Plastic fracture mechanics prediction of fracture instability in a circumferentially cracked pipe in bending - 1. J-integral analysis

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Kanninen, M.F.

1981-11-01

A method of evaluating the J-integral for a circumferentially cracked pipe in bending is proposed. The method allows a J-resistance curve to be evaluated directly from the load-displacement record obtained in a pipe fracture experiment. It permits an analysis for fracture instability in a circumferential crack growth using a J-resistance curve and the tearing modulus parameter. The influence of the system compliance on fracture instability is discussed in conjunction with the latter application. The importance of using a J-resistance curve that is consistent with the type of constraint for a given application is emphasized. The possibility of a pipe fracture emanating from a stress corrosion crack in the heat-affected zones of girth-welds in Type 304 stainless steel pipes was investigated. The J-resistance curve was employed. A pipe fracture experiment was performed using a spring-loaded four-point bending system that simulated an 8.8-m long section of unsupported 102-mm-dia pipe. An initial through-wall crack of length equal to 104 mm was used. Fracture instability was predicted to occur between 15.2 and 22.1 mm of stable crack growth at each tip. In the actual experiment, the onset of fracture instability occurred beyond maximum load at an average stable crack growth of 11.7 to 19 mm at each tip. 24 refs.

Plastic fracture mechanics prediction of fracture instability in a circumferentially cracked pipe in bending - 1. J-integral analysis

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1981-01-01

A method of evaluating the J-integral for a circumferentially cracked pipe in bending is proposed. The method allows a J-resistance curve to be evaluated directly from the load-displacement record obtained in a pipe fracture experiment. It permits an analysis for fracture instability in a circumferential crack growth using a J-resistance curve and the tearing modulus parameter. The influence of the system compliance on fracture instability is discussed in conjunction with the latter application. The importance of using a J-resistance curve that is consistent with the type of constraint for a given application is emphasized. The possibility of a pipe fracture emanating from a stress corrosion crack in the heat-affected zones of girth-welds in Type 304 stainless steel pipes was investigated. The J-resistance curve was employed. A pipe fracture experiment was performed using a spring-loaded four-point bending system that simulated an 8.8-m long section of unsupported 102-mm-dia pipe. An initial through-wall crack of length equal to 104 mm was used. Fracture instability was predicted to occur between 15.2 and 22.1 mm of stable crack growth at each tip. In the actual experiment, the onset of fracture instability occurred beyond maximum load at an average stable crack growth of 11.7 to 19 mm at each tip. 24 refs

Study on unstable fracture characteristics of light water reactor piping

International Nuclear Information System (INIS)

Kurihara, Ryoichi

1998-08-01

Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

Fracture assessment of Savannah River Reactor carbon steel piping

International Nuclear Information System (INIS)

Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

1991-01-01

The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki [and others

1997-04-01

The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program.

Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

International Nuclear Information System (INIS)

Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki

1997-01-01

The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program

Plastic fracture mechanics prediction of fracture instability in a circumferentially cracked pipe in bending--1. J-integral analysis

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1980-01-01

A method of evaluating the J-integral for a circumferentially cracked pipe in bending is proposed. The method allows a J-resistance curve to be evaluated directly from the load-displacement record obtained in a pipe fracture experiment. This method also permits an analysis for fracture instability in a circumferential crack growth using a J-resistance curve and the tearing modulus parameter. The importance of using a J-resistance curve that is consistent with the type of constraint for a given application is emphasized. 18 refs

Fracture evaluation of a crack in the service water piping system to an emergency diesel generator

International Nuclear Information System (INIS)

Rudland, D.; Scott, P.; Rahman, S.; Wilkowski, G.

1995-01-01

A pipe fracture experiment was conducted on a section of 6-inch nominal diameter pipe which was degraded by microbiologically induced corrosion (MIC) at a circumferential girth weld. The pipe was a section of one of the service water piping system to one of the emergency diesel generators at the Haddam Neck (Connecticut Yankee) plant. The experimental results will help validate future ASME Section XI pipe flaw evaluation criteria for other than Class 1 piping. A critical aspect of this experiment was an assessment of the degree of conservatism embodied in the ASME definition of flaw size. The ASME flaw size definition assumes a rectangular shaped, constant depth flaw with a depth equal to its maximum depth for its entire length. Since most service flaws are very irregular in shape, this definition can be very conservative. Alternative equivalent flaw size definitions for irregular shaped flaws are explored in this paper. (author). 7 refs., 2 figs., 4 tabs

Fracture toughness evaluations of TP304 stainless steel pipes

International Nuclear Information System (INIS)

Rudland, D.L.; Brust, F.W.; Wilkowski, G.M.

1997-02-01

In the IPIRG-1 program, the J-R curve calculated for a 16-inch nominal diameter, Schedule 100 TP304 stainless steel (DP2-A8) surface-cracked pipe experiment (Experiment 1.3-3) was considerably lower than the quasi-static, monotonic J-R curve calculated from a C(T) specimen (A8-12a). The results from several related investigations conducted to determine the cause of the observed toughness difference are: (1) chemical analyses on sections of Pipe DP2-A8 from several surface-cracked pipe and material property specimen fracture surfaces indicate that there are two distinct heats of material within Pipe DP2-A8 that differ in chemical composition; (2) SEN(T) specimen experimental results indicate that the toughness of a surface-cracked specimen is highly dependent on the depth of the initial crack, in addition, the J-R curves from the SEN(T) specimens closely match the J-R curve from the surface-cracked pipe experiment; (3) C(T) experimental results suggest that there is a large difference in the quasi-static, monotonic toughness between the two heats of DP2-A8, as well as a toughness degradation in the lower toughness heat of material (DP2-A8II) when loaded with a dynamic, cyclic (R = -0.3) loading history

Evaluation of flawed-pipe experiments: Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.

1986-11-01

The purpose of this work was to perform elastic plastic fracture mechanics evaluations of experimental data that have become available from the NRC Degraded Pipe Program, Phase II (DPII) and other NRC and EPRI sponsored programs. These evaluations were used to assess flaw evaluation procedures for austenitic and ferritic steel piping. The results also have application to leak before break fracture mechanics analysis. An improved relationship was developed for computing the J-Integral for pipes containing throughwall flaws and loaded in pure bending. The results from several DPII experiments were compared to predictions based on new J estimation scheme solutions for circumferential, finite length part-throughwall flaws in pipes with bending loading. Comparisons of experimental maximum loads with those predicted using procedures in Paragraph IWB-3640, Section XI of the ASME Code indicate that the Code flaw evaluation procedures and allowables for austenitic steel pipe are appropriate and conservative. However, the comparisons also indicate that the base metal Code allowable loads may be about 15 to 20% high for small diameter piping (less than 8-inch diameter) at allowable a/t larger than about 0.5. The work further indicates that there is justification for reducing the conservatism in IWB-3640 allowable flaw sizes and loads for austenitic steel pipe with submerged or shielded metal arc welds.

Ductile fracture behaviour of primary heat transport piping material ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

Abstract. Design of primary heat transport (PHT) piping of pressurised heavy water reactors (PHWR) has to ensure implementation of leak-before-break con- cepts. In order to be able to do so, the ductile fracture characteristics of PHT piping material have to be quantified. In this paper, the fracture resistance of SA333, Grade.

Survey on application of probabilistic fracture mechanics approach to nuclear piping

International Nuclear Information System (INIS)

Kashima, Koichi

1987-01-01

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

Concepts for benchmark problem development for fracture mechanics application in safety evaluation of nuclear piping in subcreep service

International Nuclear Information System (INIS)

Reich, M.; Esztergar, E.P.; Erdogan, F.; Gray, T.G.F.; Spence, J.

1979-01-01

This report provides basic concepts and a review of the problem areas associated with the development of analytical and experimental programs for a systematic evaluation and comparison of the currently available fracture mechanics theories. The basis for such an evaluation is conceived as a series of benchmark problems which are accurately specified examples of geometry, loading, and environmental conditions, characteristic of large diameter thin wall piping systems in nuclear service. Starting from the simplest test coupons for cracked plate specimens, the program is to be designed in such a way that the range of validity and relative merit of the competing assessment methods can be evaluated and the results applied to increasingly more complex test configurations and ultimately to real piping systems. (Auth.)

Plastic fracture mechanics prediction of fracture instability in a circumferentially cracked pipe in bending--2. Experimental verification on a Type 304 stainless steel pipe

International Nuclear Information System (INIS)

Wilkowski, G.M.; Zahoor, A.; Kanninen, M.F.

1980-01-01

The possibility of a pipe fracture emanating from a stress corrosion crack in the heat-affected zones of girth-welds in Type 304 stainless steel pipes was investigated. The J-resistance curve--tearing modulus parameter for the prediction of crack initiation, stable growth and fracture instability--was employed. In the actual experiment, the onset of fracture instability occurred beyond maximum load at an average stable crack growth of 16 to 19 mm (0.63 to 0.75-in.) at each tip. 6 refs

Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

International Nuclear Information System (INIS)

Lee, Sang-Min; Chang, Yoon-Suk; Choi, Jae-Boong; Kim, Young-Jin

2006-01-01

The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition

Unified pipe network method for simulation of water flow in fractured porous rock

Science.gov (United States)

Ren, Feng; Ma, Guowei; Wang, Yang; Li, Tuo; Zhu, Hehua

2017-04-01

Rock masses are often conceptualized as dual-permeability media containing fractures or fracture networks with high permeability and porous matrix that is less permeable. In order to overcome the difficulties in simulating fluid flow in a highly discontinuous dual-permeability medium, an effective unified pipe network method is developed, which discretizes the dual-permeability rock mass into a virtual pipe network system. It includes fracture pipe networks and matrix pipe networks. They are constructed separately based on equivalent flow models in a representative area or volume by taking the advantage of the orthogonality of the mesh partition. Numerical examples of fluid flow in 2-D and 3-D domain including porous media and fractured porous media are presented to demonstrate the accuracy, robustness, and effectiveness of the proposed unified pipe network method. Results show that the developed method has good performance even with highly distorted mesh. Water recharge into the fractured rock mass with complex fracture network is studied. It has been found in this case that the effect of aperture change on the water recharge rate is more significant in the early stage compared to the fracture density change.

Tensile and fracture properties of primary heat transport system piping material

International Nuclear Information System (INIS)

Singh, P.K.; Chattopadhyay, J.; Kushwaha, H.S.

1997-07-01

The fracture mechanics calculations in leak-before-break analysis of nuclear piping system require material tensile data and fracture resistance properties in the form of J-R curve. There are large variations in fracture parameters due to variation in chemical composition and process used in making the steel components. Keeping this in view, a comprehensive program has been planned to generate the material data base for primary heat transport system piping using the specimens machined from actual pipes used in service. The material under study are SA333 Gr.6 (base as well as weld) and SA350 LF2 (base). Since the operating temperatures of 500 MWe Indian PHWR PHT system piping range from 260 degC to 304 degC the test temperature chosen are 28 degC, 200 degC, 250 degC and 300 degC. Tensile and compact tension specimens have been fabricated from actual pipe according to ASTM standard. Fracture toughness of base metal has been observed to be higher compared to weld metal in SA333 Gr.6 material for the temperature under consideration. Fracture toughness has been observed to be higher for LC orientation (notch in circumferential direction) compared to CL orientation (notch is in longitudinal direction) for the temperature range under study. Fracture toughness value decreases with increase in temperature for the materials under study. Finally, chemical analysis has been carried out to investigate the reason for high toughness of the material. It has been concluded that low percentage of carbon and nitrogen, low inclusion rating and fine grain size has enhanced the fracture toughness value

Effect of combined loading on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Chung Yeonki

1999-01-01

Considering a rational maintenance rule of Light Water Reactor piping, reliable flaw evaluation criteria are essential to determine how a detected flaw is detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes to be considered for carbon steel piping, and can be analyzed by the elastic-plastic fracture mechanics. Currently the analytical results are provided as flaw evaluation criteria using load correction factors such like the Z-factor in ASME Code Section 6. The present correction factors were conventionally determined taken a conservatism and a simplicity into account, however, the effect of internal pressure which would be an important factor under an actual plant condition was not adequately considered. Recently, a J-estimation scheme, 'LBB.ENGC' for ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was newly developed to have a better prediction with more realistic manner. This method is explicitly incorporated the contribution of both bending and tension due to internal pressure by means of the scheme compatible with an arbitrary combined loading history. In this paper, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. A correction factor based on the new J-estimation scheme was compared with the present correction factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of internal pressure. (author)

Fracture studies on stainless steel straight pipes under earthquake-type cyclic loading

International Nuclear Information System (INIS)

Raghava, G.; Vishnuvardhan, S.; Gandhi, P.; Vaze, K.K.

2014-01-01

In order to study the crack growth and cyclic fracture behaviour, which are required for realistic assessment of Leak Before Break (LBB) applicability, experimental investigations were carried out on straight pipes under quasi-crystal loading. Totally 13 pipes were tested; three were stainless steel welded (SSW) using conventional shielded metal arc welding (SMAW) technique and the remaining specimens were Narrow Gap Welded (NGW). The fracture tests were carried out under load control, displacement control and combination of the two; the pipes were subjected to different amplitudes of load or load-line displacement (LLD), which were decided based on the response of the pipes under monotonic loading. Cyclic tearing and crack growth studies on eight straight pipes of the same material reported earlier in published literature are also considered for studying the results and understanding the behaviour. Under load control, with almost equal load amplitude, the NGW pipe exhibited improved life in comparison with SMAW pipe when both are subjected to cyclic loading. The crack growth and tearing instability behaviour of the pipes were studied. The same were found to be different for load control, displacement control and combined control tests. Based in the load-controlled experimental results, material specific plot between cyclic load amplitude (as a percentage of maximum load carrying capacity of a specimen under monotonic fracture) and number of cycles to failure was obtained. The results indicate that the piping components subjected to quasi-cyclic loading may fail in very less number of cycles even when the load amplitude is sufficiently below the monotonic fracture/collapse load. These studies will be helpful in designing nuclear power plant (NPP) piping components subjected to earthquake-type cyclic loading. (author)

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

User's manuals of probabilistic fracture mechanics analysis code for aged piping, PASCAL-SP

International Nuclear Information System (INIS)

Itoh, Hiroto; Nishikawa, Hiroyuki; Onizawa, Kunio; Kato, Daisuke; Osakabe, Kazuya

2010-03-01

As a part of research on the material degradation and structural integrity assessment for aged LWR components, a PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welded joints of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) and fatigue crack growth in piping, according to the approaches of NISA and JSME FFS Code. 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 latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This code has been developed for a cross-check use by the regulatory body in Japan. In addition to this, this code can also be used for a research purpose by researchers in academia and industries. This report provides the user's manual and theoretical background of the code. (author)

Evaluation of J-integral estimation scheme for flawed throughwall pipes

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses.

Evaluation of J-integral estimation scheme for flawed throughwall pipes

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses. (orig.)

Fracture mechanics evaluation for the cast duplex stainless steel after thermal aging

Energy Technology Data Exchange (ETDEWEB)

Urata, Shigeru [Kansai Electric Power Co., Inc., Osaka (Japan)

1998-12-31

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years. (author)

Elastic-plastic fracture analysis of carbon steel piping using the latest CEGB R6 approach

International Nuclear Information System (INIS)

Kanno, S.; Hasegawa, K.; Shimizu, T.; Kobayashi, H.

1991-01-01

The elastic-plastic fracture of carbon steel piping having various pipe diameters and circumferential crack angles and subjected to a bending moment is analyzed using the latest United Kingdom Central Electricity Generating Board R6 approach. The elastic-plastic fracture criterion must be applied instead of the plastic collapse criterion with increase of the pipe diameter and the crack angle. A simplified elastic-plastic fracture analysis procedure based on the R6 approach is proposed. (author)

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

International Nuclear Information System (INIS)

Tujikura, Y.; Urata, S.

1999-01-01

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel, which is excellent in terms of strength, corrosion resistance and weldability, has conventionally been used. Cast duplex stainless steel contains the ferrite phase in the austenite matrix, and thermal aging after long-term service is known to decrease fracture toughness. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secure, even when such through-wall crack length is assumed to be as large as the fatigue crack length grown for a service period of up to 60 years. (orig.)

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

Energy Technology Data Exchange (ETDEWEB)

Tujikura, Y.; Urata, S. [Kansai Electr. Power Co., Inc., Osaka (Japan). General Office of Nucl. and Fossil Power Production

1999-07-01

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel, which is excellent in terms of strength, corrosion resistance and weldability, has conventionally been used. Cast duplex stainless steel contains the ferrite phase in the austenite matrix, and thermal aging after long-term service is known to decrease fracture toughness. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secure, even when such through-wall crack length is assumed to be as large as the fatigue crack length grown for a service period of up to 60 years. (orig.)

Combined loading effects on the fracture mechanics behavior of line pipes

Energy Technology Data Exchange (ETDEWEB)

Bravo, R.E.; Cravero, S.; Ernst, H.A. [Tenaris Group, Campana (Argentina). SIDERCA R and D Center

2009-12-19

For certain applications, pipelines may be submitted to biaxial loading situations. In these cases, it is not clear the influence of the biaxial loading on the fracture mechanics behavior of cracked pipelines. For further understanding of biaxial loading effects, this work presents a numerical simulation of ductile tearing in a circumferentially surface cracked pipe under biaxial loading using the computational cell methodology. The model was adjusted with experimental results obtained in laboratory using single edge cracked under tension (SENT) specimens. These specimens appear as the better alternative to conventional fracture specimens to characterize fracture toughness of cracked pipes. The negligible effect of biaxial loadings on resistance curves was demonstrated. To guarantee the similarities of stress and strains fields between SENT specimens and cracked pipes subjected to biaxial loading, a constraint study using the J-Q methodology and the h parameter was used. The constraint study gives information about the characteristics of the crack-tip conditions. (author)

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

International Nuclear Information System (INIS)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size? The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size. The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service.

Comparison of fracture toughness values from large-scale pipe system tests and C(T) specimens

International Nuclear Information System (INIS)

Olson, R.; Scott, P.; Marschall, C.; Wilkowski, G.

1993-01-01

Within the International Piping Integrity Research Group (IPIRG) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe and the experiments were conducted at 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment of standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria. 4 refs., 14 figs., 1 tab

Ductile fracture of circumferentially cracked pipes subjected to bending loads

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1981-01-01

A plastic fracture mechanics methodology is presented for part-through cracks in pipes under bending. A previous analysis result on the behavior of part-through cracks in pipes is reviewed. Example quantitative results for the initiation and instability of radial growth of part-through cracks are presented and compared with the experimental data to demonstrate the applicability of the method. The analyses in our previous work are further developed to include the instability of circumferential growth of part-through cracks. Numerical results are then presented for a compliant piping system, under displacement controlled bending, which focus on (1) instability of radial growth (unstable wall breakthrough) and (2) instability of circumferential growth of the resulting throughthe-thickness crack. The combined results of the above two types of analyses are presented on a safety assessment diagram. This diagram defines a curve of critical combination of length and depth of part-through cracks which delineates leak from fracture. The effect of piping compliance on the leak-before-break assessment is discussed

Ductile fracture of circumferentially cracked pipes subjected to bending loads

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Kanninen, M.F.

1981-10-01

A plastic fracture mechanics methodology is presented for part-through cracks in pipes under bending. A previous analysis result on the behavior of part-through cracks in pipes is reviewed. Example quantitative results for the initiation and instability of radial growth of part-through cracks are presented and compared with the experimental data to demonstrate the applicability of the method. The analyses in our previous work are further developed to include the instability of circumferential growth of part-through cracks. Numerical results are then presented for a compliant piping system, under displacement controlled bending, which focus on (1) instability of radial growth (unstable wall breakthrough) and (2) instability of circumferential growth of the resulting throughthe-thickness crack. The combined results of the above two types of analyses are presented on a safety assessment diagram. This diagram defines a curve of critical combination of length and depth of part-through cracks which delineates leak from fracture. The effect of piping compliance on the leak-before-break assessment is discussed.

Evaluation methods of vibration stress of small bore piping

Energy Technology Data Exchange (ETDEWEB)

Hiramatsu, Miki; Sasaki, Toru [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Fatigue fracture by vibration stress is one of the main causes of troubles which occur at small bore piping in nuclear power plants. Therefore at the plants they manage small bore piping using a method in which their vibration accelerations are measured and the vibration stresses are calculated. In this work, vibration tests for two sets of mock-ups simulating actual piping in the plants by sinusoidal oscillation and by that obtained at an actual plant were carried out, and then an evaluation method was developed to obtain proper value of vibration stress from the measured data by the vibration tests. In comparison of the vibration stress obtained from the measured acceleration with that directly measured using strain gauges, it is confirmed that accurate vibration stress can be evaluated by a formula in which the real center of gravity of small bore piping and the acceleration of main (system) piping are considered. (author)

Ductile fracture behavior of 6-inch diameter type 304 stainless steel and STS 42 carbon steel piping containing a through-wall or part-through crack

International Nuclear Information System (INIS)

Shibata, Katsuyuki; Ohba, Toshihiro; Kawamura, Takaichi; Miyazono, Shohachiro; Kaneko, Tadashi; Yokoyama, Norio.

1986-05-01

The double ended guillotine break philosophy in the design base accident of the nuclear power plant is considered to be overly conservative from the view point of piping design. Through the past experiences and developments of the fabrication, inspection, and operation of nuclear power plants, it has been recognized that the Leak-Before-Break (LBB) concept can be justified in the LWR pressure boundary pipings. In order to verify the LBB concept, extensive experimental and theoretical works are being conducted in many countries. Furthermore, a revised piping design standard, in which LBB concept is introduced, is under preparation in Japan, U.S.A., and European countries. At JAERI, a research program to investigate the unstable ductile fracture behavior of LWR piping under bending load has been carried out as a part of the LBB verification researches since 1983. This report summarizes the result of the ductile fracture tests conducted at room temperature in 1983 and 84. The 6-inch diameter pipes of type 304 stainless steel and STS 42 carbon steel pipe with a through-wall or part-through crack were tested under bending load with low or high compliance condition at room temperature. Pipe fracture data were obtained from the test as regards to load- displacement curve, crack extension, net section stress, J-resistance curve, and so on. Besides, the influence of the compliance on the fracture behavior was examined. Discussions are performed on the ductile pipe fracture criterion, flaw evaluation criterion, and LBB evaluation method. (author)

Residual stress state in pipe cut ring specimens for fracture toughness testing

Energy Technology Data Exchange (ETDEWEB)

Damjanovic, Darko [J.J. Strossmayer Univ. of Osijek, Slavonski Brod (Croatia). Mechanical Engineering Faculty; Kozak, Drazan [Zagreb Univ. (Croatia). Dept. for Mechanical Design; Marsoner, Stefan [Materials Center, Leoben (Austria).; Gubeljak, Nenad [Maribor Univ. (Slovenia). Chair of Mechanics

2017-07-01

Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

Residual stress state in pipe cut ring specimens for fracture toughness testing

International Nuclear Information System (INIS)

Damjanovic, Darko; Kozak, Drazan; Marsoner, Stefan; Gubeljak, Nenad

2017-01-01

Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

Structural integrity evaluation of nuclear piping cracket

International Nuclear Information System (INIS)

Cadiz Deleito, J.C.

1985-01-01

The methodology to evaluation of cracks in nuclear piping is exposed. Linear elastic fracture mechanic is used to prediction of growing crack and the net section collapse theory compared with acceptation criteria of both ASME III and ASME XI code. A case allowable under ASME XI criteria is analysed under ASME III requirements. Consideration must be given to local phenomenon in crack area and local stress evaluated and compared with ASME III acceptation criteria. (author)

Current results for the NRC's short cracks in piping and piping welds research program

International Nuclear Information System (INIS)

Wilkowski, G.; Krishnaswamy, P. Brust, F.; Francini, R.; Ghadiali, N.; Kilinski, T.; Marschall, C.; Rahman, S.; Rosenfield, A.; Scott, P.

1994-01-01

The overall objective of the Short Cracks in Piping and Piping Welds Program is to verify and improve engineering analyses to predict the fracture behavior of circumferentially cracked pipe under quasi-static loading with particular attention to crack lengths typically used in LBB or flaw evaluation criteria. The program consists of 8 technical tasks as listed below. Task 1 Short through-wall-cracked (TWC) pipe evaluations. Task 2 Short surface-cracked pipe evaluations. Task 3 Bi-metallic weld crack evaluations. Task 4 Dynamic strain aging and crack instabilities. Task 5 Fracture evaluations of anisotropic pipe. Task 6 Crack-opening-area evaluations. Task 7 NRCPIPE Code improvements. Task 8 Additional efforts. Since the last WRSM meeting several additional tasks have been initiated in this program. These are discussed in Task 8. Based on results to date, the first seven tasks have also been modified as deemed necessary. The most significant accomplishments in each of these tasks since the last WRSIM meeting are discussed below. The details of all the results presented here are published in the semiannual reports from this program

Recent evaluations of crack-opening-area in circumferentially cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Rahman, S.; Brust, F.; Ghadiali, N.; Wilkowski, G.; Miura, N.

1997-04-01

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. The leak rates depend on the crack-opening area of the through-wall crack in the pipe. In addition to LBB analyses which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section XI. The objectives of this study were to review, evaluate, and refine current predictive models for performing crack-opening-area analyses of circumferentially cracked pipes. The results from twenty-five full-scale pipe fracture experiments, conducted in the Degraded Piping Program, the International Piping Integrity Research Group Program, and the Short Cracks in Piping and Piping Welds Program, were used to verify the analytical models. Standard statistical analyses were performed to assess used to verify the analytical models. Standard statistical analyses were performed to assess quantitatively the accuracy of the predictive models. The evaluation also involved finite element analyses for determining the crack-opening profile often needed to perform leak-rate calculations.

Numerical studies of temperature effect on the extrusion fracture and swell of plastic micro-pipe

Science.gov (United States)

Ren, Zhong; Huang, Xingyuan; Xiong, Zhihua

2018-03-01

Temperature is a key factor that impacts extrusion forming quality of plastic micro-pipe. In this study, the effect of temperature on extrusion fracture and swell of plastic micro-pipe was investigated by numerical method. Under a certain of the melt’s flow volume, the extrusion pattern, extrusion swelling ratio of melt are obtained under different temperatures. Results show that the extrusion swelling ratio of plastic micro-pipe decreases with increasing of temperature. In order to study the reason of temperature effect, the physical distributions of plastic micro-pipe are gotten. Numerical results show that the viscosity, pressure, stress value of melt are all decreased with the increasing of temperature, which leads to decrease the extrusion swell and fracture phenomenon for the plastic micro-pipe.

State-of-the-art report on piping fracture mechanics

International Nuclear Information System (INIS)

Wilkowski, G.M.; Olson, R.J.; Scott, P.M.

1998-01-01

This report is an in-depth summary of the state-of-the-art in nuclear piping fracture mechanics. It represents the culmination of 20 years of work done primarily in the US, but also attempts to include important aspects from other international efforts. Although the focus of this work was for the nuclear industry, the technology is also applicable in many cases to fossil plants, petrochemical/refinery plants, and the oil and gas industry. In compiling this detailed summary report, all of the equations and details of the analysis procedure or experimental results are not necessarily included. Rather, the report describes the important aspects and limitations, tells the reader where he can go for further information, and more importantly, describes the accuracy of the models. Nevertheless, the report still contains over 150 equations and over 400 references. The main sections of this report describe: (1) the evolution of piping fracture mechanics history relative to the developments of the nuclear industry, (2) technical developments in stress analyses, material property aspects, and fracture mechanics analyses, (3) unresolved issues and technically evolving areas, and (4) a summary of conclusions of major developments to date

State-of-the-art report on piping fracture mechanics

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.M.; Olson, R.J.; Scott, P.M. [Battelle, Columbus, OH (United States)

1998-01-01

This report is an in-depth summary of the state-of-the-art in nuclear piping fracture mechanics. It represents the culmination of 20 years of work done primarily in the US, but also attempts to include important aspects from other international efforts. Although the focus of this work was for the nuclear industry, the technology is also applicable in many cases to fossil plants, petrochemical/refinery plants, and the oil and gas industry. In compiling this detailed summary report, all of the equations and details of the analysis procedure or experimental results are not necessarily included. Rather, the report describes the important aspects and limitations, tells the reader where he can go for further information, and more importantly, describes the accuracy of the models. Nevertheless, the report still contains over 150 equations and over 400 references. The main sections of this report describe: (1) the evolution of piping fracture mechanics history relative to the developments of the nuclear industry, (2) technical developments in stress analyses, material property aspects, and fracture mechanics analyses, (3) unresolved issues and technically evolving areas, and (4) a summary of conclusions of major developments to date.

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)

Experimental study on dynamic pipe fracture in consideration of hydropower plant model

Directory of Open Access Journals (Sweden)

Kazumi Ishikawa

2009-12-01

Full Text Available In the case of sudden valve closure, water hammer creates the most powerful pressure and damage to pipeline systems. The best way to protect the pipeline system is to eliminate water hammer. The main reasons for water hammer occurrence are valve closure, high initial velocity, and static pressure. However, it is difficult to eliminate water hammer. Water hammer tends to occur when the valve is being closed. In this study, the pipe fracture caused by static water pressure, gradually increasing pressure, and suddenly increasing pressure were compared experimentally in a breaking PVC test pipe. The quasi-static zone, the dynamic zone, and the transition zone are defined through the results of those experiments, with consideration of the fracture patterns of test pipes and impulses. The maximum pressure results were used to design the pipeline even though it is in the dynamic zone.

Low-frequency fluid waves in fractures and pipes

Energy Technology Data Exchange (ETDEWEB)

Korneev, Valeri

2010-09-01

Low-frequency analytical solutions have been obtained for phase velocities of symmetrical fluid waves within both an infinite fracture and a pipe filled with a viscous fluid. Three different fluid wave regimes can exist in such objects, depending on the various combinations of parameters, such as fluid density, fluid viscosity, walls shear modulus, channel thickness, and frequency. Equations for velocities of all these regimes have explicit forms and are verified by comparisons with the exact solutions. The dominant role of fractures in rock permeability at field scales and the strong amplitude and frequency effects of Stoneley guided waves suggest the importance of including these wave effects into poroelastic theories.

Evaluation of flaws in carbon steel piping. Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.; Mehta, H.S.; Yukawa, S.; Ranganath, S.

1986-10-01

The objective of this program was to develop flaw evaluation procedures and allowable flaw sizes for ferritic piping used in light water reactor (LWR) power generation facilities. The program results provide relevant ASME Code groups with the information necessary to define flaw evaluation procedures, allowable flaw sizes, and their associated bases for Section XI of the code. Because there are several possible flaw-related failure modes for ferritic piping over the LWR operating temperature range, three analysis methods were employed to develop the evaluation procedures. These include limit load analysis for plastic collapse, elastic plastic fracture mechanics (EPFM) analysis for ductile tearing, and linear elastic fracture mechanics (LEFM) analysis for non ductile crack extension. To ensure the appropriate analysis method is used in an evaluation, a step by step procedure also is provided to identify the relevant acceptance standard or procedure on a case by case basis. The tensile strength and toughness properties required to complete the flaw evaluation for any of the three analysis methods are included in the evaluation procedure. The flaw evaluation standards are provided in tabular form for the plastic collapse and ductile tearing modes, where the allowable part through flaw depth is defined as a function of load and flaw length. For non ductile crack extension, linear elastic fracture mechanics analysis methods, similar to those in Appendix A of Section XI, are defined. Evaluation flaw sizes and procedures are developed for both longitudinal and circumferential flaw orientations and normal/upset and emergency/faulted operating conditions. The tables are based on margins on load of 2.77 and 1.39 for circumferential flaws and 3.0 and 1.5 for longitudinal flaws for normal/upset and emergency/faulted conditions, respectively.

Evaluation of flaws in carbon steel piping. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Gamble, R.M.; Mehta, H.S.; Yukawa, S.; Ranganath, S.

1986-10-01

The objective of this program was to develop flaw evaluation procedures and allowable flaw sizes for ferritic piping used in light water reactor (LWR) power generation facilities. The program results provide relevant ASME Code groups with the information necessary to define flaw evaluation procedures, allowable flaw sizes, and their associated bases for Section XI of the code. Because there are several possible flaw-related failure modes for ferritic piping over the LWR operating temperature range, three analysis methods were employed to develop the evaluation procedures. These include limit load analysis for plastic collapse, elastic plastic fracture mechanics (EPFM) analysis for ductile tearing, and linear elastic fracture mechanics (LEFM) analysis for non ductile crack extension. To ensure the appropriate analysis method is used in an evaluation, a step by step procedure also is provided to identify the relevant acceptance standard or procedure on a case by case basis. The tensile strength and toughness properties required to complete the flaw evaluation for any of the three analysis methods are included in the evaluation procedure. The flaw evaluation standards are provided in tabular form for the plastic collapse and ductile tearing modes, where the allowable part through flaw depth is defined as a function of load and flaw length. For non ductile crack extension, linear elastic fracture mechanics analysis methods, similar to those in Appendix A of Section XI, are defined. Evaluation flaw sizes and procedures are developed for both longitudinal and circumferential flaw orientations and normal/upset and emergency/faulted operating conditions. The tables are based on margins on load of 2.77 and 1.39 for circumferential flaws and 3.0 and 1.5 for longitudinal flaws for normal/upset and emergency/faulted conditions, respectively

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.

Application of fracture mechanics leak-before-break analyses for protection against pipe rupture in SEP plants

International Nuclear Information System (INIS)

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

1984-01-01

In accordance with the latest NRC guidance the leak-before-break technique was evaluated for high-energy piping systems in a nuclear power plant. The elements of this evaluation include determination of: 1) largest crack size which will remain stable; 2) leak rate resulting from a crack with length twice the pipe wall thickness; 3) size of crack which will leak at a rate greater than 1 gpm, if 2) results in less than 1 gpm; and 4) analysis of part-through cracks for subcritical crack growth rates to establish in-service inspection (ISI) intervals. Conclusions reached are: 1) The fracture mechanics leak-before-break approach is shown as a viable option to prevent pipe rupture. 2) Austenitic stainless steel pipes possess significant toughness, and large cracks are required for rupture. 3) The net section plastic collapse analysis is more conservative than tearing modulus evaluations. 4) Leak rates are large enough to assure detection well before cracks reach a critical size. 5) In the case studied, subcritical crack growth is slow enough to require ISI intervals of about 10 years to detect part-through cracks

Evaluation of mechano-chemical degradation induced stresses of polyolefin pipes

Energy Technology Data Exchange (ETDEWEB)

Choi, Byoung Ho [Korea Univ., Seoul (Korea, Republic of); Chudnovsky, Alexander [The University of Illinois, Chicago (United States)

2008-07-01

The fracture phenomena in engineering thermoplastics resulting from chemical degradation is usually observed in the form of a microcrack network within a surface layer of degraded polymer exposed to a combined action of mechanical stresses and chemically aggressive environment. Degradation of polymers is usually manifested in a reduction of molecular weight, increase of crystallinity in semi crystalline polymers, increase of material density, a subtle increase in yield strength, and a dramatic reduction in toughness. The critical level of degradation for fracture initiation depends on the rates of toughness deterioration and build-up of the degradation related stresses as well as on the manufacturing and service stresses. In this paper, the evaluation of mechano-chemical degradation induced stress is attempted, and the application of the evaluated stress to the fracture initiation of polymer pipes is presented.

Evaluation of mechano-chemical degradation induced stresses of polyolefin pipes

International Nuclear Information System (INIS)

Choi, Byoung Ho; Chudnovsky, Alexander

2008-01-01

The fracture phenomena in engineering thermoplastics resulting from chemical degradation is usually observed in the form of a microcrack network within a surface layer of degraded polymer exposed to a combined action of mechanical stresses and chemically aggressive environment. Degradation of polymers is usually manifested in a reduction of molecular weight, increase of crystallinity in semi crystalline polymers, increase of material density, a subtle increase in yield strength, and a dramatic reduction in toughness. The critical level of degradation for fracture initiation depends on the rates of toughness deterioration and build-up of the degradation related stresses as well as on the manufacturing and service stresses. In this paper, the evaluation of mechano-chemical degradation induced stress is attempted, and the application of the evaluated stress to the fracture initiation of polymer pipes is presented

Using acoustic emission technique to monitor fractures on the analogous pressure pipes

International Nuclear Information System (INIS)

Zhang Lichen

1989-01-01

By using the acoustic emission technique to monitor the fractures on analogous pressure pipes of the primary circuit which has had cracks and loading with pressure was investigated. The dynamical process, from cracking to fracturing, was recorded by the acoustic emission technique. Comparing with the conventional method, this method gives more informations, such as pre-cracking, cracking growing, fast fracturing and the pressure values at different phases. During testing time a microcomputer was used for real-time data processing and locating the fracturing position. These data are useful for the mechanical analysis of the reactor components

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2006-01-01

Considering a rational maintenance rule of Light Water Reactor piping, reliable flaw evaluation criteria are essential to determine how a detected flaw is detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes to be considered for carbon steel piping, and can be analyzed by the elastic-plastic fracture mechanics. Some analytical efforts have been provided as flaw evaluation criteria using load correction factors such like the Z-factors in the JSME codes on fitness-for-service for nuclear power plants or the ASME boiler and pressure vessel code section XI. The present correction factors were conventionally determined taken conservatism and simplicity into account, however, the effect of internal pressure which would be an important factor under an actual plant condition was not adequately considered. Recently, a J-estimation scheme, 'LBB. ENGC' for ductile fracture analysis of circumferentially through-wall-cracked pipes subjected combined loading was newly developed to have a better prediction with more realistic manner. This method is explicitly incorporated the contribution of both bending and tension due to internal pressure by means of the scheme compatible with an arbitrary combined loading history. In this paper, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. A correction factor based on the new J-estimation scheme was compared with the present correction factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of internal pressure. (author)

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 3. Evaluation of potential for pipe breaks

Energy Technology Data Exchange (ETDEWEB)

1984-11-01

The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations.

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 3. Evaluation of potential for pipe breaks

International Nuclear Information System (INIS)

1984-11-01

The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations

Evaluation of fatigue crack growth and fracture resistance of SA350 LF2 material

International Nuclear Information System (INIS)

Singh, P.K.; Dubey, J.S.; Chakrabarty, J.K.; Vaze, K.K.; Kushwaha, H.S.

2003-01-01

The aim of the present paper is to evaluate the tensile and fracture mechanics properties of the SA350 LF2 carbon steel material used as the Header material in the primary heat transport (PHT) system piping of the Indian pressurized heavy water reactors (PHWR). Tensile, fatigue crack growth rate and fracture toughness tests have been carried out on specimens machined from the Header of the actual PHT pipes. The effect of temperature on tensile properties has been discussed. The effect of temperature and notch orientation on fracture resistance behavior of the material and fatigue crack growth rate dependence on the notch orientation and stress ratio has also been discussed. (author)

Loading rate effects on strength and fracture toughness of pipe steels used in Task 1 of the IPIRG program

International Nuclear Information System (INIS)

Marschall, C.W.; Landow, M.P.; Wilkowski, G.M.

1993-10-01

Material characterization tests were conducted on laboratory specimens machined from pipes to determine the effect of dynamic loading (i.e., rates comparable to those for high amplitude seismic events) on tensile properties and fracture resistance at 288 C (550 F). Specimens were fabricated from seven different pipes, including carbon steels and stainless steels (both base metal and weld metal), which were to be subjected to full-scale pipe tests in IPIRG Task 1.0. For the stainless steels tested at 288 C (550 F), tensile strength was unchanged, while yield strength and fracture resistance were increased. The increase in fracture resistance was modest for the wrought base metals and substantial for the weld metal and the cast base metal. The carbon steels tested were sensitive to dynamic strain aging, and hence the strength and toughness was affected by both temperature and strain rate effects. The carbon steel base metal and welds exhibited ultimate tensile strength values at 288 C (550 F) that were greater than at room temperature. Furthermore, the ultimate tensile strength at 288 C (550 F) was lowered significantly by increased strain rate and, in the carbon steel base metals, increased strain rate also lowered the fracture resistance, substantially in the base metal of one pipe. In comparing these results to the IPIRG pipe test results to date, it was found that the trends of these tests agree well with the Subtask 1.2 quasi-static and dynamic pipe fracture experiments. Loads measured in the Subtask 1.1 pipe experiments were, however, somewhat higher than would have been expected by the trends observed in the laboratory tests

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2008-01-01

Considering a rule for the rationalization of maintenance of Light Water Reactor piping, reliable flaw evaluation criteria are essential for determining how a detected flaw will be detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes that must be considered for carbon steel piping and can be analyzed by elastic-plastic fracture mechanics. Some analytical efforts have provided various flaw evaluation criteria using load correction factors, such as the Z-factors in the JSME codes on fitness-for-service for nuclear power plants and the section XI of the ASME boiler and pressure vessel code. The present Z-factors were conventionally determined, taking conservativity and simplicity into account; however, the effect of internal pressure, which is an important factor under actual plant conditions, was not adequately considered. Recently, a J-estimation scheme, LBB.ENGC for the ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was developed for more accurate prediction under more realistic conditions. This method explicitly incorporates the contributions of both bending and tension due to internal pressure by means of a scheme that is compatible with an arbitrary combined-loading history. In this study, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. The Z-factor obtained in this study was compared with the presently used Z-factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of the internal pressure. (author)

Special fracture mechanics specimens for multilayer plastic pipes testing

Czech Academy of Sciences Publication Activity Database

Hutař, Pavel; Šestáková, Lucie; Knésl, Zdeněk; Nezbedová, E.; Náhlík, Luboš

2009-01-01

Roč. 28, č. 8 (2009), s. 785-792 ISSN 0142-9418 R&D Projects: GA ČR GA106/09/0279; GA ČR GC101/09/J027 Institutional research plan: CEZ:AV0Z20410507 Keywords : Multilayer plastic pipes * C-type specimen * K-calibration * Fracture toughness * Slow crack growth * Non-homogenous specimens Subject RIV: JL - Material s Fatigue, Friction Mechanics Impact factor: 1.667, year: 2009

Ductile failure analysis of API X65 pipes with notch-type defects using a local fracture criterion

International Nuclear Information System (INIS)

Oh, Chang-Kyun; Kim, Yun-Jae; Baek, Jong-Hyun; Kim, Young-Pyo; Kim, Woo-Sik

2007-01-01

A local failure criterion for API X65 steel is applied to predict ductile failure of full-scale API X65 pipes with simulated corrosion and gouge defects under internal pressure. The local failure criterion is the stress-modified fracture strain as a function of the stress triaxiality (defined by the ratio of the hydrostatic stress to the effective stress). Based on detailed finite element (FE) analyses with the proposed local failure criterion, burst pressures of defective pipes are estimated and compared with experimental data. For pipes with simulated corrosion defects, FE analysis with the proposed local fracture criterion indicates that predicted failure takes place after the defective pipes attain maximum loads for all cases, possibly due to the fact that the material has sufficient ductility. For pipes with simulated gouge defects, on the other hand, it is found that predicted failure takes place before global instability, and the predicted burst pressures are in good agreement with experimental data, providing confidence in the present approach

J-integral evaluation and stability analysis in the unstable ductile fracture

International Nuclear Information System (INIS)

Miyoshi, Toshiro; Yoshida, Yuichiro; Shiratori, Masaki.

1984-01-01

Concerning unstable ductile fracture, which is an important problem on the structural stability of line pipes, nuclear reactor piping and so on, the research on fracture mechanics parameters which control the beginning of the stable growth and unstable growth of cracks attracts interest. At present, as the parameters, the T-modulus based on J-integral crack tip opening angle, crack opening angle averaged over crack developing part, plastic work coefficient and so on have been proposed. The research on the effectiveness and inter-relation of these parameters is divided into generation phase and application phase, and by these researches, it was reported that all T-modulus, CTOA and COA took almost constant values in relation to crack development, except initial transition period. In order to decide which parameter is most appropriate, the detailed analysis is required. In this study, the analysis of unstable ductile fracture of a central crack test piece and a small tensile test piece was carried out by finite element method, and the evaluation of J-integral in relation to crack development, J-integral resistance value when COA is assumed to be a constant, the form of an unstable fracture occurring point and the compliance dependence were examined. The method of analysis, the evaluation of J-integral, J-integral resistance value, unstable fracture occurring point and stability diagram are described. (Kako, I.)

Leak before break piping evaluation diagram

International Nuclear Information System (INIS)

Fabi, R.J.; Peck, D.A.

1994-01-01

Traditionally Leak Before Break (LBB) has been applied to the evaluation of piping in existing nuclear plants. This paper presents a simple method for evaluating piping systems for LBB during the design process. This method produces a piping evaluation diagram (PED) which defines the LBB requirements to the piping designer for use during the design process. Several sets of LBB analyses are performed for each different pipe size and material considered in the LBB application. The results of this method are independent of the actual pipe routing. Two complete LBB evaluations are performed to determine the maximum allowable stability load, one evaluation for a low normal operating load, and the other evaluation for a high normal operating load. These normal operating loads span the typical loads for the particular system being evaluated. In developing the allowable loads, the appropriate LBB margins are included in the PED preparation. The resulting LBB solutions are plotted as a set of allowable curves for the maximum design basis load, such is the seismic load versus the normal operating load. Since the required margins are already accounted for in the LBB PED, the piping designer can use the diagram directly with the results of the piping analysis and determine immediately if the current piping arrangement passes LBB. Since the LBB PED is independent of pipe routing, changes to the piping system can be evaluated using the existing PED. For a particular application, all that remains is to confirm that the actual materials and pipe sizes assumed in creating the particular design are built into the plant

Application of the results of pipe stress analyses into fracture mechanics defect analyses for welds of nuclear piping components; Uebernahme der Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) fuer bruchmechanische Fehlerbewertungen fuer Schweissnaehte an Rohrleitungsbauteilen in kerntechnischen Anlagen

Energy Technology Data Exchange (ETDEWEB)

Dittmar, S.; Neubrech, G.E.; Wernicke, R. [TUeV Nord SysTec GmbH und Co.KG (Germany); Rieck, D. [IGN Ingenieurgesellschaft Nord mbH und Co.KG (Germany)

2008-07-01

For the fracture mechanical assessment of postulated or detected crack-like defects in welds of piping systems it is necessary to know the stresses in the un-cracked component normal to the crack plane. Results of piping stress analyses may be used if these are evaluated for the locations of the welds in the piping system. Using stress enhancing factors (stress indices, stress factors) the needed stress components are calculated from the component specific sectional loads (forces and moments). For this procedure the tabulated stress enhancing factors, given in the standards (ASME Code, German KTA regulations) for determination and limitation of the effective stresses, are not always and immediately adequate for the calculation of the stress component normal to the crack plane. The contribution shows fundamental possibilities and validity limits for adoption of the results of piping system analyses for the fracture mechanical evaluation of axial and circumferential defects in welded joints, with special emphasis on typical piping system components (straight pipe, elbow, pipe fitting, T-joint). The lecture is supposed to contribute to the standardization of a code compliant and task-related use of the piping system analysis results for fracture mechanical failure assessment. [German] Fuer die bruchmechanische Bewertung von postulierten oder bei der wiederkehrenden zerstoerungsfreien Pruefung detektierten rissartigen Fehlern in Schweissnaehten von Rohrsystemen werden die Spannungen in der ungerissenen Bauteilwand senkrecht zur Rissebene benoetigt. Hierfuer koennen die Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) genutzt werden, wenn sie fuer die Orte der Schweissnaehte im Rohrsystem ausgewertet werden. Mit Hilfe von Spannungserhoehungsfaktoren (Spannungsindizes, Spannungsbeiwerten) werden aus den komponentenweise berechneten Schnittlasten (Kraefte und Momente) die benoetigten Spannungskomponenten berechnet. Dabei sind jedoch die in den Regelwerken (ASME

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Norris, D.M.

1984-11-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior.

Ductile fracture of circumferentially cracked type-304 stainless steel pipes in tension

International Nuclear Information System (INIS)

Zahoor, A.; Norris, D.M.

1984-01-01

Circumferentially cracked pipes subjected to tensile load were analyzed for finite length and constant depth part-through cracks located at the inside of the pipe wall. The analysis postulated loads sufficient to cause net-section yielding of the flawed section. It was demonstrated that a propensity for predominantly radial growth exists for part-through cracks loaded in tension. This result is similar to the result for bend loading, except that bend loading causes more favorable conditions for wall breakthrough than tension loading. Numerical results were developed for 4-in. and 24-in-dia pipes. Safety margins for displacement controlled loads were described by a safety assessment diagram. This diagram defines a curve delineating leak from fracture in a space of nondimensional crack length and crack depth. 4-india schedule 80 Type-304 stainless steel pipes with length to radius ratio (L/R) of up to 100 exhibited leak-before-break behavior

Application of cyclic J-integral to low cycle fatigue crack growth of Japanese carbon steel pipe

Energy Technology Data Exchange (ETDEWEB)

Miura, N.; Fujioka, T.; Kashima, K. [and others

1997-04-01

Piping for LWR power plants is required to satisfy the LBB concept for postulated (not actual) defects. With this in mind, research has so far been conducted on the fatigue crack growth under cyclic loading, and on the ductile crack growth under excessive loading. It is important, however, for the evaluation of the piping structural integrity under seismic loading condition, to understand the fracture behavior under dynamic and cyclic loading conditions, that accompanies large-scale yielding. CRIEPI together with Hitachi have started a collaborative research program on dynamic and/or cyclic fracture of Japanese carbon steel (STS410) pipes in 1991. Fundamental tensile property tests were conducted to examine the effect of strain rate on tensile properties. Cracked pipe fracture tests under some loading conditions were also performed to investigate the effect of dynamic and/or cyclic loading on fracture behavior. Based on the analytical considerations for the above tests, the method to evaluate the failure life for a cracked pipe under cyclic loading was developed and verified. Cyclic J-integral was introduced to predict cyclic crack growth up to failure. This report presents the results of tensile property tests, cracked pipe fracture tests, and failure life analysis. The proposed method was applied to the cracked pipe fracture tests. The effect of dynamic and/or cyclic loading on pipe fracture was also investigated.

Calculation and evaluation methodology of the flawed pipe and the compute program development

International Nuclear Information System (INIS)

Liu Chang; Qian Hao; Yao Weida; Liang Xingyun

2013-01-01

Background: The crack will grow gradually under alternating load for a pressurized pipe, whereas the load is less than the fatigue strength limit. Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics (EPFM) criteria. Methods: In the compute, the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++. Results: The fluctuating load of the Reactor Coolant System transients, the initial flaw shape, the initial flaw orientation are all accounted here. Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not. (authors)

Causes of pipe ruptures in distribution lines. Evaluation of long-term observations in a metropolitan pipe network

Energy Technology Data Exchange (ETDEWEB)

Kottmann, A

1978-01-01

Pipe ruptures and their causes are examined from the viewpoints of pipe material, corrosion, traffic, internal pressure, air temperature, ground temperature, ground frost, gas or water temperature, and ground moisture level. The examination relies on 17 years of statistics (1958-74) from (1) Technische Werke der Stadt Stuttgart AG on 11,986 pipe ruptures and (2) German weather-service data on ground-moisture readings at depths down to 80 in. in the Stuttgart area. Faced with replacing up to 280 miles (450 km) of cast-iron gas-distribution lines that seemed extraordinarily prone to rupture (company records showed at least 20 breaks/month) after the conversion to natural gas, TWS authorized this study to determine the boundary conditions that make cast-iron pipe susceptible to fracture, thus minimizing the extent of the replacement program. The investigation showed that corrosion had only a slight effect upon cracking. No significant effect was found for any of the following: temperature-caused changes in material properties, internal pressure or pressure changes, fluctuations in gas temperature, changes in air temperature, and summertime changes in ground temperature. Stress loading by heavy traffic, however, doubled the fracture incidence.

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

International Nuclear Information System (INIS)

Gamble, R.M.; Wichman, K.R.

1997-01-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials

A simplified LBB evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Wichman, K.R.

1997-04-01

The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

International Nuclear Information System (INIS)

Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

2010-01-01

This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

Experimental and numerical approach on fracture behaviour of four inches diameter carbon-manganese cracked welded pipes in four point bending

International Nuclear Information System (INIS)

Semete, P.; Faidy, C.; Lautier, J.L.

2001-01-01

EDF has conducted a research programme to demonstrate the fracture resistance of carbon-manganese welded pipes. The main task of this programme consisted of testing three four inches diameter (114.3 mm O.D.) thin welded pipes (8.56 mm thick) which are representative of those of the sites. The three pipes were loaded under four point bending at a quasi-static rate at -20 C till their maximum bending moment was reached. This paper presents the experimental results, finite element calculations and their comparison with the simplified fracture assessment method of the RSE-M Code. (author)

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

Thermal aging evaluation of cast austenitic stainless steel pipe

International Nuclear Information System (INIS)

Song, T. H.; Jung, I. S.

2002-01-01

24 years have been passed since Kori Unit 1 began its commercial operation, and 19 years have been passed since Kori Unit 2 began its commercial operation. As the end point of design life become closer, plant life extension and periodic safety assessment is paid more and more attention to by utility company. In this paper, the methodologies and results of cast austenitic stainless steel pipe thermal aging evaluations of both units have been presented in association with aging time of 10, 20, and 30 years and operating temperature, respectively. Life extension cases respectively. As a result of this, at the operating temperature of 280 .deg. C, thermal aging was not a problem as long as Charpy V-notch room temperature minimum impact energy is concerned. However, more than 300 .deg. C and 30 years of operating condition, we should perform detailed fracture mechanics analysis with CMTR of NPP pipe

Evaluation of throughwall crack pipes under displacement controlled loading

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

Tearing modulus solutions are developed for flawed throughwall pipes subjected to displacement controlled loading. Two cases of loading were considered: (1) a displacement controlled bending loading, and (2) a displacement controlled axial tension loading. A revised version of the EPRI J-integral estimation scheme is used in the development of the solutions. These solutions can be used for the entire range of elastic-plastic loading, from linear elastic, contained yielding, to large scale yielding of the crack section. Experimental data from pipes in bending were used to assess the accuracy of the compliant loading solutions. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. These solutions are shown to have good accuracy when used to predict the experimental results. The methodology and procedure can also be applied to part-throughwall cracks. These solutions have application to the leak before break fracture mechanics analyses.

Evaluation of throughwall crack pipes under displacement controlled loading

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

Tearing modulus solutions are developed for flawed throughwall pipes subjected to displacement controlled loading. Two cases of loading were considered: (1) a displacement controlled bending loading, and (2) a displacement controlled axial tension loading. A revised version of the EPRI J-integral estimation scheme is used in the development of the solutions. These solutions can be used for the entire range of elastic-plastic loading, from linear elastic, contained yielding, to large scale yielding of the crack section. Experimental data from pipes in bending were used to assess the accuracy of the compliant loading solutions. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. These solutions are shown to have good accuracy when used to predict the experimental results. The methodology and procedure can also be applied to part-throughwall cracks. These solutions have application to the leak before break fracture mechanics analyses. (orig.)

Comparison of fracture toughness values from an IPIRG-1 large-scale pipe system test and C(T) specimens on wrought TP304 stainless steel

International Nuclear Information System (INIS)

Olson, R.J.; Scott, P.; Marschall, C.W.; Wilkowski, G.M.

1994-01-01

Within the First International Piping Integrity Research Group (IPIRG-1) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe, and the experiments were conducted at a pressure of 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. Three independent analyses were used to calculate the toughness directly from one of these pipe experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment to standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

International Nuclear Information System (INIS)

Lin, Y.C.; Xie, Y.J.; Wang, X.H.

2004-01-01

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

Specialist meeting on leak before break in reactor piping and vessels

Energy Technology Data Exchange (ETDEWEB)

Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

1997-04-01

A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

Crack stability in a representative piping system under combined inertial and seismic/dynamic displacement-controlled stresses. Subtask 1.3 final report

International Nuclear Information System (INIS)

Scott, P.; Olson, R.; Wilkowski, O.G.; Marschall, C.; Schmidt, R.

1997-06-01

This report presents the results from Subtask 1.3 of the International Piping Integrity Research Group (IPIRG) program. The objective of Subtask 1.3 is to develop data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system under combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The piping system evaluated is an expansion loop with over 30 meters of 16-inch diameter Schedule 100 pipe. The experimental facility is equipped with special hardware to ensure system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe-system experiments. The uncracked experiment was conducted to evaluate piping system damping and natural frequency characteristics. The cracked-pipe experiments evaluated the fracture behavior, pipe system response, and stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Results from all pipe-system experiments and material characterization efforts are presented. Results of fracture mechanics analyses, dynamic finite element stress analyses, and stability analyses are presented and compared with experimental results

Crack stability in a representative piping system under combined inertial and seismic/dynamic displacement-controlled stresses. Subtask 1.3 final report

Energy Technology Data Exchange (ETDEWEB)

Scott, P.; Olson, R.; Wilkowski, O.G.; Marschall, C.; Schmidt, R.

1997-06-01

This report presents the results from Subtask 1.3 of the International Piping Integrity Research Group (IPIRG) program. The objective of Subtask 1.3 is to develop data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system under combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The piping system evaluated is an expansion loop with over 30 meters of 16-inch diameter Schedule 100 pipe. The experimental facility is equipped with special hardware to ensure system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe-system experiments. The uncracked experiment was conducted to evaluate piping system damping and natural frequency characteristics. The cracked-pipe experiments evaluated the fracture behavior, pipe system response, and stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Results from all pipe-system experiments and material characterization efforts are presented. Results of fracture mechanics analyses, dynamic finite element stress analyses, and stability analyses are presented and compared with experimental results.

Leak before break evaluation for main steam piping system made of SA106 Gr.C

Energy Technology Data Exchange (ETDEWEB)

Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

1997-04-01

The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

Short cracks in piping and piping welds. Seventh program report, March 1993-December 1994. Volume 4, Number 1

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P. [Battelle, Columbus, OH (United States)

1995-04-01

This is the seventh progress report of the U.S. Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports.

Short cracks in piping and piping welds. Seventh program report, March 1993-December 1994. Volume 4, Number 1

International Nuclear Information System (INIS)

Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P.

1995-04-01

This is the seventh progress report of the U.S. Nuclear Regulatory Commission's research program entitled open-quotes Short Cracks in Piping and Piping Weldsclose quotes. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports

Plastic fracture instability analysis of wall breakthrough in a circumferentially cracked pipe subjected to bending loads

International Nuclear Information System (INIS)

Zahoor, A.; Kanninen, M.F.

1981-01-01

A method of analyzing internal surface circumferential cracks in ductile reactor piping is presented. The method utilizes an alternate but equivalent definition of the J-integral based on nonlinear structural compliance. The analysis is valid for situations where the cross section containing the crack is fully yielded. Results are obtained for radial and circumferential crack growth for pipes subjected to bending. The stability of radial crack growth (wall breakthrough) is assessed using the J-integral-based tearing modulus approach. The analysis is shown to be in agreement with experimental results on the stability of surface crack growth in Type 304 stainless stee pipes. Example quantitative results for fracture instability assessments for nuclear piping are presented. 23 refs

Plastic fracture instability analysis of wall breakthrough in a circumferentially cracked pipe subjected to bending loads

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Kanninen, M.F.

1981-07-01

A method of analyzing internal surface circumferential cracks in ductile reactor piping is presented. The method utilizes an alternate but equivalent definition of the J-integral based on nonlinear structural compliance. The analysis is valid for situations where the cross section containing the crack is fully yielded. Results are obtained for radial and circumferential crack growth for pipes subjected to bending. The stability of radial crack growth (wall breakthrough) is assessed using the J-integral-based tearing modulus approach. The analysis is shown to be in agreement with experimental results on the stability of surface crack growth in Type 304 stainless stee pipes. Example quantitative results for fracture instability assessments for nuclear piping are presented. 23 refs.

Practical application of fracture mechanics with consideration of multiaxiality of stress state to degraded nuclear piping

International Nuclear Information System (INIS)

Kussmaul, K.; Blind, D.; Herter, K.H.; Eisele, U.; Schuler, X.

1995-01-01

Within the scope of a research project nuclear piping components (T-branches and elbows) with dimensions like the primary coolant lines of PWR plants were investigated. In addition to the experimental full scale tests, extensive numerical calculations by means of the finite element method (FEM) as well as fracture mechanics analyses were performed. The applicability of these methods was verified by comparison with the experimental results. The calculation of fracture mechanics parameters as well as the calculated component stress enabled a statement on crack initiation. The failure behavior could be evaluated by means of the multiaxiality of stress state in the ligament (gradient of the quotient of the multiaxiality of stress state q). With respect to practical application on other pressurized components it is shown how to use the procedure (e.g. in a LBB analysis). A quantitative assessment with regard to crack initiation is possible by comparison of the effective crack initiation value J ieff with the calculated component stress. If the multiaxiality of stress state and the q gradient in the ligament of the fracture ligament of the fracture mechanics specimen and the pressurized component to be evaluated is comparable a quantitative assessment is possible as for crack extension and maximum load. If there is no comparability of the gradients a qualitative assessment is possible for the failure behavior

Probabilistic procedure to evaluate integrity of degraded pipes under internal pressure and bending moment

International Nuclear Information System (INIS)

Roos, E.; Herter, K.-H.; Julisch, P.; Otremba, F.; Schuler, X.

2003-01-01

The determination of critical crack sizes or permissible/allowable loading levels in pipes with degraded pipe sections (circumferential cracks) for the assurance of component integrity is usually based on deterministic approaches. Therefore along with numerical calculational methods (finite element (FE) analyses) limit load calculations, such as e.g. the 'Plastic limit load concept' and the 'Flow stress concept' as well as fracture mechanics approximation methods as e.g. the R-curve method or the 'Ductile fracture handbook' and the R6-Method are currently used for practical application. Numerous experimental tests on both ferritic and austenitic pipes with different pipe dimensions were investigated at MPA Stuttgart. The geometries of the pipes were comparable to actual piping systems in Nuclear Power Plants, both BWR as well as PWR. Through wall cracks and part wall through cracks on the inside surface of the pipes were considered. The results of these tests were used to determine the flow stresses used within the limit load calculations. Therefore the deterministic concepts assessing the integrity of degraded pipes are available A new post-calculation of the above mentioned tests was performed using probabilistic approaches to assure the component integrity of degraded piping systems. As a result the calculated probability of failure was compared to experimental behaviour during the pipe test. Different reliability techniques were used for the verification of the probabilistic approaches. (author)

Fracture mechanics assessment of thermal aged nuclear piping based on the Leak-Before-Break concept

Energy Technology Data Exchange (ETDEWEB)

Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Yu, Weiwei [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen PSI (Switzerland); Wang, Rongshan; Lu, Feng; Zhang, Guodong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China)

2016-05-15

Highlights: • The effects of thermal aging on crack unstable tearing are studied. • The critical size of crack unstable tearing is calculated by different methods. • The critical failure models are compared. • The conservatism of J–T diagram is shown. - Abstract: The Leak-Before-Break (LBB) concept has been accepted to design the primary piping system of the pressurized water reactor (PWR). Due to thermal aging of long term operation, the cast stainless steels (CSSs) which are used for the primary piping of PWR, suffer a significant loss of fracture toughness, and as a consequence the safety margin of the thermal aged pipe decreases. Therefore, the aged piping should be analyzed and validated by the LBB concept. In this paper, elastic–plastic fracture mechanics (EPFM) assessments of the thermal aged piping are presented according to the LBB concept. The critical break size of crack unstable tearing is calculated by the EPFM method. The crack driving force diagram (J–a diagram), the stability assessment diagram (J–T diagram) and a numerical method are applied to calculate the critical crack size of crack break. The effects of thermal aging on the plastic limit load, J–T diagram, critical crack size of the EPFM and the critical failure mode are studied. The results show that the thermal aging effect decreases the maximum allowed J-integral at a certain ductile tearing modulus by more than 50% and it increases the flow stress and plastic limit load by 11.78%. The results based on the J–T diagram are about 40% conservative than those based on the direct numerical method for the high loading case. For the thermal aged piping, it is important to consider the competition failure modes between plastic collapse and unstable ductile tearing.

A proposal for evaluation method of crack growth due to cyclic overload for piping materials based on an elastic-plastic fracture mechanics parameter

International Nuclear Information System (INIS)

Yamaguchi, Yoshihito; Katsuyama, Jinya; Onizawa, Kunio; Li, Yinsheng; Sugino, Hideharu

2011-01-01

The magnitude of Niigata-ken Chuetsu-Oki earthquake in 2007 was beyond the assumed one provided in seismic design. Therefore it becomes an important issue to evaluate the crack growth behaviors due to the cyclic overload like large earthquake. Fatigue crack growth is usually evaluated by Paris's law using the range of stress intensity factor (ΔK). However, ΔK is inappropriate in a loading condition beyond small scale yielding. In this study, the crack growth behaviors for piping materials were investigated based on an elastic-plastic fracture mechanics parameter, J-integral. It was indicated that the crack growth due to the cyclic overload beyond small scale yielding could be the sum of fatigue and ductile crack growth. The retardation effect of excessive loading on the crack growth was observed after the loading. The modified Wheeler model using J-integral has been proposed for the prediction of retardation effect. Finally, an evaluation method for crack growth behaviors due to the cyclic overload is suggested. (author)

Evaluation of elastic-plastic fracture of toughness and fracture resistance of carbon steel STS42

International Nuclear Information System (INIS)

Kobayashi, Hideo; Nakamura, Haruo; Kashiwagi, Kohmei

1987-01-01

The elastic-plastic fracture toughness (J Ic ) and fracture resistance (J-R curve) of a carbon steel, STS42, used for piping in a nuclear reactor were evaluated according to the several evaluating methods recommended or proposed so far, to discuss their applicability and utility. The results obtained are as follows: (1) In evaluating J Ic , the multiple specimen method recommended by the Japan Society for Mechanical Engineers (JSME standard S001) gives the most reliable results by using smaller sized specimens. (2) The single-specimen methods by using the compliance technique, adopted in the ASTM standards (E813, E813 modified, Tentative test procedure for determining the plain strain J-R curve), do not give an accurate J-R curve or J Ic , due to an error in the calculated crack length. (3) In evaluating the J-R curve, it is necessary to account for crack extension in calculating the J-integral. (4) According to the above results, a new standard method for determining the J-R curve including the J Ic test method should be poprosed. (author)

Fatigue and fracture behavior of coiled pipes; Comportamento a fratura e fadiga de tubos bobinados

Energy Technology Data Exchange (ETDEWEB)

Pope, Alexandre M.; Silva, Renato M.C. [PETROBRAS, Rio de Janeiro, RJ (Brazil); Frainer, Vitor J; Tarnowski, Gabriel A.; Strohaecker, Telmo R. [Rio Grande do Sul Univ., Porto Alegre, RS (Brazil)

2005-07-01

The possibility of applying coiled pipes in dynamic situations, such as risers, was investigated through full scale fatigue and fracture tests. A total of sixteen X-65 pipe specimens 41/2'' Od x 0.3'' x 13', containing a bias weld at the mid length, were fatigue tested in a resonance rig at 27 Hz. Six specimens were tested in the as fabricated condition while the ten others were previously subjected to five plastic deformation cycles, simulating reeling operations.. Tests were run until a crack propagated through thickness or 10{sup 7} cycles were achieved. Two cracked specimens were fracture tested in tension. All tests were carried out at room temperature. Test results showed that the fatigue lives were above the Bs-7608 mean class 'B' curves independently whether the specimen were plastically deformed or not. Post fatigue inspection and fractographic examination revealed that fatigue cracks propagate mostly from pits or surface scratches rather than from the bias weld. The two full size tensile specimens failed by ductile fracture at 80% of the material measured strength. (author)

ADIMEW: Fracture assessment and testing of an aged dissimilar metal weld pipe assembly

International Nuclear Information System (INIS)

Wintle, J.B.; Hayes, B.; Goldthorpe, M.R.

2004-01-01

ADIMEW (Assessment of Aged Piping Dissimilar Metal Weld Integrity) was a three-year collaborative research programme carried out under the EC 5th Framework Programme. The objective of the study was to advance the understanding of the behaviour and safety assessment of defects in dissimilar metal welds between pipes representative of those found in nuclear power plant. ADIMEW studied and compared different methods for predicting the behaviour of defects located near the fusion boundaries of dissimilar metal welds typically used to join sections of austenitic and ferritic piping operating at high temperature. Assessment of such defects is complicated by issues that include: severe mis-match of yield strength of the constituent parent and weld metals, strong gradients of material properties, the presence of welding residual stresses and mixed mode loading of the defect. The study includes the measurement of material properties and residual stresses, predictive engineering analysis and validation by means of a large-scale test. The particular component studied was a 453mm diameter pipe that joins a section of type A508 Class 3 ferritic pipe to a section of type 316L austenitic pipe by means of a type 308 austenitic weld with type 308/309L buttering laid on the ferritic pipe. A circumferential, surface-breaking defect was cut using electro discharge machining into the 308L/309L weld buttering layer parallel to the fusion line. The test pipe was subjected to four-point bending to promote ductile tearing of the defect. This paper presents the results of TWI contributions to ADIMEW including: fracture toughness testing, residual stress measurements and assessments of the ADIMEW test using elastic-plastic, cracked body, finite element analysis. (orig.)

International piping integrity research group (IPIRG) program final report

International Nuclear Information System (INIS)

Schmidt, R.; Wilkowski, G.; Scott, P.; Olsen, R.; Marschall, C.; Vieth, P.; Paul, D.

1992-04-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Programme. The IPIRG Programme was an international group programme managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United states. The objective of the programme was to develop data needed to verify engineering methods for assessing the integrity of nuclear power plant piping that contains circumferential defects. The primary focus was an experimental task that investigated the behaviour of circumferentially flawed piping and piping systems to high-rate loading typical of seismic events. To accomplish these objectives a unique pipe loop test facility was designed and constructed. The pipe system was an expansion loop with over 30 m of 406-mm diameter pipe and five long radius elbows. Five experiments on flawed piping were conducted to failure in this facility with dynamic excitation. The report: provides background information on leak-before-break and flaw evaluation procedures in piping; summarizes the technical results of the programme; gives a relatively detailed assessment of the results from the various pipe fracture experiments and complementary analyses; and, summarizes the advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG Program

Investigation of the ductile fracture properties of Type 304 stainless steel plate, welds, and 4-inch pipe

International Nuclear Information System (INIS)

Vassilaros, M.G.; Hays, R.A.; Gudas, J.P.

1985-01-01

J-integral fracture toughness tests were performed on welded 304 stainless steel 2-inch plate and 4-inch diameter pipe. The 2-inch plate was welded using a hot-wire automatic gas tungsten arc process. The tests were performed at 550 0 F, 300 0 F and room temperature. The results of the J-integral tests indicate that the Jsub(Ic) of the base plate ranged from 4400 to 6100 in lbs/in 2 at 550 0 F. The Jsub(Ic) values for the tests performed at 300 0 F and room temperature were beyond the measurement capacity of the specimens and appear to indicate that Jsub(Ic) was greater than 8000 in lb/in 2 . The J-integral tests performed on the weld metal specimens indicate that the Jsub(Ic) values ranged from 930 to 2150 in lbs/in 2 at 550 0 F. The Jsub(Ic) values of the weld metal specimens tested at 300 0 F and room temperature were 2300 and 3000 in lbs/in 2 respectively. One HAZ specimen was tested at 550 0 F and found to have a Jsub(Ic) value of 2980 in lbs/in 2 which indicates that the HAZ is an average of the base metal and weld metal toughness. These test results indicate that there is a significant reduction in the initiation fracture toughness as a result of welding. The second phase of this task dealt with the fracture toughness testing of 4-inch diameter 304 stainless steel pipes containing a gas tungsten arc weld. The pipes were tested at 550 0 F in four point bending. Three tests were performed, two with a through wall flaw growing circumferentially and the third pipe had a part through radial flaw in combination with the circumferential flaw. These tests were performed using unloading compliance and d.c. potential drop crack length estimate methods. The results of these tests indicate that the presence of a complex crack (radial and circumferential) reduces in the initiation toughness and the tearing modulus of the pipe material compared to a pipe with only a circumferentially growing crack. (orig.)

Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

International Nuclear Information System (INIS)

Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

1981-08-01

This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

International Piping Integrity Research Group (IPIRG) Program. Final report

International Nuclear Information System (INIS)

Wilkowski, G.; Schmidt, R.; Scott, P.

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program

International Piping Integrity Research Group (IPIRG) Program. Final report

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.; Schmidt, R.; Scott, P. [and others

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program.

Application of the cracked pipe element to creep crack growth prediction

Energy Technology Data Exchange (ETDEWEB)

Brochard, J.; Charras, T.

1997-04-01

The modification of a computer code for leak before break analysis is very briefly described. The CASTEM2000 code was developed for ductile fracture assessment of piping systems with postulated circumferential through-wall cracks under static or dynamic loading. The modification extends the capabilities of the cracked pipe element to the determination of fracture parameters under creep conditions (C*, {phi}c and {Delta}c). The model has the advantage of evaluating significant secondary effects, such as those from thermal loading.

An advanced revised universal slope method for low cycle fatigue evaluation of elbow piping subjected to in-plane cyclic bending displacement

International Nuclear Information System (INIS)

Urabe, Yoshio

2015-01-01

In order to rationalize the low cycle fatigue evaluation of elbow piping subjected to in-plane cyclic bending displacement, an advanced revised universal slope method is proposed. In the proposed method, the coefficient of the first term of the fatigue life equation which resembles Manson's equation is expressed by parameters of the multi-axial degree, the tensile strength and the fracture strength. Also, the coefficient of the second term is expressed by the multi-axial degree, the fracture ductility and the minimum fracture ductility under the maximum multi-axial degree. Here equivalent strain range is used for the fatigue life estimation. The previously carried out pipe elbow test data were reanalyzed using the proposed method. As the result, the experimentally obtained fatigue lives had considerably good coincidences with the predicted fatigue lives by the proposed method. Application of the proposed method is also discussed. (author)

Structural dynamics and fracture mechanics calculations of the behaviour of a DN 425 test piping system subjected to transient loading by water hammer

International Nuclear Information System (INIS)

Kussmaul, K.; Kobes, E.; Diem, H.; Schrammel, D.; Brosi, S.

1994-01-01

Within the scope of the German HDR safety programme, several tests were carried out to investigate transient pipe loading initiated by a simulated double-ended guillotine break event, and subsequent closure of a feedwater check valve (water hammer, blow-down). Numerical analyses by means of finite element programmes were performed in parallel to the experiments. Using water hammer tests of a DN 425 piping system with predamaged components, the procedure of such analyses will be demonstrated. The results are presented, beginning with structural dynamic calculations of the undamaged piping; followed by coupling of structural dynamics and fracture mechanics computations with simple flaw elements (line spring); and finishing with costly three-dimensional fracture mechanics analyses. A good description of the real piping behaviour can be made by the numerical methods, even in the case of high plastification processes. ((orig.))

A Study of Laser-generated Ultrasound for Evaluation of Thickness Reduction in Piping System

International Nuclear Information System (INIS)

Seo, Kyung Chul; Choi, Sang Woo; Lee, Jun Hyun; Kang, Seung Hyun

2005-01-01

In carbon steel pipes of nuclear power plants, local wall thinning may result from erosion-corrosion or FAC(Flow Accelerated Corrosion) damage. Local wall thinning is one of the major causes for the structural fracture of these pipes. Therefore, assessment of local wall thinning due to corrosion is an important issue in nondestructive evaluation for the integrity of nuclear power plants. In this study, laser-generated ultrasound technique was employed to evaluate local wall thinning due to corrosion. Guided waves were generated in the thermoelastic regime using a Q-switched pulsed Nd:YAG laser with an linear array slit. Time-frequency analysis of ultrasonic waveforms using wavelet transform and FRT(Fast Fourier Transform) allowed the identification of generated guided wave modes by comparison with the theoretical dispersion curves. This study shows some experimental results about optimization of generating laser ultrasound using various linear array slits.

Development of new Z-factors for the evaluation of the circumferential surface crack in nuclear pipes

International Nuclear Information System (INIS)

Choi, Y.H.; Chung, Y.K.; Park, Y.W.; Lee, J.B.

1997-01-01

The purpose of this study is to develop new Z-factors to evaluate the behavior of a circumferential surface crack in nuclear pipe. Z-factor is a load multiplier used in the Z-factor method, which is one of the ASME Code Sec. XI's recommendations for the estimation of a surface crack in nuclear pipe. It has been reported that the load carrying capacities predicted from the current ASME Code Z-factors, are not well in agreement with the experimental results for nuclear pipes with a surface crack. In this study, new Z-factors for ferritic base metal, ferritic submerged arc welding (SAW) weld metal, austenitic base metal, and austenitic SAW weld metal are obtained by use of the surface crack for thin pipe (SC.TNP) method based on GE/EPRI method. The desirability of both the SC.TNP method and the new Z-factors is examined using the results from 48 pipe fracture experiments for nuclear pipes with a circumferential surface crack. The results show that the SC.TNP method is good for describing the circumferential surface crack behavior and the new Z-factors are well in agreement with the measured Z-factors for both ferritic and austenitic pipes. (orig.)

Effect of plastic strain on fracture strength of cracked components

International Nuclear Information System (INIS)

Kamaya, Masayuki

2009-01-01

Nuclear power plant components are occasionally subjected to large load by earthquake and may suffer plastic strain. Although the plastic strain induced in materials increases the strength, it may reduce the fracture toughness due to a crack in the components. In this study, the effect of the plastic strain on strength of cracked components was investigated. Firstly, the change in the tensile properties and fracture toughness due to plastic strain were examined for Type 316 stainless steel and carbon steel (SM490). The degree of nominal plastic strain was 5%, 10%, 20% and 40% (only for stainless steel). Secondly, the J-integral values of surface crack on a pipe were evaluated by finite element analyses. Finally, the critical load for fracture of the cracked pipe was evaluated for various pipe and crack geometries using the J-integral values and the fracture toughness obtained. It was concluded that the plastic strain enhances the fracture strength of the cracked components when the induced plastic strain is less than 10%, although the extremely large plastic strain could reduce the strength. (author)

Effect of plastic strain on fracture strength of cracked components

International Nuclear Information System (INIS)

Kamaya, Masayuki

2010-01-01

Nuclear power plant components are occasionally subjected to excessive load by earthquake and may suffer plastic strain. Although the plastic strain introduced in materials increases the strength, it may reduce the fracture toughness. In this study, the effect of the plastic strain on strength of cracked components was investigated. Firstly, the change in the tensile properties and fracture toughness due to plastic strain were examined for Type 316 stainless steel and carbon steel (SM 490). The degree of nominal plastic strain was 5%, 10%, 20% and 40% (only for stainless steel). Secondly, the J-integral values of surface crack on a pipe were evaluated by finite element analyses. Finally, the critical load for fracture of the cracked pipe was evaluated for various pipe and crack geometries using the J-integral values and the fracture toughness obtained. It was concluded that the plastic strain enhances the fracture strength of the cracked components when the induced plastic strain is less than 10%, although the extremely large plastic strain could reduce the strength. (author)

Structural integrity assessment of piping components

International Nuclear Information System (INIS)

Kushwaha, H.S.; Chattopadhyay, J.

2008-01-01

Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

Investigation on vibrational evaluation criteria for small-bore pipe

International Nuclear Information System (INIS)

Tsuji, Takashi; Maekawa, Akira; Takahashi, Tsuneo; Kato, Minoru; Torigoe, Yuichi

2013-01-01

The well-known organization such as API and SwRI in USA developed criteria for piping vibrational evaluation. These criteria are targeted for main pipes, but not branch pipes with small bore. In this study, applicability of criteria of API and SwRI to branch pipes was investigated. Vibration test using piping system with small bore branch pipe was conducted and amplitudes of vibrational stress and displacement were measured for various exciting force. In comparison of the measurements with the two criteria, though the criteria of API and SwRI were applicable to small bore branch pipe, they made too conservative evaluation. (author)

Fracture dynamics of a propagating crack in a pressurized ductile cylinder

International Nuclear Information System (INIS)

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

1977-01-01

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

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.

Variations of fracture toughness and stress-strain curve of cold worked stainless steel and their influence on failure strength of cracked pipe

International Nuclear Information System (INIS)

Kamaya, Masayuki

2016-01-01

In order to assess failure probability of cracked components, it is important to know the variations of the material properties and their influence on the failure load assessment. In this study, variations of the fracture toughness and stress-strain curve were investigated for cold worked stainless steel. The variations of the 0.2% proof and ultimate strengths obtained using 8 specimens of 20% cold worked stainless steel (CW20) were 77 MPa and 81 MPa, respectively. The respective variations were decreased to 13 and 21 MPa for 40% cold worked material (CW40). Namely, the variation in the tensile strength was decreased by hardening. The COVs (coefficients of variation) of fracture toughness were 7.3% and 16.7% for CW20 and CW40, respectively. Namely, the variation in the fracture toughness was increased by hardening. Then, in order to investigate the influence of the variations in the material properties on failure load of a cracked pipe, flaw assessments were performed for a cracked pipe subjected to a global bending load. Using the obtained material properties led to variation in the failure load. The variation in the failure load of the cracked pipe caused by the variation in the stress-strain curve was less than 1.5% for the COV. The variation in the failure load caused by fracture toughness variation was relatively large for CW40, although it was less than 2.0% for the maximum case. It was concluded that the hardening induced by cold working does not cause significant variation in the failure load of cracked stainless steel pipe. (author)

Application of the cracked pipe element to creep crack growth prediction

Energy Technology Data Exchange (ETDEWEB)

Brochard, J.; Charras, T. [C.E.A.-C.E.-Saclay DRN/DMT, Gif Sur Yvette (France); Ghoudi, M. [C.E.A.-C.E.-Saclay, Gif Sur Yvette (France)

1997-04-01

Modifications to a computer code for ductile fracture assessment of piping systems with postulated circumferential through-wall cracks under static or dynamic loading are very briefly described. The modifications extend the capabilities of the CASTEM2000 code to the determination of fracture parameters under creep conditions. The main advantage of the approach is that thermal loads can be evaluated as secondary stresses. The code is applicable to piping systems for which crack propagation predictions differ significantly depending on whether thermal stresses are considered as primary or secondary stresses.

The Second International Piping Integrity Research Group (IPIRG-2) program. Final report, October 1991--April 1996

International Nuclear Information System (INIS)

Hopper, A.; Wilowski, G.; Scott, P.; Olson, R.

1997-03-01

The IPIRG-2 program was an international group program managed by the US NRC and funded by organizations from 15 nations. The emphasis of the IPIRG-2 program was the development of data to verify fracture analyses for cracked pipes and fittings subjected to dynamic/cyclic load histories typical of seismic events. The scope included: (1) the study of more complex dynamic/cyclic load histories, i.e., multi-frequency, variable amplitude, simulated seismic excitations, than those considered in the IPIRG-1 program, (2) crack sizes more typical of those considered in Leak-Before-Break (LBB) and in-service flaw evaluations, (3) through-wall-cracked pipe experiments which can be used to validate LBB-type fracture analyses, (4) cracks in and around pipe fittings, such as elbows, and (5) laboratory specimen and separate effect pipe experiments to provide better insight into the effects of dynamic and cyclic load histories. Also undertaken were an uncertainty analysis to identify the issues most important for LBB or in-service flaw evaluations, updating computer codes and databases, the development and conduct of a series of round-robin analyses, and analyst's group meetings to provide a forum for nuclear piping experts from around the world to exchange information on the subject of pipe fracture technology. 17 refs., 104 figs., 41 tabs

Ductile failure analysis of defective API X65 pipes based on stress-modified fracture strain criterion

International Nuclear Information System (INIS)

Oh, Chang Kyun; Kim, Yun Jae; Baek, Jong Hyun; Kim, Young Pyo; Kim, Woo Sik

2006-01-01

A local failure criterion for the API X65 steel is applied to predict ductile failure of full-scale API X65 pipes with simulated corrosion and gouge defects under internal pressure. The local failure criterion is the stress-modified fracture strain for the API X65 steel as a function of the stress triaxiality (defined by the ratio of the hydrostatic stress to the effective stress). Based on detailed FE analyses with the proposed local failure criteria, burst pressures of defective pipes are estimated and compared with experimental data. The predicted burst pressures are in good agreement with experimental data. Noting that an assessment equation against the gouge defect is not yet available, parametric study is performed, from which a simple equation is proposed to predict burst pressure for API X65 pipes with gouge defects

Progress on the degraded piping program - Phase II. Battelle Columbus Division

International Nuclear Information System (INIS)

Wilkowski, Gery; Ahmad, J.; Barnes, C.; Brust, F.; Guerrieri, D.; Kramer, G.; Landow, M.; Marschall, C.; Nakagaki, M.; Papaspyropoulos; Scott, P.

1988-01-01

The overall objective of the Degraded Piping Program is to verify and improve simple estimation schemes to predict the fracture behavior of circumferentially cracked pipe. The program is limited to quasi-static fracture and cracks in straight pipe. There are a variety of materials, flaw geometries, pipe sizes, and loading conditions evaluated. The Degraded Piping Program,which has been extended for one more year, will supply results that provide a basis for regulatory decisions regard applications for leak-before-break (LBB) and In-service flaw assessment. The significance of our results are summarized relative to how they may affect regulatory technical needs. The scope of the work in The Degraded Piping Program includes both analytical and experimental efforts. The experimental efforts have concentrated on testing circumferentially cracked pipe at 550 F (288 C) under si-static loading. Many of the tasks within this program were undertaken with the objective of determining if any detailed efforts were needed. This is true for both the analytical and experimental efforts. i e of the tasks have been slightly expanded during the course of the gram, while others were found to be of lesser concern and further efforts in those areas were not pursued. The results of this summary include the efforts of the third year. These efforts have contributed considerably to the understanding of the application of elastic-plastic fracture mechanics to nuclear piping systems. Rather than listing the significant technical contributions, these contributions are summarized below in relation to their application to LBB analyses, in-service flaw assessment criteria, and (3) material characterization and unusual behavior of nuclear piping materials at light water reactor (LWR) temperatures

Seismic evaluation of piping systems using screening criteria

International Nuclear Information System (INIS)

Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

1994-01-01

This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

A fracture mechanics assessment of surface cracks existing in protective layers of multi-layer composite pipes

Czech Academy of Sciences Publication Activity Database

Hutař, Pavel; Náhlík, Luboš; Šestáková, Lucie; Ševčík, Martin; Knésl, Zdeněk; Nezbedová, E.

2010-01-01

Roč. 92, č. 5 (2010), s. 1120-1125 ISSN 0263-8223 R&D Projects: GA ČR GA106/09/0279 Institutional research plan: CEZ:AV0Z20410507 Keywords : Protective layers * Multi-layer pipes damage * Fracture mechanics * Bi-material interface * Generalized stress intensity factor Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 2.028, year: 2010

Characterisation of girth pipe weld for primary heat transport system of pressurised heavy water reactors

International Nuclear Information System (INIS)

Singh, P.K.; Vaze, K.K.; Kushwaha, H.S.

2002-01-01

The weld and heat affected zone (HAZ) associated with the girth weld are most vulnerable regions of the piping system. The different regions of the weld joint such as the weld metal, HAZ and base metal lead to heterogeneous mechanical and metallurgical properties of the joints. Due to their different metallurgical and mechanical properties, the amounts of damage produced in these regions are different when the component is subjected to service condition. Thus, it is imperative to know the characteristics of these regions of a pipe weld in order to identify the weakest zone for safe designing of high energy piping components. In view of this necessity the present study has been planned to carry out complete characterisation of the weld joint of SA 333 Gr.6 steel pipe, in terms of its metallurgical, mechanical and fracture properties. The mechanical and fracture mechanics properties of the base metal, weld deposit and HAZ have been compared and correlated with reference to their microstructures. Weld joints of SA 333 Gr.6 steel pipe have been prepared by using GTAW root pass and SMAW filling of V-grove as per recommended welding procedure specifications (WPS) conforming to ASME Sec IX commonly used to fabricate nuclear piping system components. The emphasis of the study is to characterise base, weld and HAZ of the pipe weld in terms of chemical, metallurgical, mechanical and fracture mechanics properties. The fracture toughness behaviour of the welds and HAZ has been characterised by J-integral parameters. The fatigue crack growth rate has been characterised by Paris Law. Stretched zone width (SZW) has been measured under SEM to evaluate initiation fracture toughness. The estimated initiation fracture toughness based on SZW and blunting line given by EGF recommendation have been compared. The fracture mechanics properties of base, weld and HAZ has been determined and compared. The fracture mechanics properties of the weld and HAZ have been correlated to their

Evaluation of LBB margin of nuclear piping systems

International Nuclear Information System (INIS)

Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun; Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok

1999-04-01

Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material

Evaluation of LBB margin of nuclear piping systems

Energy Technology Data Exchange (ETDEWEB)

Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun [Seoul Nationl Univ., Seoul (Korea, Republic of); Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-04-15

Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material.

Long-Term Strength of a Thick-Walled Pipe Filled with an Aggressive Medium, with Account for Damageability of the Pipe Material and Residual Strength

Science.gov (United States)

Piriev, S. A.

2018-01-01

This paper describes the study of scattered fracture of a thick-walled pipe filled with an aggressive medium, which creates uniform pressure on the inner surface of the pipe. It is assumed that the aggressive medium affects only the value of instantaneous strength. Damageability is described by an integral operator of the hereditary type. The problem is solved with allowance for residual strength of the pipe material behind the fracture front. Numerical calculation is carried out, and relationships between the fracture front coordinate and time for various concentrations of the aggressive medium and residual strength behind the fracture front are constructed.

Efficient improvement of nuclear power plant safety by reorganization of risk-informed safety importance evaluation methods for piping welded portions

Energy Technology Data Exchange (ETDEWEB)

Irie, Takashi; Hanafusa, Hidemitsu; Suyama, Takeshi [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Morota, Hidetsugu; Kojima, Sigeo; Mizuno, Yoshinobu [Computer Software Development Co., Ltd., Tokyo (Japan)

2002-09-01

In this work, risk information was used to evaluate the safety importance of piping welded portions which were important for plant operation and maintenance of nuclear power plants. There are two types of risk-informed safety importance evaluation methods, namely the ASME method and the EPRI method. Since both methods have advantages and disadvantages, elements of each method were combined and reorganized. Considerations included whether the degradation mechanisms would be objectively evaluated and whether plant safety would be efficiently improved. The most objective and efficient method was as follows. Piping failure potential is quantitatively and objectively evaluated for failure with probabilistic fracture mechanics (PFM) and for other degradation mechanisms with empirical failure rates, and conditional core damage probability (CCDP) is calculated with PSA. This method reduces the inspected segment numbers to 1/4 of the deterministic method and increases the ratio of risk, which is covered by the inspected segments, to total risk from 80% of the deterministic method to 95%. Piping inspection numbers decreased for safety injection systems that were required the inspections by the deterministic method. Piping inspections were required for part of main feed water and main steam systems that were not required the inspections by the deterministic method. (author)

Review of Acceleration Methods for Seismic Analysis of Through-Wall Cracked Piping from the Viewpoint of Linear Elastic Fracture Mechanics

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Sung; Kim, Yong Woo [Sunchon National University, Suncheon (Korea, Republic of)

2014-10-15

Two acceleration methods, an effective force method (or inertia method) and a large mass method, have been applied for performing time history seismic analysis. The acceleration methods for uncracked structures have been verified via previous studies. However, no study has identified the validity of these acceleration methods for cracked piping. In this study, the validity of the acceleration methods for through-wall cracked piping is assessed via time history implicit dynamic elastic seismic analysis from the viewpoint of linear elastic fracture mechanics. As a result, it is identified that both acceleration methods show the same results for cracked piping if a large mass magnitude and maximum time increment are adequately selected.

Review of Acceleration Methods for Seismic Analysis of Through-Wall Cracked Piping from the Viewpoint of Linear Elastic Fracture Mechanics

International Nuclear Information System (INIS)

Kim, Jong Sung; Kim, Yong Woo

2014-01-01

Two acceleration methods, an effective force method (or inertia method) and a large mass method, have been applied for performing time history seismic analysis. The acceleration methods for uncracked structures have been verified via previous studies. However, no study has identified the validity of these acceleration methods for cracked piping. In this study, the validity of the acceleration methods for through-wall cracked piping is assessed via time history implicit dynamic elastic seismic analysis from the viewpoint of linear elastic fracture mechanics. As a result, it is identified that both acceleration methods show the same results for cracked piping if a large mass magnitude and maximum time increment are adequately selected

Effects of toughness anisotropy and combined tension, torsion, and bending loads on fracture behavior of ferritic nuclear pipe

Energy Technology Data Exchange (ETDEWEB)

Mohan, R.; Marshall, C.; Ghadiali, N.; Wilkowski, G. [Battelle, Columbus, OH (United States)

1997-04-01

This paper summarizes work on angled through-wall-crack initiation and combined loading effects on ferritic nuclear pipe performed as part of the Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks In Piping an Piping Welds{close_quotes}. The reader is referred to Reference 1 for details of the experiments and analyses conducted as part of this program. The major impetus for this work stemmed from the observation that initially circumferentially oriented cracks in carbon steel pipes exhibited a high tendency to grow at a different angle when the cracked pipes were subjected to bending or bending plus pressure loads. This failure mode was little understood, and the effect of angled crack grown from an initially circumferential crack raised questions about how cracks in a piping system subjected to combined loading with torsional stresses would behave. There were three major efforts undertaken in this study. The first involved a literature review to assess the causes of toughness anisotropy in ferritic pipes and to develop strength and toughness data as a function of angle from the circumferential plane. The second effort was an attempt to develop a screening criterion based on toughness anisotropy and to compare this screening criterion with experimental pipe fracture data. The third and more significant effort involved finite element analyses to examine why cracks grow at an angle and what is the effect of combined loads with torsional stresses on a circumferentially cracked pipe. These three efforts are summarized.

Plastic collapse moment for pipe repaired with weld overlay

International Nuclear Information System (INIS)

Li, Yinsheng; Hasegawa, Kunio; Shibuya, Akira; Deardorff, Arthur

2009-01-01

The Weld Overlay has been used in several countries as an effective method to repair the stress corrosion cracks in nuclear power plant piping. However, the method to evaluate the plastic collapse stress for the pipe repaired with Weld Overlay has not been proposed and the limit load criterion for single uniform material has been used to design its structure by now. In this paper, the equations to evaluate the plastic collapse moment for the pipe repaired with Weld Overlay have been derived considering two layer materials. Moreover, several numerical examples are given to show the validity of Weld Overlay. The equations given in this paper are simple to use like the limit load criterion showed in present standards such as JSME Rules on Fitness-for-Service for Nuclear Power Plants or ASME Boiler and Pressure Vessel Code Section XI, and they can not only be used to evaluate the fracture of the pipe, but also be applied to design the weld structure. (author)

Development of modified piping evaluation diagram for LBB application to Korean next generation reactor

International Nuclear Information System (INIS)

Huh, Nam Su; Kim, Young Jin; Pyo, Chang ryul; Yu, Young Jun; Yang, Jun Seog

1999-01-01

Recently, the Piping Evaluation Diagram (PED) is accepted in nuclear industry for simple application of Leak-Before-Break (LBB) concept to piping system. By utilizing the PED, the LBB concept is applied before the piping layout is finalized. However, the developed PED may have to be modified to account for the difference between the material properties of the PED development stage and those of the assembly stage. The objective of this paper is to develop the modified PED which can account for the variation of material properties. For this purpose, a parametric study was performed to investigate the effect of stress-strain curve on the detectable crack length and the effect of fracture resistance curve on the LBB allowable load. Finite element analyses were also performed to investigate the effect of stress-strain curve on the LBB allowable load. Finally a modified PED is developed as a function of crack length (DLC and 2xDLC) and the allowable Safe shutdown Earthquake (SSE) load. By adopting the modified PED, the variation of material properties can be considered in the LBB analysis and the computer runs required for the LBB analysis can be considerably reduced

Evaluation of pipe weld NDE indications

International Nuclear Information System (INIS)

Brasse, M.

2007-01-01

This paper discusses the evaluation of non-destructive examination (NDE) indications in pipe welds. The evaluation procedure is described in a practical engineer's view and examples are also given. (author)

Evaluation of flaws in ferritic piping: ASME Code Appendix J, Deformation Plasticity Failure Assessment Diagram (DPFAD)

International Nuclear Information System (INIS)

Bloom, J.M.

1991-08-01

This report summarizes the methods and bases used by an ASME Code procedure for the evaluation of flaws in ferritic piping. The procedure is currently under consideration by the ASME Boiler and Pressure Vessel Code Committee of Section 11. The procedure was initially proposed in 1985 for the evaluation of the acceptability of flaws detected in piping during in-service inspection for certain materials, identified in Article IWB-3640 of the ASME Boiler and Pressure Vessel Code Section 11 ''Rules for In-service Inspection of Nuclear Power Plant Components.'' for which the fracture toughness is not sufficiently high to justify acceptance based solely on the plastic limit load evaluation methodology of Appendix C and IWB-3641. The procedure, referred to as Appendix J, originally included two approaches: a J-integral based tearing instability (J-T) analysis and the deformation plasticity failure assessment diagram (DPFAD) methodology. In Appendix J, a general DPFAD approach was simplified for application to part-through wall flows in ferritic piping through the use of a single DPFAD curve for circumferential flaws. Axial flaws are handled using two DPFAD curves where the ratio of flaw depth to wall thickness is used to determine the appropriate DPFAD curve. Flaws are evaluated in Appendix J by comparing the actual pipe applied stress with the allowable stress with the appropriate safety factors for the flaw size at the end of the evaluation period. Assessment points for circumferential and axial flaws are plotted on the appropriate failure assessment diagram. In addition, this report summarizes the experimental test predictions of the results of the Battelle Columbus Laboratory experiments, the Eiber experiments, and the JAERI tests using the Appendix J DPFAD methodology. Lastly, this report also provides guidelines for handling residual stresses in the evaluation procedure. 22 refs., 13 figs., 5 tabs

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

Probabilistic fracture mechanics applied for lbb case study: international benchmark

International Nuclear Information System (INIS)

Radu, V.

2015-01-01

An application of probabilistic fracture mechanics to evaluate the structural integrity for a case study chosen from experimental Mock-ups of FP7 STYLE project is described. The reliability model for probabilistic structural integrity, focused on the assessment of TWC in the pipe weld under complex loading (bending moment and residual stress) has been setup. The basic model is the model of fracture for through-wall cracked pipe under elastic-plastic conditions. The corresponding structural reliability approach is developed with the probabilities of failure associated with maximum load for crack initiation, net-section collapse but also the evaluation the instability loads. The probabilities of failure for a through-wall crack in a pipe subject to pure bending are evaluated by using crude Monte Carlo simulations. The results from the international benchmark are presented for the mentioned case in the context of ageing and lifetime management of pressure boundary/pressure circuit component. (authors)

Mis-Match Limit Load Analyses and Fracture Mechanics Assessment for Welded Pipe with Circumferential Crack at the Center of Weldment

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Jeon, Jun Young; Shim, Kwang Bo; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Kim, Jong Sung [Sunchon University, Suncheon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company, Daejeon (Korea, Republic of)

2010-01-15

In this paper, limit load analyses and fracture mechanics analyses were conducted via finite element analyses for the welded pipe with circumferential crack at the center of the weldment. Systematic changes for strength mismatch ratio, width of weldment, crack shape and thickness ratio of the pipe were considered to provide strength mismatch limit load. And J-integral calculations based on reference stress method were conducted for two materials, stainless steel and ferritic steel. Reference stress defined by provided strength mis-match limit load gives much more accurate J-integral.

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.)

BWR pipe crack remedies evaluation

International Nuclear Information System (INIS)

Shack, W.J.; Kassner, T.F.; Maiya, P.S.; Park, J.Y.; Ruther, W.; Kuzay, T.; Rybicki, E.F.; Stonesifer, R.B.

1988-01-01

Piping in light-water-reactor power systems has been affected by several types of environmental degradation. This paper presents results from studies of (1) stress corrosion crack growth in fracture mechanics specimens of modified Type 347 SS and Type 304/308L SS weld overlay material, (2) heat-to-heat variations in stress corrosion cracking (SCC) of Types 316NG and 347 SS, (3) SCC of sensitized Type 304 SS in water with cupric ion or organic acid impurities, (4) electrochemical potential (ECP) measurements under gamma irradiation, (5) SCC of ferritic steels, (6) strain-controlled fatigue of Type 316NG SS in air at ambient temperature, and (7) through-wall residual stress measurements and finite-element calculation of residual stresses in weldments treated by a mechanical stress improvement process (MSIP). Fracture-mechanics crack-growth-rate tests on Type 316NG SS have shown that transgranular cracking can occur even in high purity environments, whereas no crack growth was observed in Type 347 SS even in impurity environments. In tests on weld overlay specimens, no cracks penetrated into the overlay even in impurity environments. Instead, the cracks branched when they approached the overlay, and then grew parallel to interface. In SCC tests on sensitized Type 304 SS, cupric ions at concentrations greater than ∼1 ppm were found to be deleterious, whereas organic acids at this concentration were not detrimental. Tests on several ferritic steels indicate a strong correlation between the sulfur content of the steels and susceptibility to SCC. External gamma radiation fields produced a large positive shift in the ECP of Type 304 SS at low dissolved-oxygen concentrations (<5 ppb), whereas in the absence of an external gamma field there was no difference in the ECP values of irradiated and nonirradiated material. Fatigue data for Type 316NG SS are consistent with the ASME code mean curve at high strains, but fall below the curve at low strains. Calculations of the

Environmental Assisted Fatigue Evaluation of Direct Vessel Injection Piping Considering Thermal Stratification

International Nuclear Information System (INIS)

Kim, Taesoon; Lee, Dohwan

2016-01-01

As the environmentally assisted fatigue (EAF) due to the primary water conditions is to be a critical issue, the fatigue evaluation for the components and pipes exposed to light water reactor coolant conditions has become increasingly important. Therefore, many studies to evaluate the fatigue life of the components and pipes in LWR coolant environments on fatigue life of materials have been conducted. Among many components and pipes of nuclear power plants, the direct vessel injection piping is known to one of the most vulnerable pipe systems because of thermal stratification occurred in that systems. Thermal stratification occurs because the density of water changes significantly with temperature. In this study, fatigue analysis for DVI piping using finite element analysis has been conducted and those results showed that the results met design conditions related with the environmental fatigue evaluation of safety class 1 pipes in nuclear power plants. Structural and fatigue integrity for the DVI piping system that thermal stratification occurred during the plant operation has conducted. First of all, thermal distribution of the piping system is calculated by computational fluid dynamic analysis to analyze the structural integrity of that piping system. And the fatigue life evaluation considering environmental effects was carried out. Our results showed that the DVI piping system had enough structural integrity and fatigue life during the design lifetime of 60 years

IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

Energy Technology Data Exchange (ETDEWEB)

Scott, P.; Olson, R.; Marschall, C.; Rudland, D. [and others

1997-02-01

This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program.

IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

International Nuclear Information System (INIS)

Scott, P.; Olson, R.; Marschall, C.; Rudland, D.

1997-02-01

This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program

Leak-before-break behaviour of nuclear piping systems

International Nuclear Information System (INIS)

Bartholome, G.; Wellein, R.

1992-01-01

The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

Analysis of Fracture Behaviour of Multilayer Pipes

Czech Academy of Sciences Publication Activity Database

Nezbedová, E.; Knésl, Zdeněk; Vlach, B.

2007-01-01

Roč. 36, č. 5 (2007), s. 207-212 ISSN 1465-8011. [Plastic Pipes /13./. Washington, D. C., 02.10.2006-05.10.2006] R&D Projects: GA ČR GA106/07/1284 Institutional research plan: CEZ:AV0Z20410507 Keywords : multi-layer pipes Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 0.431, year: 2007

Structural integrity evaluation of FTL in-pool piping

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y

1998-05-01

HANARO fuel test loop will be equipped in HANARO to obtain the development betterment of advanced fuel and materials through the irradiation test. The object of this study is to evaluate the structural integrity of FTL in-pool piping by investigating a dynamic analysis of the loop containing a postulated rupture section. The method to perform the dynamic analysis and structural integrity evaluation caused by the pipe whip in water environment can be a reference for a similar structural integrity evaluation. (author). 7 refs., 39 tabs., 34 figs.

Determination of leakage areas in nuclear piping

Energy Technology Data Exchange (ETDEWEB)

Keim, E. [Siemens/KWU, Erlangen (Germany)

1997-04-01

For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack.

Determination of leakage areas in nuclear piping

International Nuclear Information System (INIS)

Keim, E.

1997-01-01

For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack

Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

International Nuclear Information System (INIS)

Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

1995-01-01

This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

Development of piping evaluation diagram for LBB application to KNGR surge line

International Nuclear Information System (INIS)

Yoon, K. S.; Park, W. B.; Kim, J. M.; Choi, T. S.; Yang, J. S.; Park, C. Y.

1998-01-01

Plant specific data, such as pipe geometry, material properties and pipe loads, are required in order to evaluate Leak-Before-Break (LBB) applicability to piping systems in nuclear power plant under the construction. However, the existing method of LBB evaluation for KSNP's can not be used for newly developed nuclear plants such as Korean Next Generation Reactor (KNGR) which material properties is not available and LBB evaluation is required during design process. In order to solve this problem during developing process for KNGR surge line LBB Piping Evaluation Diagram (PED), which is independent of piping geometry and has a function of the loads applied in piping system, is developed in this paper. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the PED. The PED, therefore, can be used for quick LBB evaluation of KNGR surge line in the process of both design and construction. The benefit obtained by using the PED is : 1) to be able to very quickly confirm LBB applicability without calculating any leakage crack length for all concerned piping locations in the process of both iterative design for optimal routing and construction and 2) to save significantly a lot of computing times required for the corresponding LBB analyses

Setting reinspection intervals for seam welded piping by use of probabilistic fracture mechanics and target reliability values

International Nuclear Information System (INIS)

Harris, D.O.; Dedhia, D.

1995-01-01

The purpose of this paper is to describe a procedure for the selection of a reinspection interval for defects found during an inspection. The procedure is based on probabilistic fracture mechanics calculations of the reliability of the component into the future and selection of an inspection time based on maintaining the target value reliability. The selection of a target value based on the risk of everyday activities is discussed. The procedure is applied to high temperature seam welded piping as an example, because the probabilistic fracture mechanics tools are relatively readily available and this is a problem of great current interest. The results obtained in the example problem indicate reinspection intervals much shorter than field experience would suggest. This indicates a conservatism in the fracture mechanics procedures and/or lack of accurate characterization of scatter in material properties due to lack of data. The general procedure should prove useful in the disposition of detected cracks in a wide variety of situations

Fatigue test results of straight pipe with flaws in inner surface

International Nuclear Information System (INIS)

Shibata, Katsuyuki; Oba, Toshihiro; Kawamura, Takaichi; Yokoyama, Norio; Miyazono, Shohachiro

1981-01-01

Fatigue and fracture tests of piping models with flaws in the inner surface were carried out to investigate the fatigue crack growth, coalescence of multiple cracks and fracture behavior. Two straight test pipes with and without weldment in the test section of SUS304L stainless steel were tested under almost the same test conditions. Three artificial defects were machined in the inner surface of the test section of the test pipes. The fatigue test were performed untill the cracks coalesced and grew through the thickness. Subsequently, a static load was imposed on test pipe which contained a large crack in the test section. The test results show that the fatigue crack growth is slower than that predicted by the method specified in the Section XI of ASME Boiler and Pressure Vessel Code, and that the test pipes can endure more than the static load of 3Sm without an unstable fracture. (author)

Piping research program plan

International Nuclear Information System (INIS)

1988-09-01

This document presents the piping research program plan for the Structural and Seismic Engineering Branch and the Materials Engineering Branch of the Division of Engineering, Office of Nuclear Regulatory Research. The plan describes the research to be performed in the areas of piping design criteria, environmentally assisted cracking, pipe fracture, and leak detection and leak rate estimation. The piping research program addresses the regulatory issues regarding piping design and piping integrity facing the NRC today and in the foreseeable future. The plan discusses the regulatory issues and needs for the research, the objectives, key aspects, and schedule for each research project, or group of projects focussing of a specific topic, and, finally, the integration of the research areas into the regulatory process is described. The plan presents a snap-shot of the piping research program as it exists today. However, the program plan will change as the regulatory issues and needs change. Consequently, this document will be revised on a bi-annual basis to reflect the changes in the piping research program. (author)

Stability of cracked pipe under inertial stresses. Subtask 1.1 final report

International Nuclear Information System (INIS)

Scott, P.; Wilson, M.; Olson, R.; Marschall, C.; Schmidt, R.; Wilkowski, G.

1994-08-01

This report presents the results of the pipe fracture experiments, analyses, and material characterization efforts performed within Subtask 1.1 of the IPIRG Program. The objective of Subtask 1.1 was to experimentally verify the analysis methodologies for circumferentially cracked pipe subjected primarily to inertial stresses. Eight cracked-pipe experiments were conducted on 6-inch nominal diameter TP304 and A106B pipe. The experimental procedure was developed using nonlinear time-history finite element analyses which included the nonlinear behavior due to the crack. The model did an excellent job of predicting the displacements, forces, and times to maximum moment. The comparison of the experimental loads to the predicted loads by the Net-Section-Collapse (NSC), Dimensionless Plastic-Zone Parameter, J-estimation schemes, R6, and ASME Section XI in-service flaw assessment criteria tended to underpredict the measured bending moments except for the NSC analysis of the A106B pipe. The effects of flaw geometry and loading history on toughness were evaluated by calculating the toughness from the pipe tests and comparing these results to C(l) values. These effects were found to be variable. The surface-crack geometry tended to increase the toughness (relative to CM results), whereas a negative load-ratio significantly decreased the TP304 stainless steel surface-cracked pipe apparent toughness. The inertial experiments tended to achieve complete failure within a few cycles after reaching maximum load in these relatively small diameter pipe experiments. Hence, a load-controlled fracture mechanics analysis may be more appropriate than a displacement-controlled analysis for these tests

The stress analysis evaluation and pipe support layout for pressurizer discharge system

International Nuclear Information System (INIS)

Mao Qing; Wang Wei; Zhang Yixiong

2000-01-01

The author presents the stress analysis and evaluation of pipe layout and support adjustment process for Qinshan phase II pressurizer discharge system. Using PDL-SYSPIPE INTERFACE software, the characteristic parameters of the system are gained from 3-D CAD engineering design software PDL and outputted as the input date file format of special pipe stress analysis program SYSPIPE. Based on that, SYSPIPE program fast stress analysis function is applied in adjusting pipe layout , support layout and support types. According to RCC-M standard, the pipe stress analysis and evaluation under deadweight, internal pressure, thermal expansion, seismic, pipe rupture and discharge loads are fulfilled

Assessment of short through-wall circumferential cracks in pipes. Experiments and analysis: March 1990--December 1994

Energy Technology Data Exchange (ETDEWEB)

Brust, F.W.; Scott, P.; Rahman, S. [Battelle, Columbus, OH (United States)] [and others

1995-04-01

This topical report summarizes the work performed for the Nuclear Regulatory Commission`s (NRC) research program entitled ``Short Cracks in Piping and Piping Welds`` that specifically focuses on pipes with short through-wall cracks. Previous NRC efforts, conducted under the Degraded Piping Program, focused on understanding the fracture behavior of larger cracks in piping and fundamental fracture mechanics developments necessary for this technology. This report gives details on: (1) material property determinations, (2) pipe fracture experiments, and (3) development, modification, and validation of fracture analysis methods. The material property data required to analyze the experimental results are included. These data were also implemented into the NRC`s PIFRAC database. Three pipe experiments with short through-wall cracks were conducted on large diameter pipe. Also, experiments were conducted on a large-diameter uncracked pipe and a pipe with a moderate-size through-wall crack. The analysis results reported here focus on simple predictive methods based on the J-Tearing theory as well as limit-load and ASME Section 11 analyses. Some of these methods were improved for short-crack-length predictions. The accuracy of the various methods was determined by comparisons with experimental results from this and other programs. 69 refs., 124 figs, 49 tabs.

Assessment of short through-wall circumferential cracks in pipes. Experiments and analysis: March 1990--December 1994

International Nuclear Information System (INIS)

Brust, F.W.; Scott, P.; Rahman, S.

1995-04-01

This topical report summarizes the work performed for the Nuclear Regulatory Commission's (NRC) research program entitled ''Short Cracks in Piping and Piping Welds'' that specifically focuses on pipes with short through-wall cracks. Previous NRC efforts, conducted under the Degraded Piping Program, focused on understanding the fracture behavior of larger cracks in piping and fundamental fracture mechanics developments necessary for this technology. This report gives details on: (1) material property determinations, (2) pipe fracture experiments, and (3) development, modification, and validation of fracture analysis methods. The material property data required to analyze the experimental results are included. These data were also implemented into the NRC's PIFRAC database. Three pipe experiments with short through-wall cracks were conducted on large diameter pipe. Also, experiments were conducted on a large-diameter uncracked pipe and a pipe with a moderate-size through-wall crack. The analysis results reported here focus on simple predictive methods based on the J-Tearing theory as well as limit-load and ASME Section 11 analyses. Some of these methods were improved for short-crack-length predictions. The accuracy of the various methods was determined by comparisons with experimental results from this and other programs. 69 refs., 124 figs, 49 tabs

Development of engineering program for integrity evaluation of pipes with local wall thinned defects

International Nuclear Information System (INIS)

Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong; Park, Sang Kyu

2008-01-01

Integrity evaluation of pipes with local wall thinning by erosion and corrosion is increasingly important in maintenance of wall thinned carbon steel pipes in nuclear power plants. Though a few program for integrity assessment of wall thinned pipes have been developed in domestic nuclear field, however those are limited to straight pipes and methodology proposed in ASME Sec.XI Code Case N-597. Recently, the engineering program for integrity evaluation of pipes with all kinds of local wall defects such as straight, elbow, reducer and branch pipes was developed successfully. The program was designated as PiTEP (Pipe Thinning Evaluation Program), which name was registered as a trademark in the Korea Intellectual Property Office. A developed program is carried out by sequential step of four integrity evaluation methodologies, which are composed of construction code, code case N-597, its engineering method and two developed owner evaluation method. As PiTEP program will be performed through GUI (Graphic User Interface) with user's familiarity, it would be conveniently used by plant engineers with only measured thickness data, basic operation conditions and pipe data

Evaluation of underground pipe-structure interface for surface impact load

Energy Technology Data Exchange (ETDEWEB)

Wang, Shen, E-mail: swang@terrapower.com

2017-06-15

Highlights: • A simple method is proposed for the evaluation of underground pipelines for surface impact load considering the effect of a nearby pipe-structure interface. • The proposed simple method can be used to evaluate the magnitude of damage within a short period of time after accidental drop occurs. • The proposed method is applied in a practical example and compared by using finite element analysis. - Abstract: Nuclear safety related buried pipelines need to be assessed for the effects of postulated surface impact loads. In published solutions, the buried pipe is often considered within an elastic half space without interference with other underground structures. In the case that a surface impact occurs in short distance from an underground pipe-structure interface, this boundary condition will further complicate the buried pipe evaluation. Neglecting such boundary effect in the assessment may lead to underestimating potential damage of buried pipeline, and jeopardizing safety of the nuclear power plant. Comprehensive analysis of such structure-pipe-soil system is often subjected to availability of state-of-art finite element tools, as well as costly and time consuming. Simple, but practical conservative techniques have not been established. In this study, a mechanics based solution is proposed in order to assess the magnitude of damage to a buried pipeline beneath a heavy surface impact considering the effect of a nearby pipe-structure interface. The proposed approach provides an easy to use tool in the early stage of evaluation before the decision of applying more costly technique can be made by owner of the nuclear facility.

Evaluation of underground pipe-structure interface for surface impact load

International Nuclear Information System (INIS)

Wang, Shen

2017-01-01

Highlights: • A simple method is proposed for the evaluation of underground pipelines for surface impact load considering the effect of a nearby pipe-structure interface. • The proposed simple method can be used to evaluate the magnitude of damage within a short period of time after accidental drop occurs. • The proposed method is applied in a practical example and compared by using finite element analysis. - Abstract: Nuclear safety related buried pipelines need to be assessed for the effects of postulated surface impact loads. In published solutions, the buried pipe is often considered within an elastic half space without interference with other underground structures. In the case that a surface impact occurs in short distance from an underground pipe-structure interface, this boundary condition will further complicate the buried pipe evaluation. Neglecting such boundary effect in the assessment may lead to underestimating potential damage of buried pipeline, and jeopardizing safety of the nuclear power plant. Comprehensive analysis of such structure-pipe-soil system is often subjected to availability of state-of-art finite element tools, as well as costly and time consuming. Simple, but practical conservative techniques have not been established. In this study, a mechanics based solution is proposed in order to assess the magnitude of damage to a buried pipeline beneath a heavy surface impact considering the effect of a nearby pipe-structure interface. The proposed approach provides an easy to use tool in the early stage of evaluation before the decision of applying more costly technique can be made by owner of the nuclear facility.

Corrosion evaluation in insulated pipes by non destructive testing method

International Nuclear Information System (INIS)

Abd Razak Hamzah; Azali Muhammad; Mohammad Pauzi Ismail; Abd Nassir Ibrahim; Abd Aziz Mohamed; Sufian Saad; Saharuddin Sayuti; Shukri Ahmad

2002-01-01

In engineering plants, detection of corrosion and evaluation of deposit in insulated pipes using radiography method are considered as a very challenging tasks. In General this degradation problem is attributed to water condensation. It causes the formation of deposit and scale inside the pipe, as well as between the insulation and pipe in cold temperature pipes. On the other hand, for hot temperature pipes the main problem is mainly due to corrosion/erosion attack inside the pipe. In the study of corrosion in pipelines, one of the most important parameters to be monitored and measured is the wall thickness. Currently, most pipeline corrosion monitoring and evaluation for both insulated and non-insulated pipes is performed using an ultrasonic method. The most common technique is that based on the A-Scan, using either a normal flaw detector or some form of dedicated equipment. However, with recent development of ultrasonic technology, more advance method, namely B-Scan and C-scan techniques are also available. The most notable disadvantage of using this method is that the insulation covering the pipe has to be removed before the inspection can be carried out and this is considered as not so cost effective. Due to this reason, the possibility of employing other alternative NDT method, namely radiographic testing method were studied. The technique used in this studied are known as tangential technique. In this study it was found that the result found using tangential technique is consistent with the actual thickness of the pipe. Result of this study is presented and discussed in this paper. (Author)

Microstructural characterization of pipe bomb fragments

International Nuclear Information System (INIS)

Gregory, Otto; Oxley, Jimmie; Smith, James; Platek, Michael; Ghonem, Hamouda; Bernier, Evan; Downey, Markus; Cumminskey, Christopher

2010-01-01

Recovered pipe bomb fragments, exploded under controlled conditions, have been characterized using scanning electron microscopy, optical microscopy and microhardness. Specifically, this paper examines the microstructural changes in plain carbon-steel fragments collected after the controlled explosion of galvanized, schedule 40, continuously welded, steel pipes filled with various smokeless powders. A number of microstructural changes were observed in the recovered pipe fragments: deformation of the soft alpha-ferrite grains, deformation of pearlite colonies, twin formation, bands of distorted pearlite colonies, slip bands, and cross-slip bands. These microstructural changes were correlated with the relative energy of the smokeless powder fillers. The energy of the smokeless powder was reflected in a reduction in thickness of the pipe fragments (due to plastic strain prior to fracture) and an increase in microhardness. Moreover, within fragments from a single pipe, there was a radial variation in microhardness, with the microhardness at the outer wall being greater than that at the inner wall. These findings were consistent with the premise that, with the high energy fillers, extensive plastic deformation and wall thinning occurred prior to pipe fracture. Ultimately, the information collected from this investigation will be used to develop a database, where the fragment microstructure and microhardness will be correlated with type of explosive filler and bomb design. Some analyses, specifically wall thinning and microhardness, may aid in field characterization of explosive devices.

Application of failure assessment diagram methods to cracked straight pipes and elbows

International Nuclear Information System (INIS)

Ainsworth, R.A.; Gintalas, M.; Sahu, M.K.; Chattopadhyay, J.; Dutta, B.K.

2016-01-01

This paper reports fracture assessments of large-scale straight pipes and elbows of various pipe diameters and crack sizes. The assessments estimate the load for ductile fracture initiation using the failure assessment diagram method. Recent solutions in the literature for stress intensity factor and limit load provide the analysis inputs. An assessment of constraint effects is also performed using recent solutions for elastic T-stress. It is found that predictions of initiation load are close to the experimental values for straight pipes under pure bending. For elbows, there is generally increased conservatism in the sense that the experimental loads are greater than those predicted. The effects of constraint are found not to be a major contributor to the initiation fracture assessments but may have some influence on the ductile crack extension. - Highlights: • This paper presents assessments of the loads for ductile fracture initiation in 21 large-scale piping tests. • Modern stress intensity factor and limit load solutions were used for standard failure assessment diagram methods. • This leads to generally accurate assessments of the loads for ductile crack initiation. • The effects of constraint are found not to be a major contributor to the initiation fracture assessments.

Probabilistic analyses of failure in reactor coolant piping

International Nuclear Information System (INIS)

Holman, G.S.

1984-01-01

LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

Effect of yield stress matching on ductile fracture behavior of girth welds for X line pipe

Energy Technology Data Exchange (ETDEWEB)

Motohashi, Hiroyuki; Hagiwara, Naoto [Tokyo Gas Co., Ltd. (Japan)

2005-07-01

This paper describes the effects of yield stress matching on the ductile fracture behavior of girth welded joints for X linepipes. Three welded joints were made on an X line pipe using several consumables to obtain about a 20% overmatched, even matched and about a 20% under matched weld metal. For these three welded joints, curved wide plate tensile tests were then conducted with a surface notch in the weld metal. To determine the ductile crack initiation from the surface notch, these tests employed a direct-current electric potential (d-c E P) method. Crack opening displacement, gauge length strain and local strain adjacent to the surface notch were also measured. The ductile crack initiation was successfully detected using the d-c E P method. The yield stress matching significantly affected the ductile crack initiation and fracture behavior, that is, the overmatched welded joint had a higher resistance to ductile fracture than that of the under matched welded joint. The allowable strength matching level was determined from the relationship between the strength matching and the gauge length strain at the ductile crack initiation detected using the d-c E P method. (author)

Crack propagation and arrest simulation of X90 gas pipe

International Nuclear Information System (INIS)

Yang, Fengping; Huo, Chunyong; Luo, Jinheng; Li, He; Li, Yang

2017-01-01

To determine whether X90 steel pipe has enough crack arrest toughness or not, a damage model was suggested as crack arrest criterion with material parameters of plastic uniform percentage elongation and damage strain energy per volume. Fracture characteristic length which characterizes fracture zone size was suggested to be the largest mesh size on expected cracking path. Plastic uniform percentage elongation, damage strain energy per volume and fracture characteristic length of X90 were obtained by five kinds of tensile tests. Based on this criterion, a length of 24 m, Φ1219 × 16.3 mm pipe segment model with 12 MPa internal gas pressure was built and computed with fluid-structure coupling method in ABAQUS. Ideal gas state equation was used to describe lean gas behavior. Euler grid was used to mesh gas zone inside the pipe while Lagrangian shell element was used to mesh pipe. Crack propagation speed and gas decompression speed were got after computation. The result shows that, when plastic uniform percentage elongation is equal to 0.054 and damage strain energy per volume is equal to 0.64 J/mm"3, crack propagation speed is less than gas decompression speed, which means the simulated X90 gas pipe with 12 MPa internal pressure can arrest cracking itself. - Highlights: • A damage model was suggested as crack arrest criterion. • Plastic uniform elongation and damage strain energy density are material parameters. • Fracture characteristic length is suggested to be largest mesh size in cracking path. • Crack propagating simulation with coupling of pipe and gas was realized in ABAQUS. • A Chinese X90 steel pipe with 12 MPa internal pressure can arrest cracking itself.

A New Method for Fracturing Wells Reservoir Evaluation in Fractured Gas Reservoir

Directory of Open Access Journals (Sweden)

Jianchun Guo

2014-01-01

Full Text Available Natural fracture is a geological phenomenon widely distributed in tight formation, and fractured gas reservoir stimulation effect mainly depends on the communication of natural fractures. Therefore it is necessary to carry out the evaluation of this reservoir and to find out the optimal natural fractures development wells. By analyzing the interactions and nonlinear relationships of the parameters, it establishes three-level index system of reservoir evaluation and proposes a new method for gas well reservoir evaluation model in fractured gas reservoir on the basis of fuzzy logic theory and multilevel gray correlation. For this method, the Gaussian membership functions to quantify the degree of every factor in the decision-making system and the multilevel gray relation to determine the weight of each parameter on stimulation effect. Finally through fuzzy arithmetic operator between multilevel weights and fuzzy evaluation matrix, score, rank, the reservoir quality, and predicted production will be gotten. Result of this new method shows that the evaluation of the production coincidence rate reaches 80%, which provides a new way for fractured gas reservoir evaluation.

Fatigue evaluation of piping systems with limited vibration test data

International Nuclear Information System (INIS)

Huang, S.N.

1990-11-01

The safety-related piping in a nuclear power plant may be subjected to pump- or fluid-induced vibrations that, in general, affect only local areas of the piping systems. Pump- or fluid-induced vibrations typically are characterized by low levels of amplitudes and a high number of cycles over the lifetime of plant operation. Thus, the resulting fatigue damage to the piping systems could be an important safety concern. In general, tests and/or analyses are used to evaluate and qualify the piping systems. Test data, however, may be limited because of lack of instrumentation in critical piping locations and/or because of difficulty in obtaining data in inaccessible areas. This paper describes and summarizes a method to use limited pipe vibration test data, along with analytical harmonic response results from finite-element analyses, to assess the fatigue damage of nuclear power plant safety-related piping systems. 5 refs., 2 figs., 11 tabs

Fracture dynamics of a propagating crack in a pressurized ductile cylinder

International Nuclear Information System (INIS)

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

1977-01-01

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

Applications of a fracture mechanics model of structural reliability to the effects of seismic events on reactor piping

International Nuclear Information System (INIS)

Harris, D.O.; Lim, E.Y.

1982-01-01

A fracture mechanics model of structural reliability is described. The model assumes that failure occurs due to the subcritical and catastrophic growth of as-fabricated defects. The material properties, stress history, number and dimensions of the initial cracks are treated as random variables. Crack growth is calculated using fracture mechanics principles. The model has been used to estimate the influence of earthquakes on the integrity of circumferential girth butt welds in the large (diameter greater than 30 in.) primary coolant system pipes of a commercial pressurized water reactor. In the absence of earthquakes, the probability of leaks and catastrophic double-ended guillotine breaks is estimated to be 10 -6 and 10 -12 per plant lifetime, respectively. These probabilities were only slightly increased by the occurrence of earthquakes. (author)

Evaluation of burst pressure prediction models for line pipes

Energy Technology Data Exchange (ETDEWEB)

Zhu, Xian-Kui, E-mail: zhux@battelle.org [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States); Leis, Brian N. [Battelle Memorial Institute, 505 King Avenue, Columbus, OH 43201 (United States)

2012-01-15

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487-492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: Black-Right-Pointing-Pointer This paper evaluates different burst pressure prediction models for line pipes. Black-Right-Pointing-Pointer The existing models are categorized into two major groups of Tresca and von Mises solutions. Black-Right-Pointing-Pointer Prediction quality of each model is assessed statistically using a large full-scale burst test database. Black-Right-Pointing-Pointer The Zhu-Leis solution is identified as the best predictive model.

Evaluation of burst pressure prediction models for line pipes

International Nuclear Information System (INIS)

Zhu, Xian-Kui; Leis, Brian N.

2012-01-01

Accurate prediction of burst pressure plays a central role in engineering design and integrity assessment of oil and gas pipelines. Theoretical and empirical solutions for such prediction are evaluated in this paper relative to a burst pressure database comprising more than 100 tests covering a variety of pipeline steel grades and pipe sizes. Solutions considered include three based on plasticity theory for the end-capped, thin-walled, defect-free line pipe subjected to internal pressure in terms of the Tresca, von Mises, and ZL (or Zhu-Leis) criteria, one based on a cylindrical instability stress (CIS) concept, and a large group of analytical and empirical models previously evaluated by Law and Bowie (International Journal of Pressure Vessels and Piping, 84, 2007: 487–492). It is found that these models can be categorized into either a Tresca-family or a von Mises-family of solutions, except for those due to Margetson and Zhu-Leis models. The viability of predictions is measured via statistical analyses in terms of a mean error and its standard deviation. Consistent with an independent parallel evaluation using another large database, the Zhu-Leis solution is found best for predicting burst pressure, including consideration of strain hardening effects, while the Tresca strength solutions including Barlow, Maximum shear stress, Turner, and the ASME boiler code provide reasonably good predictions for the class of line-pipe steels with intermediate strain hardening response. - Highlights: ► This paper evaluates different burst pressure prediction models for line pipes. ► The existing models are categorized into two major groups of Tresca and von Mises solutions. ► Prediction quality of each model is assessed statistically using a large full-scale burst test database. ► The Zhu-Leis solution is identified as the best predictive model.

Ductile fracture theories for pressurised pipes and containers

Science.gov (United States)

Erdogan, F.

1976-01-01

Two mechanisms of fracture are distinguished. Plane strain fractures occur in materials which do not undergo large-scale plastic deformations prior to and during a possible fracture deformation. Plane stress or high energy fractures are generally accompanied by large inelastic deformations. Theories for analyzing plane stress are based on the concepts of critical crack opening stretch, K(R) characterization, J-integral, and plastic instability. This last is considered in some detail. The ductile fracture process involves fracture initiation followed by a stable crack growth and the onset of unstable fracture propagation. The ductile fracture propagation process may be characterized by either a multiparameter (discrete) model, or some type of a resistance curve which may be considered as a continuous model expressed graphically. These models are studied and an alternative model is also proposed for ductile fractures which cannot be modeled as progressive crack growth phenomena.

Heat pipe effects in nuclear waste isolation: a review

International Nuclear Information System (INIS)

Doughty, C.; Pruess, K.

1985-12-01

The existence of fractures favors heat pipe development in a geologic repository as does a partially saturated medium. A number of geologic media are being considered as potential repository sites. Tuff is partially saturated and fractured, basalt and granite are saturated and fractured, salt is unfractured and saturated. Thus the most likely conditions for heat pipe formation occur in tuff while the least likely occur in salt. The relative permeability and capillary pressure dependences on saturation are of critical importance for predicting thermohydraulic behavior around a repository. Mineral redistribution in heat pipe systems near high-level waste packages emplaced in partially saturated formations may significantly affect fluid flow and heat transfer processes, and the chemical environment of the packages. We believe that a combined laboratory, field, and theoretical effort will be needed to identify the relevant physical and chemical processes, and the specific parameters applicable to a particular site. 25 refs., 1 fig

CT evaluation of acetabular fractures

Energy Technology Data Exchange (ETDEWEB)

Piazza, P; Girelli, G; Coran, F; Lutman, M

1986-01-01

The paper deals with sixteen cases of acetabular fractures studied with CT. After a short description of the normal CT findings, the different kind of fractures are reported. The usefulness of CT examination in evaluating acetabular fractures and their complications is confirmed both in conservative treatment and surgical approach.

Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

International Nuclear Information System (INIS)

Lu, S.C.; Larder, R.A.; Ma, S.M.

1981-01-01

This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Evaluation of KALIMER IHTS piping using French RCC-MR code

International Nuclear Information System (INIS)

Lee, Hyeong Yeon; Kim, J. B.; Lee, J. H.

2001-12-01

In the present report, the evaluation of design integrity for the liquid metal reactor(LMR) of KALIMER IHTS(intermediate heat transport system) piping according to the French design guideline of RCC-MR RC3600 developed for secondary piping of LMR and the evaluation procedure was presented. The evaluation results showed that the results by the simple RC-3600 procedure of design by formula were more conservative than those of ASME section III subsection NH of the design by analysis for the class I structural components

Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

International Nuclear Information System (INIS)

Jong, Rudolf Peter de

2004-01-01

Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

Degradation of safety injection system and containment spray piping and tank fracture toughness analysis

International Nuclear Information System (INIS)

Douglas, A.; Doubel, P.; Wicker, C.

2011-01-01

Extensive stress corrosion cracking (SCC), induced by the marine environment and the presence of high residual stresses arising from the respective manufacturing processes has been encountered in the safety injection system piping (RIS), containment spray system piping (EAS) and reactor and spent fuel storage tank (PTR), or refuelling water storage tank (RWST) of the Koeberg plant. Type 304L steels from the RIS system and replacement components for the RIS and RWST systems have been subject to mechanical and fracture toughness testing. The following conclusions have been drawn. -) The piping sections of both the original and replacement components exhibit residual cold work. The level of cold work imparted to the piping and elbow have been estimated to be 2, 2 to 3, 9% and 5, 7 to 7, 3% respectively. -) Re-annealing produces different responses in type 304L as a function of prior cold work level. Re-annealing of material cold worked to low levels i.e. 3.5% maintain the cold worked level of UTS but can exhibit 0, 2% PS. levels below that of the mill annealed condition. There is the potential for the ASTM A312 minimum 0, 2% level to be breached. At higher levels of cold work i.e. 7% re-annealing results in extensive grain growth, a significant reduction in 0, 2% PS from the mill annealed condition and the recovery of the UTS to the mill annealed level. -) Cold work at the levels obtained significantly reduces the SOL initiation toughness Ji. The reduction in toughness can be greater than 50%. The resistance to ductile crack propagation, dJ/da, remains unchanged at least up to 5 % cold work. -) The defect assessment for the RIS/EAS systems have used highly conservative values of initiation toughness such that no crack initiation would occur under the loading conditions considered and in a non-hostile environment. -) Under the marine environment to which the RIS/EAS components are still subjected, the limiting criterion for operation of the RIS/EAS system remains a

Evaluation on damage of pipe using ultrasonic and acoustic emission

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Kyung; Lee, Sang Pill; Lee, Moon Hee [Dongeui Univ., Busan (Korea, Republic of); Lee, Joon Hyun [Pusan National Univ., Busan (Korea, Republic of)

2008-07-01

An elastic waves like ultrasonic and acoustic emission were used to evaluate the propagating properties of the wave in pipe, and study on mode conversion of the elastic wave due to the defects on the pipe was performed. In this study an Acoustic Emission (AE) sensor was used to receive the propagated ultrasonic wave. AE technique has a advantage that it can identify the received ultrasonic wave by the analysis of the AE parameters such as count, energy, frequency, duration time and amplitude. For transmitting and receiving of the wave, an universal angle wedge was manufactured. The optimum angles for transmitting of ultrasonic wave and signal receiving at the attached AE sensor on the pipe were determined. Theoretical dispersion curve was compared with the results of the time-frequency analysis based on the wavelet transformation. The received modes showed a good agreement with theoretical one. The used ultrasonic sensor was 1MHz, and AE sensor was broadband. The artificial cracks were induced in the pipe to measure the propagation characteristics of the elastic wave for the cracks. AE parameters for the received signals were also varied with the crack types in the pipe. AE parameters of amplitude and duration time were more effective factors than the analysis of mode conversion for evaluation of the cracks in the pipe.

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.

Piping and erosion in buffer and backfill materials. Current knowledge

International Nuclear Information System (INIS)

Boergesson, Lennart; Sanden, Torbjoern

2006-09-01

The water inflow into the deposition holes and tunnels in a repository will mainly take place through fractures in the rock and will lead to that the buffer and backfill will be wetted and homogenised. But in general the buffer and backfill cannot absorb all water that runs through a fracture, which leads to that a water pressure will be generated in the fracture when the inflow is hindered. If the counter pressure and strength of the buffer or backfill is insufficiently high, piping and subsequent erosion may take place. The processes and consequences of piping and erosion have been studied in some projects and several laboratory test series in different scales have been carried through. This brief report describes these tests and the results and conclusions that have emerged. The knowledge of piping and erosion is insufficient today and additional studies are needed and running

Evaluation of seismic margins for an in-plant piping system

International Nuclear Information System (INIS)

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

1991-01-01

Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

Comparison and evaluation of flexible and stiff piping systems

International Nuclear Information System (INIS)

Hahn, W.; Tang, H.T.; Tang, Y.K.

1983-01-01

An experimental and numerical study was performed on a piping system, with various support configurations, to assess the difference in piping response for flexible and stiff piping systems. Questions have arisen concerning a basic design philosophy employed in present day piping designs. One basic question is, the reliability of a flexible piping system greater than that of a stiff piping system by virtue of the fact that a flexible system has fewer snubber supports. With fewer snubbers, the pipe is less susceptible to inadvertent thermal stresses introduced by snubber malfunction during normal operation. In addition to the technical issue, the matter of cost savings in flexible piping system design is a significant one. The costs associated with construction, in-service inspection and maintenance are all significantly reduced by reducing the number of snubber supports. The evaluation study, sponsored by the Electric Power Research Institute, was performed on a boiler feedwater line at Consolidated Edison's Indian Point Unit 1. In this study, the boiler feedwater line was tested and analyzed with two fundamentally different support systems. The first system was very flexible, employing rod and spring hangers, and represented the 'old' design philosophy. The pipe system was very flexible with this support system, due to the long pipe span lengths between supports and the fact that there was only one lateral support. This support did not provide much restraint since it was near an anchor. The second system employed strut and snubber supports and represented the 'modern' design philosophy. The pipe system was relatively stiff with this support system, primarily due to the increased number of supports, including lateral supports, thereby reducing the pipe span lengths between supports. The second support system was designed with removable supports to facilitate interchange of the supports with different support types (i.e., struts, mechanical snubbers and hydraulic

Development of new assessment methodology for locally corroded pipe

International Nuclear Information System (INIS)

Lim, Hwan; Shim, Do Jun; Kim, Yun Jae; Kim, Young Jin

2002-01-01

In this paper, a unified methodology based on the local stress concept to estimate residual strength of locally thinned pipes is proposed. An underlying idea of the proposed methodology is that the local stress in the minimum section for locally thinned pipe is related to the reference stress, popularly used in creep problems. Then the problem remains how to define the reference stress, that is the reference load. Extensive three-dimensional Finite Element (FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Based on these FE results, the reference load is proposed, which is independent of materials. A natural outcome of this method is the maximum load capacity. By comparing with existing test results, it is shown that the reference stress is related to the fracture stress, which in turn can be posed as the fracture criterion of locally thinned pipes. The proposed method is powerful as it can be easily generalised to more complex problems, such as pipe bends and tee-joints

Analysis of cantilever pipes in transverse fluid flow with motion limiting stopper at the free end

International Nuclear Information System (INIS)

Jiyavan, R.

1983-01-01

Flow-induced vibration in heat exchanger tubes can result in impact with the baffle plates and subsequent tube failure through fatigue, fracture and fretting wear. As a step towards the correlation between the random flow excitations and the rate of wear, this paper presents a general theory for predicting the tube motion and the tube baffle impact forces through a case of cantilever pipe with motion limiting stopper at the free end and simultaneously subjected to transverse fluid flow. The mathematical model has been developed using the theory of fluid-structure interactions with model superposition technique. The pipe displacement induced by lift forces is evaluated by numerical integration. When displacement increases to greater than the pipe-stopper clearance, the pipe impacts on stopper. Assuming semielastic impact, the equation of pipe motion during impact is developed using extended Hertz's theory to include the vibration of one of the colliding bodies. The stopper is assumed to be at rest before and after the impact. The constraint imposed on pipe motion, at the free end due to impact of the pipe on stopper, is considered as one of the boundary conditions and is used to evaluate the pipe natural frequencies. The nonlinear equations are solved numerically. The response of the pipe due to wake induced lift forces superposed by the impact response is evaluated. (orig./GL)

On the impact bending test technique for high-strength pipe steels

Science.gov (United States)

Arsenkin, A. M.; Odesskii, P. D.; Shabalov, I. P.; Likhachev, M. V.

2015-10-01

It is shown that the impact toughness (KCV-40 = 250 J/cm2) accepted for pipe steels of strength class K65 (σy ≥ 550 MPa) intended for large-diameter gas line pipes is ineffective to classify steels in fracture strength. The results obtained upon testing of specimens with a fatigue crack and additional sharp lateral grooves seem to be more effective. In energy consumption, a macrorelief with splits is found to be intermediate between ductile fracture and crystalline brittle fracture. A split formation mechanism is considered and a scheme is proposed for split formation.

Participation of the GRS in the 'Degraded piping program' of the USNRC

International Nuclear Information System (INIS)

Azodi, D.; Hoefler, A.; Sievers, J.

1989-01-01

During the course of the Degraded Piping Program, Battelle was the organizer of three round-robin activities for advanced elasto-plastic fracture mechanics calculations (e.g. finite element method and J-estimation scheme). GRS participated in all of them and submitted finite element and J-estimation scheme results. As a main result of round-robin on elasto-plastic fracture mechanics calculations: Based on finite element calculations, the J-integral method (energy release rate) provided the ability to describe the fracture behaviour of flawed piping even in a very ductile material. (orig./HP) [de

Evaluation of wall thinning profile by flow accelerated corrosion in separation and union pipe

International Nuclear Information System (INIS)

Watanabe, Shun; Yoneda, Kimitoshi

2013-01-01

Flow Accelerated Corrosion (FAC) is a pipe wall thinning phenomena to be monitored and managed in power plants with high priority. At present, its management has been conducted with conservative evaluation of thinning rate and residual lifetime of the piping based on wall thickness measurements. However, noticeable case of wall thinning was occurred at separation and union pipe. In such pipe system, it is a problem to manage section beneath reinforcing plate of T-tube pipe and 'crotch' of T-joint pipe; the region where wall thickness measurement is difficult to conduct with ordinary ultrasonic testing device. In this study, numerical analysis for separation and union part of T-tube and T-joint pipe was conducted, and wall thinning profile by Flow Accelerated Corrosion was evaluated by calculating mass transfer coefficient and geometry factor. Based on these results, we considered applicable wall thinning management for T-tube and T-joint pipe. In the case of union flow from main and branch pipe, the wall thinning profile of T-tube showed the tendency of increase at main pipe like semielliptical region. On the other hand, noticeable profile appeared at 'crotch' in T-joint. Although it was found that geometry factor of T-joint in this case was half the value of T-tube, an alternative evaluation method to previous one might be needed for the profiles of 'semielliptical region' and 'crotch'. (author)

Development of total systems of piping stress analysis and evaluation: ISAPPS

International Nuclear Information System (INIS)

Oki, Teizaburo; Koyanagi, Ryoichi; Fukuda, Masanao

1978-01-01

IHI has developed the systems of piping stress analysis and evaluation: ISAPPS (IHI Stress Analysis Program for Piping Systems), which are further described in this paper. In addition, the results of structural analysis and heat transfer analysis were confirmed. An example of stress evaluation in accordance with the modified ASME Code Sec. III is shown. ISAPPS consists of the following seven parts, and is designed for easy adoption of other programs by making modifications. 1. Piping design oriented language programs 2. Structural analysis programs 3. Isometric plotting programs 4. Multi-file dumping program 5. Load combination program 6. Heat transfer program 7. Stress evaluation programs As one of the examples of structural analysis programs, IHI make use of the modified SAP IV developed by the University of California. Evaluations of stresses are performed in accordance with: 1. ASME Boiler and Pressure Vessel Code, Sec. III Class 1, 2 and 3 2. ANSI Code, B31.1 and B31.3 3. MITI (Ministry of International Trade and Industry ) Code ISAPPS is very useful for design of nuclear and chemical pipings and so on. (author)

Nuclear piping and pipe support design and operability relating to loadings and small bore piping

International Nuclear Information System (INIS)

Stout, D.H.; Tubbs, J.M.; Callaway, W.O.; Tang, H.T.; Van Duyne, D.A.

1994-01-01

The present nuclear piping system design practices for loadings, multiple support design and small bore piping evaluation are overly conservative. The paper discusses the results developed for realistic definitions of loadings and loading combinations with methodology for combining loads under various conditions for supports and multiple support design. The paper also discusses a simplified method developed for performing deadweight and thermal evaluations of small bore piping systems. Although the simplified method is oriented towards the qualification of piping in older plants, this approach is applicable to plants designed to any edition of the ASME Section III or B31.1 piping codes

Development of carbon steel with superior resistance to wall thinning and fracture for nuclear piping system

International Nuclear Information System (INIS)

Rhee, Chang Kyu; Lee, Min Ku; Park, Jin Ju

2010-07-01

Carbon steel is usually used for piping for secondary coolant system in nuclear power plant because of low cost and good machinability. However, it is generally reported that carbon steel was failed catastrophically because of its low resistance to wall thinning and fracture toughness. Especially, flow accelerated corrosion (FAC) is one of main problems of the wall thinning of piping in the nuclear power plant. Therefore, in this project, fabrication technology of new advanced carbon steel materials modified by dispersion of nano-carbide ceramics into the matrix is developed first in order to improve the resistance to wall thinning and fracture toughness drastically compared to the conventional one. In order to get highly wettable fine TiC ceramic particles into molten metal, the micro-sized TiC particles were first mechanically milled by Fe (MMed TiC/Fe) in a high energy ball mill machine in Ar gas atmosphere, and then mixed with surfactant metal elements (Sn, Cr, Ni) to obtain better wettability, as this lowered surface tension of the carbon steel melt. According to microscopic images revealed that an addition of MMed TiC/Fe-surfactant mixed powders favorably disperses the fine TiC particles in the carbon steel matrix. It was also found that the grain size refinement of the cast matrix is achieved remarkably when fine TiC particles were added due to the fact that they act as nucleation sites during the solidification process. As a results, a cast carbon steel dispersed with fine TiC particles shows improved mechanical properties such as hardness, tensile strength and cavitation resistance compared to that of without particles. However, the slight decrease of toughness was found

Corrosion and deposit evaluation in large diameter pipes using radiography

International Nuclear Information System (INIS)

Boateng, A.

2012-01-01

The reliability and safety of industrial equipment in the factories and processing industries are substantially influenced by degradation processes such as corrosion, erosion, deposits and blocking of pipes. These might lead to low production, unpredictable and costly shutdowns due to repair and replacement and sometimes combined environmental pollution and risk of personnel injuries. Only periodic inspection for the integrity of pipes and equipment can reduce the risk in connection with other maintenance activities. The research explored two methods of radiographic inspection techniques, the double wall technique and the tangential radiographic technique using Ir-192 for evaluating deposits and corrosion attacks across the inner and outer walls of steel pipes with diameter greater than 150 mm with or without insulation. The application of both techniques was conducted depending on pipe diameter, wall thickness, radiation source (Ir-92) and film combination. The iridium source was positioned perpendicular with respect to the pipe axis projecting the double wall of the pipe on the plated radiographic film. With the tangential radiographic technique, the source was placed tangential to the pipe wall and because of its large diameter, the source was collimated to prevent backscatter and also to focus the beam at the target area of interest. All measurements were performed on special designed test pieces to simulate corrosion attack and deposits on industrial pipes. Pitting corrosion measurements based on Tangential Radiographic Technique were more sophisticated, and therefore magnification factor and correction were used to establish the estimated pit depth on the film. The insulating material used to conserve the thermodynamic properties of the transported media had relatively negligible attenuation coefficient compared to the concrete deposit. The two explored techniques were successful in evaluating corrosion attack and deposit on the walls of the pipe and the risk

Investigation of fatigue and mechanical properties of the pipe grade poly(vinyl chloride using recycled scraps

Directory of Open Access Journals (Sweden)

J-M. Lee

2015-04-01

Full Text Available In this paper, the effect of using pre-consumer PVC scraps on static and long-term mechanical properties is studied. The degradation characteristics of mixing virgin PVC with crushed pre-consumer and PVC pipe scraps are analyzed using various tools including Gel Permeation Chromatography (GPC, Thermogravimetric Analysis (TGA, X-ray fluorescence (XRF and Fourier Transform Infrared (FTIR spectroscopy. The variation of static mechanical properties as a function of adding pre-consumer PVC pipe scraps is investigated using the degradation analyses of recycled PVC scraps. In addition, fatigue tests are executed to evaluate the long-term durability of blending virgin PVC and recycled PVC scraps, and the fracture surface is investigated in detail to reveal the variation of the fracture mechanisms.

Evaluating Program about Performance of Circular Sodium Heat Pipe

International Nuclear Information System (INIS)

Kwak, Jae Sik; Kim, Hee Reyoung

2014-01-01

The superior heat transfer capability, structural simplicity, relatively inexpensive, insensitivity to the gravitational field, silence and reliability are some of its outstanding features. We study about heat transfer equation of heat pipe and program predicting performance which is considering geometrical shape of heat pipe by the related heat transfer equation of heat pipe. The operating temperature is 450 .deg. C - 950 .deg. C, working fluid is sodium, material for container is stainless steel, and type of wick is sintered metal. As a result of evaluating program about performance of circular sodium heat pipe based on MATLAB code, express correlation between radius and LHR, correlation between heat transfer length and LHR, correlation between wick and LHR, correlation between operating temperature and LHR. Generally radius values of heat pipe are proportional to LHR because of increase of mass flow which is main factor of heat flow. Heat transfer length values of heat pipe are inversely proportional to LHR and slightly inversely proportional to heat rate. Pore size is proportional to LHR. Although increase of pore size decrease capillary pressure, decrease more pressure drop in liquid phase. As a result, mass flow and heat rate are increase. But we have to do additional consideration about pore size and voidage in the aspect of safety and production technique

Evaluating Program about Performance of Circular Sodium Heat Pipe

Energy Technology Data Exchange (ETDEWEB)

Kwak, Jae Sik; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The superior heat transfer capability, structural simplicity, relatively inexpensive, insensitivity to the gravitational field, silence and reliability are some of its outstanding features. We study about heat transfer equation of heat pipe and program predicting performance which is considering geometrical shape of heat pipe by the related heat transfer equation of heat pipe. The operating temperature is 450 .deg. C - 950 .deg. C, working fluid is sodium, material for container is stainless steel, and type of wick is sintered metal. As a result of evaluating program about performance of circular sodium heat pipe based on MATLAB code, express correlation between radius and LHR, correlation between heat transfer length and LHR, correlation between wick and LHR, correlation between operating temperature and LHR. Generally radius values of heat pipe are proportional to LHR because of increase of mass flow which is main factor of heat flow. Heat transfer length values of heat pipe are inversely proportional to LHR and slightly inversely proportional to heat rate. Pore size is proportional to LHR. Although increase of pore size decrease capillary pressure, decrease more pressure drop in liquid phase. As a result, mass flow and heat rate are increase. But we have to do additional consideration about pore size and voidage in the aspect of safety and production technique.

A quantitative evaluation of seismic margin of typical sodium piping

International Nuclear Information System (INIS)

Morishita, Masaki

1999-05-01

It is widely recognized that the current seismic design methods for piping involve a large amount of safety margin. From this viewpoint, a series of seismic analyses and evaluations with various design codes were made on typical LMFBR main sodium piping systems. Actual capability against seismic loads were also estimated on the piping systems. Margins contained in the current codes were quantified based on these results, and potential benefits and impacts to the piping seismic design were assessed on possible mitigation of the current code allowables. From the study, the following points were clarified; 1) A combination of inelastic time history analysis and true (without margin)strength capability allows several to twenty times as large seismic load compared with the allowable load with the current methods. 2) The new rule of the ASME is relatively compatible with the results of inelastic analysis evaluation. Hence, this new rule might be a goal for the mitigation of seismic design rule. 3) With this mitigation, seismic design accommodation such as equipping with a large number of seismic supports may become unnecessary. (author)

Development on methods for evaluating structure reliability of piping components

International Nuclear Information System (INIS)

Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.

2003-01-01

In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)

Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

International Nuclear Information System (INIS)

1975-10-01

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

WHIPJET progress on piping restraint elimination at Beaver Valley - 2

International Nuclear Information System (INIS)

Server, W.L.; Szy Slow Ski, J.J.; Goldstein, N.A.

1986-01-01

Fracture mechanics technology has advanced to the point that an engineering approach using the concept of leak-before-break in lieu of postulating double-ended pipe rupture is now possible. An approach based upon this fracture mechanics technology, termed WHIPJET, is currently being applied to Beaver Valley Power Station, Unit 2 for Duquesne Light Company. The WHIPJET philosophy is simple, conservative, and provides defense-in-depth arguments for high energy piping throughout the balance-of-plant. Progress being made in applying WHIPJET to several lines is presented

Crack stability analysis of low alloy steel primary coolant pipe

Energy Technology Data Exchange (ETDEWEB)

Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

1997-04-01

At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

1998 Annual Study Report. Standardization of methods for evaluating characteristics of high-strength, large-diameter steel pipes for superhigh-pressure natural gas pipelines; 1998 nendo seika hokokusho. Chokoatsu tennen gas pipeline yo kokyodo daikei kokan no tokusei hoho no hyojunka

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-03-01

The pipelines for safely transmitting superhigh-pressure natural gas should have excellent characteristics. The steel pipe is required to have a sufficient toughness, more concretely Charpy impact-absorbing energy, to prevent propagating shear fracture characteristic of natural gas pipelines. Recently, the natural gas pipeline is increasingly required to have higher design pressures (15 Mpa or higher) and grade (X80 or higher). In order to develop the techniques for simulating crack propagation in the propagating shear fracture of natural gas pipe lines as part of the programs to cope with these trends, the 1998 efforts were directed to reviewing the research results obtained so far and analysis of the problems to be solved and tasks to be taken, based on which the analytical procedure for gas releasing phenomena during the fracture process was basically developed, the material characteristic data were collected by the laboratory scale toughness tests, and the preliminary tests with rupture disks were conducted to verify the above analytical procedure. These efforts have established the bases for evaluating the characteristics of high-strength, large-diameter steel pipes in the light of safety against fracture, and greatly advanced the program towards the final target of developing the international specification drafts for toughness. (NEDO)

Phase 2 of the International Piping Integrity Research Group programme

International Nuclear Information System (INIS)

Darlaston, B.J.

1994-01-01

The results of phase 1 of the International Piping Integrity Research Group (IPIRG-1) programme have been widely reported. The significance of the results is reviewed briefly, in order to put the phase 2 programme into perspective. The success of phase 1 led the participants to consider further development and validation of pipe and pipe component fracture analysis technology as part of another international group programme (IPIRG-2). The benefits of combined funding and of the technical exchanges and interactions are considered to be of significant advantage and value. The phase 2 programme has been designed with the overall objective of developing and experimentally validating methods of predicting the fracture behaviour of nuclear reactor safety-related piping, to both normal operating and accident loads. The programme will add to the engineering estimation analysis methods that have been developed for straight pipes. The pipe system tests will expand the database to include seismic loadings and flaws in fittings, such as bends, elbows and tees, as well as ''short'' cracks. The results will be used to validate further the analytical methods, expand the capability to make fittings and extend the quasi-static results for the USNRC's new programme on short cracks in piping and piping welds. The IPIRG-2 programme is described to provide a clear understanding of the content, strategy, potential benefits and likely significance of the work. ((orig.))

The effect of cyclic and dynamic loads on carbon steel pipe

International Nuclear Information System (INIS)

Rudland, D.L.; Scott, P.M.; Wilkowski, G.M.

1996-02-01

This report presents the results of four 152-mm (6-inch) diameter, unpressurized, circumferential through-wall-cracked, dynamic pipe experiments fabricated from STS410 carbon steel pipe manufactured in Japan. For three of these experiments, the through-wall crack was in the base metal. The displacement histories applied to these experiments were a quasi-static monotonic, dynamic monotonic, and dynamic, cyclic (R = -1) history. The through-wall crack for the third experiment was in a tungsten-inert-gas weld, fabricated in Japan, joining two lengths of STS410 pipe. The displacement history for this experiment was the same history applied to the dynamic, cyclic base metal experiment. The test temperature for each experiment was 300 C (572 F). The objective of these experiments was to compare a Japanese carbon steel pipe material with US pipe material, to ascertain whether this Japanese steel was as sensitive to dynamic and cyclic effects as US carbon steel pipe. In support of these pipe experiments, quasi-static and dynamic, tensile and fracture toughness tests were conducted. An analysis effort was performed that involved comparing experimental crack initiation and maximum moments with predictions based on available fracture prediction models, and calculating J-R curves for the pipe experiments using the η-factor method

Design Evaluation of a Piping System in the SELFA Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Son, Seok-Kwon; Jo, Young-Chul; Lee, Hyeong-Yeon; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, design evaluations on the SELFA piping system has been conducted according to the ASME B31.1 and RCC-MRx RD-3600. The conservatism of the two codes was quantified based on the evaluation results. It was shown that B31.1 was more conservative for the sustained loads while less conservative for thermal expansion loads when compare with those of RD-3600. However, all the evaluation results according to the two codes were within the code allowables. There are two main piping systems in the SELFA test loop. In this study, the integrity of the SELFA piping system has been evaluated according to the two design-by-rule (DBR) codes of ASME B31.1 and RCC-MRx RD-3600. B31.1 is an industry design code for power piping while RD-3600 is a class 3 nuclear DBR code. The conservatism of the two codes was quantified based on the evaluation results as per the two DBR codes. The sodium test facility of the SELFA is under construction at KAERI for the investigation of thermo-hydraulic behavior of finned-tube sodium-to-air heat exchanger.

Evaluation of remote piping connectors for fusion applications

International Nuclear Information System (INIS)

Welland, H.J.

1985-08-01

Four types of remotely actuated piping connectors were evaluated for their application to the fusion energy program. The evaluations are part of a study being performed by EG and G Idaho, Inc., for the fusion program at Kernforschungszentrum, Karlsruhe, GmbH. Results of the connector evaluation will be used to develop preliminary concepts for connector improvements and new connector configurations

Design and Integrity Evaluation of High-temperature Piping Systems in the STELLA-2 Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Son, Seok-Kwon; Lee, Hyeong-Yeon; Eoh, JaeHyuk; Kim, Jong-Bum; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ju, Yong-Sun [KOASIS Inc., Daejeon (Korea, Republic of)

2016-09-15

In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.

Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

International Nuclear Information System (INIS)

Rodabaugh, E.C.

1983-06-01

The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

A methodology to evaluate the fatigue life of flexible pipes

Energy Technology Data Exchange (ETDEWEB)

Sousa, Fernando J.M. de; Sousa, Jose Renato M. de; Siqueira, Marcos Q. de; Sagrilo, Luis V.S. [Coordenacao dos Programas de Pos-graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Lemos, Carlos Alberto D. de [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil)

2009-07-01

This paper focus on a methodology to perform the fatigue analysis of flexible pipes. This methodology employs functions that convert forces and moments obtained in global analyses into stresses. The stresses are then processed by well-known cycle counting methods, and S-N curves evaluate the damage at several points in the pipe cross-section. Palmgren-Miner linear damage hypothesis is assumed in order to calculate the accumulated fatigue damage. A parametric study on the fatigue life of a flexible pipe employing this methodology is presented. The main points addressed in the study are the influence of friction between layers in the results, the importance of evaluating the fatigue life in various points of the pipe cross-section and the effect of different mean stress levels. The obtained results suggest that the consideration of friction effects strongly influences the fatigue life of flexible risers and these effects have to be accounted both in the global and local analyses of the riser. Moreover, mean stress effects are also significant and at least 8 equally spaced wires in each analyzed section of the riser must be considered in fatigue analyses. (author)

Finite Element Limit Pressures for Circumferential Through-Wall Cracks in the Interface between Elbow and Pipe

International Nuclear Information System (INIS)

Jang, Yoon-Young; Han, Tae-Song; Huh, Nam-Su; Jeong, Jae-Uk

2014-01-01

Among integrity assessment method based on a fracture mechanics concept for piping system, a limit load method is one of the important way to predict a maximum load carrying capacity in the materials with high ductility in the sense that it is used to either assess directly structural integrity of pipe based on fully plastic fracture mechanics or calculate elasticplastic fracture mechanics parameters based on reference stress concept. In nuclear power plants, piping system often involves elbows welded to straight pipe. Since welded regions are vulnerable to cracking, it is important to predict an accurate limit load for pipes with a crack in the interface between elbows and attached pipes. However, although extensive works have been made for developing limit analysis methods for cracked pipes, they were mainly for straight pipes. Recently, limit moment solutions for elbow that is attached to straight pipe with a circumferential through-wall crack(TWC) in the interface were proposed, whereas limit pressure for this geometry is not suggested yet. In this context, plastic limit pressures of circumferential TWCs between elbow and straight pipe were calculated in the present study considering geometric parameters such as an elbow curvature, a pipe size and a crack length. In the present study, the FE plastic limit analyses for circumferential TWC in the interface between elbow and pipe under internal pressure were conducted based on elastic perfectly plastic assumption. Based on the present FE results, it is found that plastic limit pressures of straight pipes with circumferential TWC are not appropriate for predicting plastic limit pressures of circumferential TWC in the interface between elbow and pipe for shorter crack length

J-integral estimation analysis for circumferential throughwall cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

J-integral estimation solution is derived for pipes containing a circumferential throughwall crack. Bending moment and axial tension loadings are considered. These solutions are useful for calculating J from single load-displacement record obtained as part of pipe fracture testing, and are applicable for a wide range of flaw length to pipe circumference ratios. Results for J at initiation of crack growth generated using the solution developed in this paper agree well with J results from finite elements analyses.

Fracture behavior of circumferentially surface-cracked elbows. Technical report, October 1993--March 1996

International Nuclear Information System (INIS)

Kilinski, T.; Mohan, R.; Rudland, D.; Fleming, M.

1996-12-01

This report presents the results from Task 2 of the Second International Piping Integrity Research Group (IPIRG-2) program. The focus of the Task 2 work was directed towards furthering the understanding of the fracture behavior of long-radius elbows. This was accomplished through a combined analytical and experimental program. J-estimation schemes were developed for both axial and circumferential surface cracks in elbows. Large-scale, quasi-static and dynamic, pipe-system, elbow fracture experiments under combined pressure and bending loads were performed on elbows containing an internal surface crack at the extrados. In conjunction with the elbow experiments, material property data were developed for the A106-90 carbon steel and WP304L stainless steel elbow materials investigated. A comparison of the experimental data with the maximum stress predictions using existing straight pipe fracture prediction analysis methods, and elbow fracture prediction methods developed in this program was performed. This analysis was directed at addressing the concerns regarding the validity of using analysis predictions developed for straight pipe to predict the fracture stresses of cracked elbows. Finally, a simplified fitting flaw acceptance criteria incorporating ASME B2 stress indices and straight pipe, circumferential-crack analysis was developed

Fracture behavior of circumferentially surface-cracked elbows. Technical report, October 1993--March 1996

Energy Technology Data Exchange (ETDEWEB)

Kilinski, T.; Mohan, R.; Rudland, D.; Fleming, M. [and others

1996-12-01

This report presents the results from Task 2 of the Second International Piping Integrity Research Group (IPIRG-2) program. The focus of the Task 2 work was directed towards furthering the understanding of the fracture behavior of long-radius elbows. This was accomplished through a combined analytical and experimental program. J-estimation schemes were developed for both axial and circumferential surface cracks in elbows. Large-scale, quasi-static and dynamic, pipe-system, elbow fracture experiments under combined pressure and bending loads were performed on elbows containing an internal surface crack at the extrados. In conjunction with the elbow experiments, material property data were developed for the A106-90 carbon steel and WP304L stainless steel elbow materials investigated. A comparison of the experimental data with the maximum stress predictions using existing straight pipe fracture prediction analysis methods, and elbow fracture prediction methods developed in this program was performed. This analysis was directed at addressing the concerns regarding the validity of using analysis predictions developed for straight pipe to predict the fracture stresses of cracked elbows. Finally, a simplified fitting flaw acceptance criteria incorporating ASME B2 stress indices and straight pipe, circumferential-crack analysis was developed.

Reliability analysis of stainless steel piping using a single stress corrosion cracking damage parameter

International Nuclear Information System (INIS)

Guedri, A.

2013-01-01

This article presents the results of an investigation that combines standard methods of fracture mechanics, empirical correlations of stress-corrosion cracking, and probabilistic methods to provide an assessment of Intergranular Stress Corrosion Cracking (IGSCC) of stainless steel piping. This is done by simulating the cracking of stainless steel piping under IGSCC conditions using the general methodology recommended in the modified computer program Piping Reliability Analysis Including Seismic Events, and by characterizing IGSCC using a single damage parameter. Good correlation between the pipe end-life probability of leak and the damage values were found. These correlations were later used to generalize this probabilistic fracture model. Also, the probability of detection curves and the benefits of in-service inspection in order to reduce the probability of leak for nuclear piping systems subjected to IGSCC were discussed for several pipe sizes. It was found that greater benefits could be gained from inspections for the large pipe as compared to the small pipe sizes. Also, the results indicate that the use of a better inspection procedure can be more effective than a tenfold increase in the number of inspections of inferior quality. -- Highlights: • We simulate the pipe probability of failure under different level of SCC damages. • The residual stresses are adjusted to calibrate the model. • Good correlations between 40-year cumulative leak probabilities and D σ are found. • These correlations were used to generalize this probabilistic fracture model. • We assess the effect of inspection procedures and scenarios on leak probabilities

Design evaluation on sodium piping system and comparison of the design codes

International Nuclear Information System (INIS)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

2015-01-01

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

Design evaluation on sodium piping system and comparison of the design codes

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

2015-03-15

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

Evaluation of clamp effects on LMFBR piping systems

International Nuclear Information System (INIS)

Jones, G.L.

1980-01-01

Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

J-integral estimation analysis for circumferential throughwall cracked pipes

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

J-integral estimation solution is derived for pipes containing a circumferential throughwall crack. Bending moment and axial tension loadings are considered. These solutions are useful for calculating J from single load-displacement record obtained as part of pipe fracture testing, and are applicable for a wide range of flaw length to pipe circumference ratios. Results for J at initiation of crack growth generated using the solution developed in this paper agree well with J results from finite elements analyses. (orig.)

Procedure for seismic evaluation and design of small bore piping

International Nuclear Information System (INIS)

Bilanin, W.; Sills, S.

1991-01-01

Simplified methods for the seismic design of small bore piping in nuclear power plants have teen used for many years. Various number of designers have developed unique methods to treat the large number of class 2 and 3 small bore piping systems. This practice has led to a proliferation of methods which are not standardized in the industry. These methods are generally based on enveloping the results of rigorous dynamic or conservative static analysis and result in an excessive number of supports and unrealistically high support loadings. Experience and test data have become available which warranted taking another look at the present methods for analysis of small bore piping. A recently completed Electric Power Research Institute and NCIG (a utility group) activity developed a new procedure for the seismic design and evaluation of small bore piping which provides significant safety and cost benefits. The procedure streamlines the approach to inertial stresses, which is the main feature that achieves the new benefits. Criteria in the procedure for seismic anchor movement and support design are based analysis and focus the designer on credible failure mechanisms. A walkdown of the as-constructed piping system to identify and eliminate undesirable piping features such as adverse spatial interaction is required

LABORATORY EVALUATION ON PERFORMANCE OF GLASS FIBER REINFORCED PLASTIC MORTAR PIPE CULVERTS

Directory of Open Access Journals (Sweden)

Huawang Shi

2018-04-01

Full Text Available This paper investigated the performance and behaviour of glass fiber reinforced plastic mortar (FRPM pipes under different loading conditions. FRPM pipes with inner diameter of 1500 mm were prefabricated in factory. Mechanics performance testing (ring and axial compressive strength and elastic modulus, stiffness and fatigue test were carried out in laboratory. Ring stiffness test provided pipe stiffness (PS which is a function of geometry and material type of pipe through parallel plate loading test (PPLT. The fatigue test and micro-structure measure method were used to evaluate the durability effects of FRPM under repeated compression load. Results indicated that FRPM pipes had better mechanic performances as the road culverts under soils. It may be helpful for the design and construction of FRPM culverts.

Crack resistance of austenitic pipes with circumferential through-wall cracks

International Nuclear Information System (INIS)

Foerster, K.; Grueter, L.; Setz, W.; Bhandari, S.; Debaene, J.P.; Faidy, C.; Schwalbe, K.H.

1993-01-01

For monotonously increasing load the correct evaluation of the crack resistance properties of a structure is essential for safety analyses. Considerable attention has been given to the through-wall case, since this is generally believed to be the controlling case with regard to complete pipe failure. The maximum load conditions for circumferential crack growth in pipes under displacement-controlled loadings has been determined. The need for crack resistance curves, measured on circumferentially through-wall cracked straight pipes of austenitic stainless steel 316L under bending, is emphasized by the limitation in the data range on small specimens and by the differences in the procedures. To answer open questions and to improve calculational methods a joint fracture mechanics program is being performed by Electricite de France, Novatome and Siemens-Interatom. The working program contains experimental and theoretical investigations on the applicability of small-specimen data to real structures. 10 refs., 10 figs., 4 tabs

The evaluation of stress and piping support loads on RSG-GAS secondary cooling system

International Nuclear Information System (INIS)

Pustandyo, W.; Sitandung, Y. B.; Sujalmo, S.

1998-01-01

The evaluation of stress and piping support loads was evaluated on piping segment of secondary cooling water piping. In this paper, the analysis methods are presented with the use of computer code PS + CAEPIPE Version 3. 4. 05. W. From the selected pipe segment, the data of pipe characteristic, material properties, operation and design condition, equipment and support were used as inputs. The result of analysis show that stress and support loads if using location, kind and number of support equal with the system that have been installed for sustain load 3638 psi (node 160), thermal 13517 psi (node 90) and combination of sustain and thermal (node 90) 16747 psi. Meanwhile,if the optimization support, stress and support load for sustain load are respectively 4238 psi (node 10), thermal 13517 psi (node 90) and combination of sustain + thermal (node 90) 17350 psi. The limit values of permitted support based on Code PS+CAEPIPE of sustain load are 15000 psi, thermal 22500 psi and combination of sustain + thermal 37500 psi. The conclusion of evaluation result, that stress support load of pipe secondary cooling system are sufficiently low and using support show excessive and not economic

A multi-step approach for evaluation of pipe impact effects

International Nuclear Information System (INIS)

Vazquez Sierra, J.M.; Marti, J.

1987-01-01

The licensing of new and requalification of existing plant requires the consideration of effects arising from postulated breaks in high-energy lines. If the resulting jets or whipping pipes affect equipment or components (with safety-related functions in relation with the postulated break), their structural integrity and functionality has to be guaranteed. This can be achieved either by demonstrating sufficient ruggedness, or by obviating the problem with hardware (restraints, screens, deflectors, etc.). The present paper is orientated towards the first solution. A methodology has been developed and applied to the requalification of high-energy piping at the Santa Maria de Garona NPP in Spain. It provides techniques for evaluation of pipe-whip and jet effects on various structures inside the containment: containment liner, pedestal, shield wall, pipes and penetrations. Items of little structural strength (such as cables, conduits, etc.) were excluded from this approach for obvious reasons. (orig./GL)

Effective applied moment in circumferential through-wall cracked pipes for leak-before-break evaluation considering pipe restraint effects

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeji; Hwang, Il-Soon [Seoul National University, Seoul 08826 (Korea, Republic of); Oh, Young-Jin, E-mail: yjoh2@kepco-enc.com [KEPCO Engineering and Construction Co. Inc., Gimcheon 39660 (Korea, Republic of)

2016-05-15

Highlights: • Effective applied moment at pipe cracked section considering the pipe restraint effect. • Verification of the proposed evaluation methods using finite element analyses. • Applicability for distributed external load of the proposed methods. - Abstract: In the leak-before-break (LBB) design of nuclear power plants, crack opening displacement (COD) is an essential element for determining the length of the leakage size crack. Recent researches regarding the evaluation of COD have indicated that the current practice of the LBB evaluation without consideration of the pressure induced bending (PIB) restraint overestimates COD, which in turn gives non-conservative results. Under a free-ended boundary condition, however, the applied moment at cracked section also can be overestimated, which has conservative effects on LBB evaluation. Therefore, it is necessary to evaluate pipe restraint effects on the applied moment as well as on COD to keep the constancy. In this paper, an evaluation method for the effect of the PIB restraint on COD and an effective applied moment (=crack driving force) at cracked section was developed. Both the linear elastic and elastic–plastic behaviors of the crack were considered. By comparing the behaviors with 3-D finite element analysis results from earlier studies, it was confirmed that the proposed methods make accurate estimations of the PIB restraint effect on COD. Next, the applicability of the proposed method to other types of external loading conditions was examined.

Fracture propagation in gas pipelines - relevance to submarine lines

Energy Technology Data Exchange (ETDEWEB)

Fearnehough, G D [British Gas Corp., Newcastle upon Tyne. Engineering Research Station

1976-09-01

This paper reviews the factors which control fracture propagation in pipes and suggests how they are influenced by submarine environments. If fracture arrest capability is required then these factors should be considered in terms of the design philosophy and the maximum tolerable length of fracture which can be repaired. The paper shows that brittle fracture characteristics of submarine pipelines are probably similar to land based lines and fracture arrest can only be guaranteed by appropriate material toughness specification. Resistance to ductile fracture propagation in submarine lines is enhanced by lower design stresses, thicker pipe, concrete coating and the effect of hydrostatic head on gas dynamics. However, additional factors due to submarine design can be deleterious viz: uncertainty about backfill integrity and a tendency of thicker steels to low fracture resistance arising from 'separation' formation. Attention is drawn to problems which may arise with transportation of gases rich in hydrocarbons and the use of mechanical methods of fracture arrest.

Evaluation of stress corrosion crack growth in BWR piping systems

International Nuclear Information System (INIS)

Kassir, M.; Sharma, S.; Reich, M.; Chang, M.T.

1985-05-01

This report presents the results of a study conducted to evaluate the effects of stress intensity factor and environment on the growth behavior of intergranular stress corrosion cracks in type 304 stainless steel piping systems. Most of the detected cracks are known to be circumferential in shape, and initially started at the inside surface in the heat affected zone near girth welds. These cracks grow both radially in-depth and circumferentially in length and, in extreme cases, may cause leakage in the installation. The propagation of the crack is essentially due to the influence of the following simultaneous factors: (1) the action of applied and residual stress; (2) sensitization of the base metal in the heat affected zone adjacent to girth weld; and (3) the continuous exposure of the material to an aggressive environment of high temperature water containing dissolved oxygen and some levels of impurities. Each of these factors and their effects on the piping systems is discussed in detail in the report. The report also evaluates the time required for hypothetical cracks in BWR pipes to propagate to their critical size. The pertinent times are computed and displayed graphically. Finally, parametric study is performed in order to assess the relative influence and sensitivity of the various input parameters (residual stress, crack growth law, diameter of pipe, initial size of defect, etc.) which have bearing on the growth behavior of the intergranular stress corrosion cracks in type 304 stainless steel. Cracks in large-diameter as well as in small-diameter pipes are considered and analyzed. 27 refs., 25 figs., 10 tabs

A simplified leak-before-break evaluation procedure for austenitic and ferritic steel piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Ghassemi, B. [NOVETECH Corp., Rockville, MD (United States)

1994-10-01

A simplified procedure has been defined for computing the allowable circumferential throughwall crack length as a function of applied loads in piping. This procedure has been defined to enable leak-before-break (LBB) evaluations to be performed without complex and time consuming analyses. The development of the LBB evaluation procedure is similar to that now used in Section 11 of the ASME Code for evaluation of part-throughwall flaws found in piping. The LBB evaluation procedure was bench marked using experimental data obtained from pipes having circumferential throughwall flaws. Comparisons of the experimental and predicted load carrying capacities indicate that the method has a conservative bias, such that for at least 97% of the experiments the experimental load is equal to or greater than 90% of the predicted load. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austenitic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

Generic safety evaluation report regarding integrity of BWR scram system piping

International Nuclear Information System (INIS)

1981-08-01

Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

Test and evaluation about damping characteristics of hanger supports for nuclear power plant piping systems (Seismic Damping Ratio Evaluation Program)

International Nuclear Information System (INIS)

Shibata, H.; Ito, A.; Tanaka, K.; Niino, T.; Gotoh, N.

1981-01-01

Generally, damping phenomena of structures and equipments is caused by very complex energy dissipation. Especially, as piping systems are composed of many components, it is very difficult to evaluate damping characteristics of its system theoretically. On the other hand, the damping value for aseismic design of nuclear power plants is very important design factor to decide seismic response loads of structures, equipments and piping systems. The very extensive studies titled SDREP (Seismic Damping Ratio Evaluation Program) were performed to establish proper damping values for seismic design of piping as a joint work among a university, electric companies and plant makers. In SDREP, various systematic vibration tests were conducted to investigate factors which may contribute to damping characteristics of piping systems and to supplement the data of the pre-operating tests. This study is related to the component damping characteristics tests of that program. The object of this study is to clarify damping characteristics and mechanism of hanger supports used in piping systems, and to establish the evaluation technique of dispersing energy at hanger support points and its effect to the total damping ability of piping system. (orig./WL)

A leak-before-break strategy for CANDU primary piping systems

International Nuclear Information System (INIS)

Aggarwal, M.L.; Kozluk, M.J.; Lin, T.C.; Manning, B.W.; Vijay, D.K.

1986-01-01

Recent advances in elastic-plastic fracture mechanics have made it possible to assess the stability of cracks in ductile piping systems. These technological developments have been used by Ontario Hydro as the nucleus of an approach for demonstrating that CANDU primary heat transport piping systems will not break catastrophically; at worst they would leak at a detectable rate. This leak-before-break approach has been taken on the Darlington nuclear generating station as a design stage alternative to the provision of pipe whip restraints on large diameter, primary heat transport system piping. Positive conclusions reached via this approach are considered sufficient to exclude the requirement to provide protective devices, such as pipe whip restraints. In arriving at the proposed leak-before-break approach a review of current and proposed leak-before-break licensing positions of other jurisdictions (particularly those in the United States and the Federal Republic of Germany) was carried out. The approach presented makes use of recent American developments in the area of elastic-plastic fracture mechanics. It also gives consideration to aspects which are unique to the pressurized heavy water (CANDU) reactors used by Ontario Hydro. The proposed leak-before-break approach is described and its use is illustrated by applying it to the Darlington generating station primary heat transport system pump suction piping. (author)

Proposal of limit moment equation applicable to planar/non-planar flaw in wall thinned pipes under bending

International Nuclear Information System (INIS)

Tsuji, Masataka; Meshii, Toshiyuki

2011-01-01

Highlights: → A limit moment equation applicable to planar/non-planar flaw of 0 ≤ θ ≤ π found in wall thinned straight pipes was proposed. → An idea to rationally classify planar/non-planar flaw in wall thinned pipes was proposed. → The equation based on the experimental observation focused on the fracture mode. - Abstract: In this paper, a limit bending moment equation applicable to all types of planar and non-planar flaws in wall-thinned straight pipes under bending was proposed. A system to rationally classify the planar/non-planar flaws in wall-thinned pipes was suggested based on experimental observations focused on the fracture mode. The results demonstrate the importance of distinguishing between axial and circumferential long flaws in wall-thinned pipes.

Radiographic evaluation of maxillofacial fractures

International Nuclear Information System (INIS)

Litwan, M.; Fliegel, C.

1986-01-01

The course and configuration of typical maxillofacial fractures (type Le Fort I-III) and lateral maxillary fractures including the zygomatic arch were reconstructed in detail by application of barium paste on a bony skull and radiogrpahs in standard projections were performed and evaluated. It was obvious from the resulting radiographs that for most maxillofacial fractures a half axial or Water's view was most helpful. Lateral views only give additional information when there is a considerable degree of dislocation of fragments. Comparison with a prediatric skull of 8 years of age demonstrated that fractures of the zygomatic arch in this age group cannot be demonstrated by the typical submento-vertical view, but are shown on a Towne projection. The radiographic appearance of important maxillofacial fractures is demonstrated. The necessity of further studies in cases where reconstructive surgery appears necessary is discussed and CT rather then conventional tomography is advocated. (orig.) [de

Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

International Nuclear Information System (INIS)

Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

2014-01-01

This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

Development of a Short-term Failure Assessment of High Density Polyethylene Pipe Welds - Application of the Limit Load Analysis -

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho-Wan; Han, Jae-Jun; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of); Kim, Jong-Sung [Sunchon National University, Suncheon (Korea, Republic of); Kim, Jeong-Hyeon; Jang, Chang-Heui [KAIST, Daejeon (Korea, Republic of)

2015-04-15

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

Evaluation of 3 Inch SN-219 Failure and S and SX Tank Farm Saltwell Piping

International Nuclear Information System (INIS)

ELSEN, J.J.

2000-01-01

Evaluation of direct buried piping currently in use or designated for future Saltwell pumping in S and SX Farms. Documented evaluation of failed S-103 saltwell pumping transfer line 3 inch SN-219. This evaluation is intended to reflect current status of Saltwell piping, when taken in context with referenced documents

Evaluation of residual stress on pipe welded joints using laser interferometry

Energy Technology Data Exchange (ETDEWEB)

Chang, Ho Seob; Na, Man Gyun; Kim, Koung Suk [Chosun University, Gwangju (Korea, Republic of)

2014-02-15

Residual stresses that occur during the welding process, are the main cause of failure and defects in welded structures. This paper, presents the use of an electronic processing laser speckle interferometer to measure the residual stress of a welded pipe for a nuclear power plant. A tensile testing machine was used to evaluate a welded pipe that failed in compression. The inform plane deformation and modulus of elasticity of the base metal and welds were measured using an interferometer. Varying the load on the welded pipe had a larger effect on the deformation of the base metal the other properties of the base metal and welds. The elastic moduli of the base metal and weld of the welded pipe were 202.46 and 212.14 GPa, respectively, the residual stress was measured to be 6.29 MPa.

Dynamic fracture toughness and evaluation of fracture in a ferritic nodular cast iron for casks

International Nuclear Information System (INIS)

Yasunaka, T.; Nakano, K.

1993-01-01

The effect of loading rate and temperature on fracture toughness of a ferritic nodular cast iron obtained from a thick-walled cylindrical casting has been investigated. Based upon this result, the cast iron is evaluated as a material for casks. (1) In the ductile fracture region, fracture toughness increases with increases in loading rate. (2) Ductile-brittle transition temperature is linearly related to the logarithm of stress intensity rate. (3) In the ductile fracture region, converted plain strain fracture toughness divided by yield stress can be adopted as a material constant which is independent of loading rate and temperature. From the result of a static fracture toughness test, the evaluation of fracture in high loading rate can be made. (4) In the ductile fracture region of the material investigated, the maximum allowable flaw depth exceeded the minimum detectable flaw size by a nondestructive inspection. Ferritic nodular cast iron can be used as a material for casks in the ductile fracture region at least. (J.P.N.)

Experimental evaluation of J in cracked straight and curved pipes under bending

International Nuclear Information System (INIS)

Moulin, D.; Touboul, F.; Foucher, N.; Lebey, J.; Acker, D.

1989-01-01

An experimental program is being carried out at the CEA Saclay in collaboration with FRAMATOME and IPSN with a view to validate analysis methods applicable for evaluation of leak before break behavior in P.W.R. piping. A large experimental work was already performed in USA, Germany and Japan and cracked pipes made of stainless steel material under bending. The methods of analysis got same validations for straight pipes. However applicability to elbows and comparison with toughness values obtained on small specimens like CT specimens was not completely dealt with

Performance evaluation of buried pipe installation.

Science.gov (United States)

2010-05-01

The purpose of this study is to determine the effects of geometric and mechanical parameters characterizing the soil structure interaction developed in a buried pipe installation located under roads/highways. The drainage pipes or culverts installed ...

Fatigue evaluation of piping connections under thermal transients

International Nuclear Information System (INIS)

Aquino, C.T.E. de; Maneschy, J.E.

1993-01-01

In designing nuclear power plant piping, thermal transients, caused by non-steady operation conditions, should be considered. These events may reduce considerably the lifetime of the pipes, creating the necessity of using structural elements designed in such a way to minimize the acting thermal stresses. Typical examples of the usage of these elements are the connections between pipes of small and large diameters, in which it is usually used a weldolet. Nevertheless, in some situations, the thermal stresses caused by the transients are greater than the allowable limits, being, in this case, an alternative for best results, the introduction of a special fitting replacing the weldolet. Such a fitting is designed in a way to permit a better distribution of the stresses, reducing its maximum value to acceptable levels. This paper intends to present a fatigue evaluation of a connection, using the above mentioned fitting, when subjected to a load expressed in terms of a step thermal gradient, varying from 263 deg to 40 deg C. Two different methodologies are used in this analysis: (a) Determination of the temperature distribution from the heat transfer equations for piping, being the stresses calculated according to ASME III NB-3600. (b) Thermal and stress analyses using axisymmetric elements, according to the rules presented at ASME III NB-3200. In the first case, named simplified analysis, the computer code used is the PIPESTRESS, while in the second case, the ANSYS program was adopted

An evaluation of an operating BWR piping system damping during earthquake by applying auto regressive analysis

International Nuclear Information System (INIS)

Kitada, Y.; Makiguchi, M.; Komori, A.; Ichiki, T.

1985-01-01

The records of three earthquakes which had induced significant earthquake response to the piping system were obtained with the earthquake observation system. In the present paper, first, the eigenvalue analysis results for the natural piping system based on the piping support (boundary) conditions are described and second, the frequency and the damping factor evaluation results for each vibrational mode are described. In the present study, the Auto Regressive (AR) analysis method is used in the evaluation of natural frequencies and damping factors. The AR analysis applied here has a capability of direct evaluation of natural frequencies and damping factors from earthquake records observed on a piping system without any information on the input motions to the system. (orig./HP)

Fabrication and performance evaluation of flexible heat pipes for potential thermal control of foldable electronics

International Nuclear Information System (INIS)

Yang, Chao; Chang, Chao; Song, Chengyi; Shang, Wen; Wu, Jianbo; Tao, Peng; Deng, Tao

2016-01-01

Highlights: • A flexible and high-performance heat pipe is fabricated. • Bending effect on thermal performance of flexible heat pipes is evaluated. • Theoretical analysis is carried out to reveal the change of thermal resistance with bending. • Thermal control of foldable electronics with flexible heat pipes is demonstrated. - Abstract: In this work, we report the fabrication and thermal performance evaluation of flexible heat pipes prepared by using a fluororubber tube as the connector in the adiabatic section and using strong base treated hydrophilic copper meshes as the wick structure. Deionized water was chosen as working fluid and three different filling ratios (10%, 20%, and 30%) of working fluid were loaded into the heat pipe to investigate its impact on thermal performance. The fabricated heat pipes can be easily bended from 0"o to 180"o in the horizontal operation mode and demonstrated consistently low thermal resistances after repeated bending. It was found that with optimized amount of working fluid, the thermal resistance of flexible heat pipes increased with larger bending angles. Theoretical analysis reveals that bending disturbs the normal vapor flow from evaporator to condenser in the heat pipe, thus leads to increased liquid–vapor interfacial thermal resistance in the evaporator section. The flexible heat pipes have been successfully applied for thermal control of foldable electronic devices showing superior uniform heat-transfer performance.

Numerical evaluation of cracked pipes under dynamic loading

International Nuclear Information System (INIS)

Petit, M.; Jamet, P.

1989-01-01

In order to apply the leak-before-break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic, loadings must be studied. A simple finite element model of a cracked pipe has been developed and implemented in the general purpose computer code CASTEM 2000. The model is a generalization of the approach proposed by Paris and Tada (1). Considered loads are bending moment and axial force (representing thermal expansion and internal pressure.) The elastic characteristics of the model are determined using the Zahoor formulae for the geometry-dependent factors. Owing to the material behabior plasticity must be taken into account. To represent the crack growth, the material is defined by two characteristic values: J 1c which is the level of energy corresponding to crack initiation and the tearing modulus, T, which governs the length of propagation of the crack. For dynamic loads, unilateral conditions are imposed to represent crack closure. The model has been used for the design of dynamic tests to be conducted on shaking tables. Test principle is briefly described and numerical results are presented. Finally evaluation of margin, due to plasticity, in comparison with the standard design procedure is made

Evaluation of local allowable wall thickness of thinned pipe considering internal pressure and bending moment

International Nuclear Information System (INIS)

Kim, J. W.; Park, C. Y.; Kim, B. Y.

2000-01-01

This study proposed the local allowable wall thickness (LAWT) evaluation method for local wall thinned pipe subjected by internal pressure and bending moment. Also, LAWT was evaluated for simplified thinned pipe and the effect of axial extent of thinned area on LAWT was investigated. The results showed that LAWT predicted by present method was thinner, about 50%, than that evaluated by construction code and ASME Code Case N-597, while it was thicker, about 2 times, than that calculated by evaluation model based on pipe experiments. LAWT decreased with increasing axial extent of thinned area and was saturated above axial extent of pipe radius, which was a contrast to the results of ASME Code Case N-597 evaluation. The results of stress analysis with applied loading type indicated that the effect of axial extent of thinned area on LAWT was dependent on loading type considering in the evaluation. That is, the dependence of axial extent on LAWT is determined by magnitude of bending moment, and the contrary trend with axial extent in ASME Code Case is because ASME Code Case N-597 considers only internal pressure in the evaluation

Fatigue evaluation of socket welded piping in nuclear power plant

International Nuclear Information System (INIS)

Vecchio, R.S.

1996-01-01

Fatigue failures in piping systems occur, almost without exception, at the welded connections. In nuclear power plant systems, such failures occur predominantly at the socket welds of small diameter piping ad fillet attachment welds under high-cycle vibratory conditions. Nearly all socket weld fatigue failures are identified by leaks which, though not high in volume, generally are costly due to attendant radiological contamination. Such fatigue cracking was recently identified in the 3/4 in. diameter recirculation and relief piping socket welds from the reactor coolant system (RCS) charging pumps at a nuclear power plant. Consequently, a fatigue evaluation was performed to determine the cause of cracking and provide an acceptable repair. Socket weld fatigue life was evaluated using S-N type fatigue life curves for welded structures developed by AASHTO and the assessment of an effective cyclic stress range adjacent to each socket weld. Based on the calculated effective tress ranges and assignment of the socket weld details to the appropriate AASHTO S-N curves, the socket weld fatigue lives were calculated and found to be in excellent agreement with the accumulated cyclic life to-date

Influence of structure on static cracking resistance and fracture of welded joints of pipe steels of strength class K60

Science.gov (United States)

Tereshchenko, N. A.; Tabatchikova, T. I.; Yakovleva, I. L.; Makovetskii, A. N.; Shander, S. V.

2017-07-01

The static cracking resistance of a number of welded joints made from pipe steels of K60 strength class has been determined. It has been established that the deformation parameter CTOD varies significantly at identical parameters of weldability of steels. The character of fracture has been investigated and the zone of local brittleness of welded joints has been studied. It has been shown that the ability of a metal to resist cracking is determined by the austenite grain size and by the bainite morphology in the region of overheating in the heat-affected zone of a welded joint.

Multilayer polymer pipes failure assessment based on a fracture mechanics approach

Czech Academy of Sciences Publication Activity Database

Hutař, Pavel; Zouhar, Michal; Náhlík, Luboš; Ševčík, Martin; Máša, Bohuslav

2013-01-01

Roč. 33, OCT (2013), s. 151-162 ISSN 1350-6307 R&D Projects: GA ČR(CZ) GAP108/12/1560 Institutional support: RVO:68081723 Keywords : Multilayer pipes * Generalised stress intensity factor * Material interface * Slow crack growth * Polyolefin pipes Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 1.130, year: 2013

Rethinking ASME III seismic analysis for piping operability evaluations

International Nuclear Information System (INIS)

Adams, T.M.; Stevenson, J.D.

1994-01-01

It has been recognized since the mid 1980's that there are very large seismic margins to failure for nuclear piping systems when designed using current industry practice, design criteria, and methods. As a result of this realization there are or have been approximately eighteen initiatives within the ASME , Boiler and Pressure Vessel Code Section III, Division 1, in the form of proposed code cases and proposed code text changes designed to reduce these failure margins to more realistic values. For the most part these initiatives have concentrated on reclassifying seismic inertia stresses in the piping as secondary and increasing the allowable stress limits permitted by Section III of the ASME, Boiler Code. This paper focuses on the application of non-linear spectral analysis methods as a method to reduce the input seismic demand determination and thereby reduce the seismic failure margins. The approach is evaluated using the ASME Boiler Pressure Vessel Code Section III Subgroup on Design benchmark procedure as proposed by the Subgroup's Special Task Group on Integrated Piping Criteria. Using this procedure, criteria are compared to current code criterion and analysis methods, and several other of the currently proposed Boiler and Pressure Vessel, Section III, changes. Finally, the applicability of the non-linear spectral analysis to continued Safe Operation Evaluations is reviewed and discussed

Ductile fracture mechanics methodology for complex cracks in nuclear piping

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1988-02-01

Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions.

Ductile fracture mechanics methodology for complex cracks in nuclear piping

International Nuclear Information System (INIS)

Zahoor, A.

1988-01-01

Limit load and J-integral estimation solutions are developed for circumferentially complex-cracked pipes in bending. The limit load solution is developed using thick-walled cylinder analysis which included the effects of flaw depth accurately. J-integral estimation solutions are developed that are suitable for a wide range of loading from linear elastic, elastic-plastic to net-section yielding of the flawed section. Mode I stress intensity factor solution is developed from experimental compliance data. Two types of J solutions are developed. First, J solutions for determining the J-resistance curve from single load-displacement record are presented. Next, elastic-plastic J solution in the format of EPRI J estimation scheme is presented. The latter solution was used to predict the load carrying capacity of complex-cracked pipes made of Type-304 stainless steel, Inconel 600, and A106 GrB materials. Predictions were compared against pipe tests to demonstrate the accuracy of the limit load and J estimation solutions. (orig.)

Reactive Transport in a Pipe in Soluble Rock: a Theoretical and Experimental Study

Science.gov (United States)

Li, W.; Opolot, M.; Sousa, R.; Einstein, H. H.

2015-12-01

Reactive transport processes within the dominant underground flow pathways such as fractures can lead to the widening or narrowing of rock fractures, potentially altering the flow and transport processes in the fractures. A flow-through experiment was designed to study the reactive transport process in a pipe in soluble rock to serve as a simplified representation of a fracture in soluble rock. Assumptions were made to formulate the problem as three coupled, one-dimensional partial differential equations: one for the flow, one for the transport and one for the radius change due to dissolution. Analytical and numerical solutions were developed to predict the effluent concentration and the change in pipe radius. The positive feedback of the radius increase is captured by the experiment and the numerical model. A comparison between the experiment and the simulation results demonstrates the validity of the analytical and numerical models.

Application of leak-before-break to primary loop piping to eliminate pipe whip restraints in a Spanish nuclear power plant

International Nuclear Information System (INIS)

Rodriguez, M.; Esteban, A.

1990-01-01

The Spanish plant described in this study is a 982 MWe PWR with a three-loop primary circuit of piping made from centrifugally-cast stainless steel SA351 CF8A. The licensee requested from Consejo de Seguridad Nuclear (CSN) an exemption from the general design criterion, GDC-4, so as to avoid the need to postulate a guillotine rupture of the primary loop piping. The request was based on the generic work performed for a US PWR plant group in order to have such an exemption. As the piping material in the Spanish plant is different from that in the plants included in the generic work, CSN performed a review of the applicability of the generic results to the Spanish plant. Also, aspects such as fatigue evaluation, net section collapse, crack growth and leak detection, specifically analyzed for the Spanish plant, were reviewed. CSN found that fracture toughness test results from generic work are applicable to the Spanish plant; sufficient margin exists against unstable crack extension, and adequate leak detection capability exists with the leakage detection systems available in the plant. Exemption from GDC-4 was approved and CSN authorized the licensee to remove protection devices against dynamic loads from guillotine breaks in the primary coolant loops. (author)

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1986-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

Refinement and evaluation of crack-opening-area analyses for circumferential through-wall cracks in pipes

International Nuclear Information System (INIS)

Rahman, S.; Brust, F.; Ghadiali, N.; Krishnaswamy, P.; Wilkowski, G.; Choi, Y.H.; Moberg, F.; Brickstad, B.

1995-04-01

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet impingement shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. These leak rates depend on the crack-opening area of a through-wall crack in the pipe. In addition to LBB analyses, which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section 11. This study was requested by the NRC to review, evaluate, and refine current analytical models for crack-opening-area analyses of pipes with circumferential through-wall cracks. Twenty-five pipe experiments were analyzed to determine the accuracy of the predictive models. Several practical aspects of crack-opening such as; crack-face pressure, off-center cracks, restraint of pressure-induced bending, cracks in thickness transition regions, weld residual stresses, crack-morphology models, and thermal-hydraulic analysis, were also investigated. 140 refs., 105 figs., 41 tabs

Guidelines and criteria for nuclear piping and support evaluation and design

International Nuclear Information System (INIS)

Rehn, D.L.; Stout, D.H. Jr.; Minichiello, J.C.

1993-05-01

The EPRI Research Project 2967-2 has set its fundamental goal to be the development of realistic guidelines and criteria for piping and pipe support design and evaluation. The focus is on items that are most critical to utilities and consists of a variety of tasks relating to piping and pipe support design. One objective of this report is to summarize the recommendations from the seven task reports of the first phase of the project and to provide examples of how to use those recommendations. Criteria and methods for evaluating both short and long term system operation are addressed. Benefits gained from applying the recommendations to actual systems are discussed. The report also reviews other work currently being done within the nuclear industry and assesses the impact of that work on the recommended criteria/methods of this project. The second objective of the report is to discuss possible changes needed in the governing codes or licensing commitments in order to implement the recommendations. Finally, the report describes further research which can be done to advance the criteria presented and answer questions concerning applicability of the proposed criteria to designs not tested/investigated. The basic conclusion reached in the project is that many of the criteria/methods used today in piping analysis/design are overly conservative. The report's conclusion is supported by extensive literature searches, tests, and analyses. The report presents a robust source of reference to utilities which wish to implement changes in criteria and methods. Most of the criteria and methodologies described in the seven task reports and summarized in the following sections will require some effort in licensing or Code changes

Finite element limit analysis based plastic limit pressure solutions for cracked pipes

International Nuclear Information System (INIS)

Shim, Do Jun; Huh, Nam Su; Kim, Yun Jae; Kim, Young Jin

2002-01-01

Based on detailed FE limit analyses, the present paper provides tractable approximations for plastic limit pressure solutions for axial through-wall cracked pipe; axial (inner) surface cracked pipe; circumferential through-wall cracked pipe; and circumferential (inner) surface cracked pipe. Comparisons with existing analytical and empirical solutions show a large discrepancy in circumferential short through-wall cracks and in surface cracks (both axial and circumferential). Being based on detailed 3-D FE limit analysis, the present solutions are believed to be the most accurate, and thus to be valuable information not only for plastic collapse analysis of pressurised piping but also for estimating non-linear fracture mechanics parameters based on the reference stress approach

Dosimetric evaluation of multi-sided irradiation on HDPE pipes under 2 MeV electron beam

Energy Technology Data Exchange (ETDEWEB)

Benny, P.G., E-mail: bennypg@yahoo.com; Khader, S.A.; Sarma, K.S.S.

2014-03-01

The use of electron beam technology has enabled the production of heat resistant pipe for hot water circulation. One of the difficulties in the irradiation of pipe products is the uneven penetration of electrons. Quality of the radiation process depends on radiation dose and homogeneity of the dose distribution, which becomes a major concern when treatments of circular objects like pipes are performed. One method to achieve uniformity in the absorbed dose in the product is to use multi-sided irradiation. The paper discusses the importance of dosimetry mapping in industrial electron beam radiation processing and outlines the challenges in delivering a uniform dose to cylindrical objects. In this study, HDPE pipe of 5 mm thickness of homogeneous material (40 mm outer diameter and 30 mm inner diameter) has been chosen for multi-sided irradiation under 2 MeV scanned electron beam from the ILU-6 accelerator. - Highlights: • The paper outlines the challenges in delivering uniform dose to cylindrical objects at 2 MeV industrial electron beam facility. • HDPE pipe of 40 mm outer diameter and 30 mm inner diameter has been chosen for the study. • The circumferential dose distribution inside and outside of the pipes were evaluated by using calibrated CTA dosimeter strips. • Using stack of dosimeter strips, changes in circumferential dose distribution in the annular region of the pipe was evaluated. • Optimization of multi-sided irradiation on the HDPE pipes for better dose homogeneity is reported in the paper.

Dosimetric evaluation of multi-sided irradiation on HDPE pipes under 2 MeV electron beam

International Nuclear Information System (INIS)

Benny, P.G.; Khader, S.A.; Sarma, K.S.S.

2014-01-01

The use of electron beam technology has enabled the production of heat resistant pipe for hot water circulation. One of the difficulties in the irradiation of pipe products is the uneven penetration of electrons. Quality of the radiation process depends on radiation dose and homogeneity of the dose distribution, which becomes a major concern when treatments of circular objects like pipes are performed. One method to achieve uniformity in the absorbed dose in the product is to use multi-sided irradiation. The paper discusses the importance of dosimetry mapping in industrial electron beam radiation processing and outlines the challenges in delivering a uniform dose to cylindrical objects. In this study, HDPE pipe of 5 mm thickness of homogeneous material (40 mm outer diameter and 30 mm inner diameter) has been chosen for multi-sided irradiation under 2 MeV scanned electron beam from the ILU-6 accelerator. - Highlights: • The paper outlines the challenges in delivering uniform dose to cylindrical objects at 2 MeV industrial electron beam facility. • HDPE pipe of 40 mm outer diameter and 30 mm inner diameter has been chosen for the study. • The circumferential dose distribution inside and outside of the pipes were evaluated by using calibrated CTA dosimeter strips. • Using stack of dosimeter strips, changes in circumferential dose distribution in the annular region of the pipe was evaluated. • Optimization of multi-sided irradiation on the HDPE pipes for better dose homogeneity is reported in the paper

Evaluation of temporary non-code repairs in safety class 3 piping systems

International Nuclear Information System (INIS)

Godha, P.C.; Kupinski, M.; Azevedo, N.F.

1996-01-01

Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

The stability of through-wall circumferential cracks in cylindrical pipes subjected to bending loads

International Nuclear Information System (INIS)

Smith, E.

1983-01-01

Tada, Paris and Gamble have used the tearing modulus approach to show that when a circumferential through-wall crack exists in a 304 SS circular cylindrical pipe, and the pipe is subjected to an applied bending moment, then crack growth requires the rotation at the pipe-ends to be increased, (i.e. crack growth is stable), unless the pipe length is unduly large. On this basis it was concluded that unstable fracture is unlikely to occur in BWR SS piping, when the system is designed in accord with the ASME Code load levels for normal operation and anticipated transients. The Tada-Paris-Gamble analysis focuses on the inter-relation between instability and the onset of crack extension, and does not specifically consider the possibility that a crack might become unstable after some stable crack extension. The paper addresses this aspect of the crack stability problem using a crack tip opening angle criterion for crack extension, which has similarities with the tearing modulus approach. The results show that unstable fracture should not occur even after some stable crack extension, again provided that the pipe length is not unduly large. In other words, guillotine failure of a pipe in a BWR system is unlikely, even though the ASME Code limiting stress levels as might be exceeded, as may be the case with a very severe earthquake. (orig./HP)

Study on calculation of influence function with fracture mechanics analysis method for circumferential through-crack pipe

International Nuclear Information System (INIS)

Zheng Bin; Lu Yuechuan; Zang Fenggang; Sun Yingxue

2009-01-01

In order to widen the application of the engineering method of EPRI, with a series of analysis on the 3D elastic and elastic-plastic fracture mechanics finite element, the crack open displacements (COD) of cracked pipe were calculated and a key influence function h 2 in EPRI engineering method was studied against the COD results of FEM. A calculation method of h2 under the condition of tension and bending combined load was introduced in detail. In order to validate this method, the calculated h 2 results were compared with that of EPRI, and the calculated COD results based on the h 2 results were compared with that of PICEP. The compared results indicated that the calculated h 2 results as well as the COD results and the corresponding reference values were respectively accordant, and the calculation method in this paper was validated accordingly. (authors)

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

Analysis and Optimisation of Carcass Production for Flexible Pipes

DEFF Research Database (Denmark)

Nielsen, Peter Søe

Un-bonded flexible pipes are used in the offshore oil and gas industry worldwide transporting hydrocarbons from seafloor to floating production vessels topside. Flexible pipes are advantageous over rigid pipelines in dynamic applications and during installation as they are delivered in full length......-axial tension FLC points were attained. Analysis of weld fracture of duplex stainless steel EN 1.4162 is carried out determining strains with GOM ARAMIS automated strain measurement system, which shows that strain increases faster in the weld zone than the global strain of the parent material. Fracture...... is the analysis and optimisation of the carcass manufacturing process by means of a fundamental investigation in the fields of formability, failure modes / mechanisms, Finite Element Analysis (FEA), simulative testing and tribology. A study of failure mechanisms in carcass production is performed by being present...

Technical report on the Piping Reliability Proving Tests at the Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1993-05-01

Japan Atomic Energy Research Institute (JAERI) conducts Piping Reliability Proving Tests from 1975 to 1992 based upon the contracts between JAERI and Science and Technology Agency of Japan (STA) under the auspices of the special account law for electric power development promotion. The purpose of these tests are to prove the structural reliability of the primary cooling piping constituting a part of the pressure boundary in the light water reactor power plants. The tests with large experimental facilities had ended already in 1990. Presently piping reliability analysis by the probabilistic fracture mechanics method is being done. Until now annual reports concerning the proving tests were produced and submitted to STA, whereas this report summarizes the test results done during these 16 years. Objectives of the piping reliability proving tests are to prove that the primary piping of the light water reactor (1) be reliable throughout the service period, (2) have no possibility of rupture, (3) bring no detrimental influence on the surrounding instrumentations or equipments near the break location even if it ruptured suddenly. To attain these objectives (i) pipe fatigue tests, (ii) unstable pipe fracture tests, (iii) pipe rupture tests and also the analyses by computer codes were done. After carrying out these tests, it is verified that the piping is reliable throughout the service period. The authors of this report are T. Isozaki, K. Shibata, S. Ueda, R. Kurihara, K. Onizawa and A. Kohsaka. The parts they wrote are shown in contents. (author)

FROMS3D: New Software for 3-D Visualization of Fracture Network System in Fractured Rock Masses

Science.gov (United States)

Noh, Y. H.; Um, J. G.; Choi, Y.

2014-12-01

A new software (FROMS3D) is presented to visualize fracture network system in 3-D. The software consists of several modules that play roles in management of borehole and field fracture data, fracture network modelling, visualization of fracture geometry in 3-D and calculation and visualization of intersections and equivalent pipes between fractures. Intel Parallel Studio XE 2013, Visual Studio.NET 2010 and the open source VTK library were utilized as development tools to efficiently implement the modules and the graphical user interface of the software. The results have suggested that the developed software is effective in visualizing 3-D fracture network system, and can provide useful information to tackle the engineering geological problems related to strength, deformability and hydraulic behaviors of the fractured rock masses.

Low evaluation rate for osteoporosis among patients presenting with a rib fracture.

Science.gov (United States)

Kim, Whang; Gong, Hyun Sik; Lee, Seung Hoo; Park, Jin Woo; Kim, Kahyun; Baek, Goo Hyun

2017-12-01

This study in a regional hospital setting found a low evaluation rate for osteoporosis among patients presenting with a rib fracture. Increased emphasis or education for osteoporosis evaluation may be necessary in case of rib fractures. Rib fractures from a low-energy trauma are common in the elderly, and a history of rib fracture has been reported to increase the risk for a subsequent osteoporotic fracture. The purpose of this study was to evaluate how many of the patients presenting with an isolated rib fracture were being evaluated for osteoporosis and the risk for a subsequent fracture. We retrospectively reviewed all patients aged 50 years or older who were diagnosed with a rib fracture between January 2011 and April 2016 at a regional tertiary care university hospital near Seoul, South Korea. We excluded those who had been treated for osteoporosis or those with other concomitant fractures or fractures from a motor vehicle accident or cancer. We evaluated the frequency of dual energy X-ray absorptiometry (DXA) scan examinations in these patients. There were 231 patients with isolated rib fractures (132 men and 99 women). The mean age was 65 years. Rib fractures were most commonly diagnosed at the emergency department and most of the patients were referred to the department of thoracic surgery for follow-up evaluations. Of these 231 patients, 29 (12%) had DXA examinations after the injury, and only 9 (4%) of them did so within 6 months. Physicians specializing in orthopedic surgery, family medicine, internal medicine, rehabilitation medicine, and emergency medicine were ordering the examination. This study in a regional hospital setting found a low evaluation rate for osteoporosis among patients presenting with a rib fracture. This study suggests that increased emphasis or education for osteoporosis evaluation may be necessary for physicians who are often referred to for care of rib fractures.

Program to justify life extension of older nuclear piping systems

International Nuclear Information System (INIS)

Burr, T.K.; Dwight, J.E. Jr.; Morton, D.K.

1991-01-01

The Idaho National Engineering Laboratory (INEL) has a history of more than 40 years devoted to the operation of nuclear reactors designed for research and experiments. The Advanced Test Reactor (ATR) is one such operating reactor whose mission requires continued operation for an additional 25 years or more. Since the ATR is approaching its design life of twenty years, life extension evaluations have been initiated. Of particular importance are the associated high temperature, high pressure loop piping system supporting in--reactor experiments. Failure of this piping could challenge core safety margins. Since regulatory rules for nuclear power plant life extension are only in the formulation stage, the current technical guidance on this subject provided by the Department of Energy (DOE) or the commercial nuclear industry is incomplete. In the interim, order to assure continued safe operation of this piping beyond its initial design life, a program has been developed to provide the necessary technical justification for life extension. This paper describes a program that establishes Section 11 of the ASME Boiler and Pressure Vessel Code as the governing criteria document, retains B31.1 as the Code of record for Section 11 activities, specifies additional inservice inspection requirements more strict than Section 11, and relies heavily on flaw detection and fracture mechanics evaluations. 18 refs., 2 figs

Risk analysis of in-service pressure piping containing defects

International Nuclear Information System (INIS)

Lin, Y.C.; Xie, Y.J.; Wang, X.H.; Luo, H.

2004-01-01

The reliability of pressure piping containing defects is important in engineering. The failure probability of pressure piping containing defects may be used as a guide to the most economic deployment of resources on maintenance, inspection and repair. This paper presents a probabilistic assessment methodology for in-service pressure piping containing defects, which is especially designed for programming. It is based on three assessment codes, BS 7910, R6 and SAPV-99, considering uncertainties in operating loadings, flaw sizes, material fracture toughness and flow stress. A general sampling computation method of stress intensity factor (SIF), in the form of the relationship between SIF and axial force and bending moment and torsion, is adopted. This relationship has been successfully used in developing software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), to assess planar and non-planar flaws. A numerical example is presented to illustrate the application of SAPP-2003 for calculating the failure probabilities of separate defects and for the assessed pressure piping

Evaluation of permeable fractures in rock aquifers

Science.gov (United States)

Bok Lee, Hang

2015-04-01

In this study, the practical usefulness and fundamental applicability of a self-potential (SP) method for identifying the permeable fractures were evaluated by a comparison of SP methods with other geophysical logging methods and hydraulic tests. At a 10 m-shallow borehole in the study site, the candidates of permeable fractures crossing the borehole were first determined by conventional geophysical methods such as an acoustic borehole televiwer, temperature, electrical conductivity and gamma-gamma loggings, which was compared to the analysis by the SP method. Constant pressure injection and recovery tests were conducted for verification of the hydraulic properties of the fractures identified by various logging methods. The acoustic borehole televiwer and gamma-gamma loggings detected the open space or weathering zone within the borehole, but they cannot prove the possibility of a groundwater flow through the detected fractures. The temperature and electrical conductivity loggings had limitations to detect the fractured zones where groundwater in the borehole flows out to the surrounding rock aquifers. Comparison of results from different methods showed that there is a best correlation between the distribution of hydraulic conductivity and the variation of the SP signals, and the SP logging can estimate accurately the hydraulic activity as well as the location of permeable fractures. Based on the results, the SP method is recommended for determining the hydraulically-active fractures rather than other conventional geophysical loggings. This self-potential method can be effectively applied in the initial stage of a site investigation which selects the optimal location and evaluates the hydrogeological property of fractures in target sites for the underground structure including the geothermal reservoir and radioactive waste disposal.

A Study on an Evaluation of PWR Piping Thermal Stratification

Energy Technology Data Exchange (ETDEWEB)

Lee, H.; Kim, B.N.; Lee, S.K.; Jeong, I.S.; Chjung, B.S.; Lee, S.H. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

1997-12-31

This report presents the determination of thermal stratification phenomenon of pressurizer surge line for Kori unit No.4. With this regards, the integrity of related piping was evaluated by both various stress analysis and fatigue analysis. (author). 23 refs., 61 figs., 22 tabs.

Analysis Strategy for Fracture Assessment of Defects in Ductile Materials

Energy Technology Data Exchange (ETDEWEB)

Dillstroem, Peter; Andersson, Magnus; Sattari-Far, Iradj; Weilin Zang (Inspecta Technology AB, Stockholm (Sweden))

2009-06-15

The main purpose of this work is to investigate the significance of the residual stresses for defects (cracks) in ductile materials with nuclear applications, when the applied primary (mechanical) loads are high. The treatment of weld-induced stresses as expressed in the SACC/ProSACC handbook and other fracture assessment procedures such as the ASME XI code and the R6-method is believed to be conservative for ductile materials. This is because of the general approach not to account for the improved fracture resistance caused by ductile tearing. Furthermore, there is experimental evidence that the contribution of residual stresses to fracture diminishes as the degree of yielding increases to a high level. However, neglecting weld-induced stresses in general, though, is doubtful for loads that are mostly secondary (e.g. thermal shocks) and for materials which are not ductile enough to be limit load controlled. Both thin-walled and thick-walled pipes containing surface cracks are studied here. This is done by calculating the relative contribution from the weld residual stresses to CTOD and the J-integral. Both circumferential and axial cracks are analysed. Three different crack geometries are studied here by using the finite element method (FEM). (i) 2D axisymmetric modelling of a V-joint weld in a thin-walled pipe. (ii) 2D axisymmetric modelling of a V-joint weld in a thick-walled pipe. (iii) 3D modelling of a X-joint weld in a thick-walled pipe. t. Each crack configuration is analysed for two load cases; (1) Only primary (mechanical) loading is applied to the model, (2) Both secondary stresses and primary loading are applied to the model. Also presented in this report are some published experimental investigations conducted on cracked components of ductile materials subjected to both primary and secondary stresses. Based on the outcome of this study, an analysis strategy for fracture assessment of defects in ductile materials of nuclear components is proposed. A new

Analysis Strategy for Fracture Assessment of Defects in Ductile Materials

International Nuclear Information System (INIS)

Dillstroem, Peter; Andersson, Magnus; Sattari-Far, Iradj; Weilin Zang

2009-06-01

The main purpose of this work is to investigate the significance of the residual stresses for defects (cracks) in ductile materials with nuclear applications, when the applied primary (mechanical) loads are high. The treatment of weld-induced stresses as expressed in the SACC/ProSACC handbook and other fracture assessment procedures such as the ASME XI code and the R6-method is believed to be conservative for ductile materials. This is because of the general approach not to account for the improved fracture resistance caused by ductile tearing. Furthermore, there is experimental evidence that the contribution of residual stresses to fracture diminishes as the degree of yielding increases to a high level. However, neglecting weld-induced stresses in general, though, is doubtful for loads that are mostly secondary (e.g. thermal shocks) and for materials which are not ductile enough to be limit load controlled. Both thin-walled and thick-walled pipes containing surface cracks are studied here. This is done by calculating the relative contribution from the weld residual stresses to CTOD and the J-integral. Both circumferential and axial cracks are analysed. Three different crack geometries are studied here by using the finite element method (FEM). (i) 2D axisymmetric modelling of a V-joint weld in a thin-walled pipe. (ii) 2D axisymmetric modelling of a V-joint weld in a thick-walled pipe. (iii) 3D modelling of a X-joint weld in a thick-walled pipe. t. Each crack configuration is analysed for two load cases; (1) Only primary (mechanical) loading is applied to the model, (2) Both secondary stresses and primary loading are applied to the model. Also presented in this report are some published experimental investigations conducted on cracked components of ductile materials subjected to both primary and secondary stresses. Based on the outcome of this study, an analysis strategy for fracture assessment of defects in ductile materials of nuclear components is proposed. A new

The application of leak before break concepts piping of KWU-plants

International Nuclear Information System (INIS)

Bartholome, G.; Bieselt, R.W.

1985-01-01

The fracture of pipes with longitudinal and circumferential cracks was investigated by experiments and theoretical approaches (flow stress criteria and limit load analyses). The experiments show that the critical crack dimensions can conservatively be determined by fracture mechanics. The tests and calculations are applied to KWU primary coolant piping with hypothetical longitudinal and circumferential defects. Reactor systems, design, fabrication, stress analysis, material, non-destructive testing, quality control and inservice inspection are considered referring to the leak-before-break behaviour. On the basis of the extreme toughness of the materials, the known loads, the high level of non-destructive examinations, the leakage monitoring system and the high quality of manufacture and processing it is shown that a spontaneous failure need not be postulated. (orig.)

Model experiments on imaging subsurface fracture permeability by pulsed Doppler borehole televiewer; Pulse doppler borehole televiewer ni yoru kiretsu tosuisei hyoka ni kansuru model jikken

Energy Technology Data Exchange (ETDEWEB)

Inagaki, Y; Niitsuma, H [Tohoku University, Sendai (Japan). Faculty of Engineering

1996-05-01

This paper reports model experiments to evaluate flow rates of fluids passing through a fracture by using a Doppler borehole televiewer (DBHTV). A supersonic transducer disposed on a well axis transmits transmission pulses, and a transducer receives scattered waves generated by particulates in water and waves reflected on a well wall. This signal is applied with time gating to extract only the scattered waves from particulates in the vicinity of the well wall. Deriving spectra in the recorded Doppler signal obtains flow velocity components in the direction of the well radius. A model was made with a polyvinylchloride pipe with a diameter of 14.6 cm to simulate a well, to which an aluminum pipe with an inner diameter of 2 mm is connected to be used as a simulated fracture, and mud water is circulated in the pipe. The result of deriving a passed flow volume in this model by integrating flow rate distribution derived by using the above method to a predetermined range in the vicinity of the fracture showed a good proportional relationship with actual flow rate in the simulated fracture. 1 ref., 7 figs.

Evaluation of effectiveness of inservice inspection in aging PWR plants based on PSA

Energy Technology Data Exchange (ETDEWEB)

Hanafusa, Hidemitsu; Irie, Takashi; Suyama, Takeshi [Institute of Nuclear Safety Systems Inc., Mihama, Fukui (Japan); Morota, Hidetsugu [Computer Software Development Co., Ltd., Tokyo (Japan)

2001-09-01

By making use of a probabilistic fracture mechanics (PFM) code which had been improved to be able to evaluate the influence of thermal aging embrittlement of cast stainless steel, we evaluated the fracture probability of welding parts of the piping of the pressurized water reactor plants in Japan and the core damage frequency by the fracture. In addition, we ranked the welding parts by the safety significance and evaluated the effectiveness of inservice inspection (ISI). As a result, it is seen that the risk increases about 25% for extending the plant life from 40 to 60 years without ISI. However, the influence of the risk on the plant life becomes negligible, when the ISI is performed adequately. This assessment is useful to plan efficient inspection and test programs of a plant under limited resources, because it is available to rank piping segments by the safety significance and to evaluate the effectiveness of inspection and test quantitatively. (author)

Study on criterion for leak before break assessment of pressure pipes

International Nuclear Information System (INIS)

Yang Linjuan

2009-01-01

Based on the elastoplastic fracture mechanics, this paper established the expression formulas of limit buckling pressure P u on the ligament of axial and circumferential surface cracks and the initial pressure for the through cracks P c . A new Leak Before Break (LBB) assessment criterion was put forward to predict the failure mode of pressure pipes, i.e., when P u is less than P c , the pipe will leak; when P u is equal to or larger than P c , the pipe will break, which is verified by the test data reported in literatures. (authors)

Impact of inservice inspection on the reliability of nuclear piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-12-01

The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

Evaluations of the piping system inelastic analysis computer program PIRAX2

International Nuclear Information System (INIS)

Rodabaugh, E.C.

1977-01-01

The report contains two sets of comparisons of inelastic test data with PIRAX2-Theory; i.e., ORNL beam tests and ORNL elbow tests. The purpose of these comparisons is to evaluate the accuracy of the simplified analytical techniques used in PIRAX2. The test data are on structures that are much simpler than piping systems but provide a fundamental basis for comparison. The report includes an analysis of a 3-anchor piping system to illustrate the relative simplicity of PIRAX2 input/output data and relatively small computer running time. Some areas of needed improvements in PIRAX2 are discussed

Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

International Nuclear Information System (INIS)

Datta, Debashis

2010-02-01

A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

Development of stress-modified fracture strain criterion for ductile fracture of API X65 steel

International Nuclear Information System (INIS)

Oh, Chang Kyun; Kim, Yun Jae; Park, Jin Moo; Kim, Woo Sik; Baek, Jong Hyun

2005-01-01

This paper presents a stress-modified fracture strain for API X65 steel used for gas pipeline, as a function of stress triaxiality. To determine the stress-modified fracture strain, tension test of bars with four different notch radii, made of API X65 steel, is firstly performed, from which true fracture strains are determined as a function of notch radius. Then detailed elastic-plastic, large strain Finite Element (FE) analyses are performed to estimate variations of stress triaxiality in the notched bars with load. Combining experimental with FE results provides the true fracture strain as a function of stress triaxiality, which is regarded as a criterion of ductile fracture. Application of the developed stress-modified fracture strain to failure prediction of gas pipes made of API X65 steel with various types of defects is discussed

Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

Probabilistic fracture mechanics in the integrity evaluation of components; Mecanica de fractura probabilista en la evaluacion de la integridad de componentes

Energy Technology Data Exchange (ETDEWEB)

Franco Nava, Jose Manuel; Torres Toledano, Jose Gerardo; Sanchez Sanchez, Ramon [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1996-12-31

Because of the importance of fracture mechanics in applications to pressure vessels, piping and other components used in thermal power plants, the Mechanical Engineering Unit of the Instituto de Investigaciones Electricas (IIE) has been developing activities in this field since some years ago. In this paper, the process for the analysis of probabilistic fracture mechanics is described. As an example, an application to the calculation of fault probability in piping systems, is presented. [Espanol] Dada la importancia de la mecanica de fractura en aplicaciones de recipientes a presion, tuberias y otros componentes utilizados en centrales termicas, la Unidad de Ingenieria Mecanica del Instituto de Investigaciones Electricas (IIE) ha venido desarrollando actividades en este campo desde hace algunos anos. En el presente trabajo, se describe el proceso para el analisis de mecanica de fractura probabilista. Como ejemplo se presenta una aplicacion al calculo de probabilidad de falla en tuberias.

Probabilistic fracture mechanics in the integrity evaluation of components; Mecanica de fractura probabilista en la evaluacion de la integridad de componentes

Energy Technology Data Exchange (ETDEWEB)

Franco Nava, Jose Manuel; Torres Toledano, Jose Gerardo; Sanchez Sanchez, Ramon [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1997-12-31

Because of the importance of fracture mechanics in applications to pressure vessels, piping and other components used in thermal power plants, the Mechanical Engineering Unit of the Instituto de Investigaciones Electricas (IIE) has been developing activities in this field since some years ago. In this paper, the process for the analysis of probabilistic fracture mechanics is described. As an example, an application to the calculation of fault probability in piping systems, is presented. [Espanol] Dada la importancia de la mecanica de fractura en aplicaciones de recipientes a presion, tuberias y otros componentes utilizados en centrales termicas, la Unidad de Ingenieria Mecanica del Instituto de Investigaciones Electricas (IIE) ha venido desarrollando actividades en este campo desde hace algunos anos. En el presente trabajo, se describe el proceso para el analisis de mecanica de fractura probabilista. Como ejemplo se presenta una aplicacion al calculo de probabilidad de falla en tuberias.

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

2001-03-15

This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

International Nuclear Information System (INIS)

Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

1982-01-01

One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

Fatigue crack growth rate studies on pipes and pipe welds made of austenitic stainless steel and carbon steel

International Nuclear Information System (INIS)

Arora, Punit; Singh, P.K.; Bhasin, Vivek; Vaze, K.K.; Pukazhendhi, D.M.; Gandhi, P.; Raghava, G.

2011-01-01

The objective of the present study is to understand the fatigue crack growth behavior in austenitic stainless steel and carbon steel pipes and pipe welds by carrying out analysis/predictions and experiments. The Paris law has been used for the prediction of fatigue crack growth life. To carry out the analysis, Paris constants have been determined for pipe (base) and pipe weld materials by using Compact Tension (CT)/Three Point Bend (TPB) specimens machined from the actual pipe/pipe weld. Analyses have been carried out to predict the fatigue crack growth life of pipes/pipe welds having part through cracks on the outer surface. In the analyses, Stress Intensity Factors (K) have been evaluated through two different schemes. The first scheme considers the 'K' evaluations at two points of the crack front i.e. maximum crack depth and crack tip at the outer surface. The second scheme accounts for the area averaged root mean square stress intensity factor (K RMS ) at deepest and surface points. In order to validate the analytical procedure/results, experiments have been carried out on full scale pipe and pipe welds with part through circumferential crack. Fatigue crack growth life evaluated using both schemes have been compared with experimental results. Use of stress intensity factor (K RMS ) evaluated using second scheme gives better fatigue crack growth life prediction compared to that of first scheme. (author)

Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

International Nuclear Information System (INIS)

Sudardjo; Histori; Triyadi, Ari

1998-01-01

The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

Safety evaluation of socket weld integrity in nuclear piping

International Nuclear Information System (INIS)

Choi, Y.H.; Kim, H.J.; Choi, S.Y.; Kim, Y.J.; Kim, Y.J.

2004-01-01

The purposes of this paper are to evaluate the integrity of socket weld in nuclear piping and prepare the technical basis for a new guideline on radiographic testing (RT) for the socket weld. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because lots of failures and leaks have been reported in the socket weld. The root causes of the socket weld failure are known as unanticipated loadings such as vibration or thermal fatigue and improper weld joint during construction. The ASME Code sec. III requires 1/16 inch gap between the pipe and fitting in the socket weld. Many failure cases, however, showed that the gap requirement was not satisfied. The Code also requires magnetic particle examination (MT) or liquid penetration examination (PT) on the socket weld, but not radiographic examination (RT). It means that it is not easy to examine the 1/16 inch gap in the socket weld by using the NDE methods currently required in the Code. In this paper, the effects of the requirements in the ASME Code sec. III on the socket weld integrity were evaluated by using finite element method. The crack behavior in the socket weld was also investigated under vibration event in nuclear power plants. The results showed that the socket weld was very susceptible to the vibration if the requirements in ASME Code were not satisfied. The constraint between the pipe and fitting due to the contact significantly affects the integrity of the socket weld. This paper also suggests a new guideline on the RT for the socket weld during construction stage in nuclear power plants. (orig.)

An investigation of wall temperature characteristics to evaluate thermal fatigue at a T-junction pipe

International Nuclear Information System (INIS)

Miyoshi, Koji; Nakamura, Akira; Utanohara, Yoichi; Takenaka, Nobuyuki

2014-01-01

Thermal fatigue cracking may initiate at a T-junction pipe where high and low temperature fluids mix. In this study, wall temperature characteristics at a T-junction pipe were investigated to improve the evaluation method for thermal fatigue. The stainless steel test section consisted of a horizontal main pipe (diameter, 150 mm) and a T-junction connected to a vertical branch pipe (diameter, 50 mm). The inlet flow velocities in the main and branch pipes were set to 0.99 m/s and 0.66 m/s respectively to produce a wall jet pattern in which the jet from the branch pipe was bent by the main pipe flow and made to flow along the pipe wall. The temperature difference was 34.1 K. A total of 148 thermocouples were installed to measure the wall temperature on the pipe inner surface in the downstream region. The maximum of temperature fluctuation intensity on the pipe inner surface was measured as 5% of the fluid temperature difference at the inlets. The dominant frequency of the large temperature fluctuations in the region downstream from z = 0.5D m was equal to 0.2 of the Strouhal number, which was equal to the frequency caused by the vortex streets generated around the jet flow. The large temperature fluctuation was also observed with the period of about 10 s. The fluctuation was caused by spreading of the heated region in the circumferential direction. (author)

Seismic ratchet-fatigue failure of piping systems

International Nuclear Information System (INIS)

Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

1987-01-01

Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

Evaluation on the thermal-hydraulic behavior of condensation pool and piping system

Energy Technology Data Exchange (ETDEWEB)

Kwon, Soon Bum; Lee, B. E.; Baek, S. C.; Joo, S. Y.; Lee, D. E.; Woo, S. W. [Kyungpook National Univ., Daegu (Korea, Republic of)

2002-03-15

The In-containment Refueling Water Storage Tank (IRWST) has the function of heat sink, when steam is released from the pressurizer. The hydrodynamic behaviors occurring at the piping system and sparger are very complex because of the wide variety of operating conditions and the complex geometry. Hydrodynamic behavior when air is discharged through a sparger in a condensation pool is investigated using CFD techniques in the present study. The effect of pressure acting on the sparger header during both water and air discharge through the sparger is studied. In addition, pressure oscillation occurring in the IRWST during air discharge through the sparger is studied for a better understanding of mechanisms of air discharge md an advanced evaluation technology of reactor safety. Understanding of flow behaviors occurring m the various types of pipes when POSRV is opened are also very important because those are very complex and may damage the structures of reactor coolant system. The principle of shock tube has been applied to analyze flow behaviors occurring in the piping system and several important phenomena which can be used for the evaluation of nuclear reactor safety has been obtained.

An evaluation of fracture toughness of bituminous coal

International Nuclear Information System (INIS)

Pathan, A.G.

2005-01-01

The role of fracture mechanics in the design of rock structures is vitally important. However, because of the complexities of rock structures and lack of understanding of the fundamentals of the failure mechanism, it has become customary to use the engineering properties approach in the design of stable rock structures. Recently considerable attention has been given and attempts are being made to apply the fracture mechanics approach to the design of safe mining structures. In mining engineering the fracture mechanics may be applied to calculate the formation of fracture zones around mine opening, thus estimating support requirements and formulating guide lines for the selection of mine roadway support system. The research work presented here is concerned with the evaluation of fracture toughness of coal under laboratory conditions. Diametral compression test method is used to determine the fracture toughness parameter of coal in the opening model failure. The effect of crack length and dimensionless crack length on the fracture toughness was studied also. A laboratory investigation of fracture toughness of coal in tensile mode failure has led to the conclusion that fracture toughness could be treated as a material property. (author)

Battelle integrity of nuclear piping program. Summary of results and implications for codes/standards

International Nuclear Information System (INIS)

Miura, Naoki

2005-01-01

The BINP(Battelle Integrity of Nuclear Piping) program was proposed by Battelle to elaborate pipe fracture evaluation methods and to improve LBB and in-service flaw evaluation criteria. The program has been conducted from October 1998 to September 2003. In Japan, CRIEPI participated in the program on behalf of electric utilities and fabricators to catch up the technical backgrounds for possible future revision of LBB and in-service flaw evaluation standards and to investigate the issues needed to be reflected to current domestic standards. A series of the results obtained from the program has been well utilized for the new LBB Regulatory Guide Program by USNRC and for proposal of revised in-service flaw evaluation criteria to the ASME Code Committee. The results were assessed whether they had implications for the existing or future domestic standards. As a result, the impact of many of these issues, which were concerned to be adversely affected to LBB approval or allowable flaw sizes in flaw evaluation criteria, was found to be relatively minor under actual plant conditions. At the same time, some issues that needed to be resolved to address advanced and rational standards in the future were specified. (author)

Dynamic Stability of Pipe Conveying Fluid with Crack and Attached Masses

International Nuclear Information System (INIS)

Ahn, Tae Soo; Yoon, Han Ik; Son, In Soo; Ahn, Sung Jin

2007-01-01

In this paper, the dynamic stability of a cracked simply supported pipe conveying fluid with an attached mass is investigated. Also, the effect of attached masses on the dynamic stability of a simply supported pipe conveying fluid is presented for the different positions and depth of the crack. Based on the Euler-Bernoulli beam theory, the equation of motion can be constructed by the energy expressions using extended Hamilton's principle. The crack section is represented by a local flexibility matrix connecting two undamaged pipe segments. The crack is assumed to be in the first mode of a fracture and to be always opened during the vibrations. Finally, the critical flow velocities and stability maps of the pipe conveying fluid are obtained by changing the attached masses and crack severity

X-ray diffraction study of microstructural changes during fatigue damage initiation in steel pipes

Energy Technology Data Exchange (ETDEWEB)

Pinheiro, B., E-mail: bianca@lts.coppe.ufrj.br [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Lesage, J. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Pasqualino, I. [Subsea Technology Laboratory (LTS), Ocean Engineering Department, COPPE/Federal University of Rio de Janeiro, PO Box 68508, Cidade Universitaria, CEP 21945-970, Rio de Janeiro/RJ (Brazil); Benseddiq, N. [Laboratory of Mechanics of Lille (LML), UMR CNRS 8107, University of Lille 1, Boulevard Paul Langevin, Cite Scientifique, 59655 Villeneuve d' Ascq (France); Bemporad, E. [Interdepartmental Laboratory of Electron Microscopy (LIME), University of Rome TRE, Via Della Vasca Navale 79, 00146 Rome (Italy)

2012-01-15

Highlights: Black-Right-Pointing-Pointer In this work we study the fatigue damage evolution in an API 5L X60 steel. Black-Right-Pointing-Pointer Microstructural changes and residual stresses are evaluated during fatigue tests. Black-Right-Pointing-Pointer Microdeformations and macro residual stresses are estimated by X-ray diffraction. Black-Right-Pointing-Pointer Results are discussed in view of an indicator of fatigue damage initiation. Black-Right-Pointing-Pointer This indicator could allow the prediction of residual life before macrocracking. - Abstract: Steel pipes used in the oil and gas industry undergo the action of cyclic loads that can cause their failure by fatigue. A consistent evaluation of the fatigue damage during the initiation phase should fundamentally be based on a nanoscale approach, i.e., at the scale of the dislocation network, in order to take into account the micromechanisms of fatigue damage that precede macrocrack initiation and propagation until the final fracture. In this work, microstructural changes related to fatigue damage initiation are investigated in the API 5L X60 grade steel, used in pipe manufacturing. Microdeformations and macro residual stress are evaluated using X-ray diffraction in real time during alternating bending fatigue tests performed on samples cut off from an X60 steel pipe. The aim of this ongoing work is to provide ground for further development of an indicator of fatigue damage initiation in X60 steel. This damage indicator could allow a good residual life prediction of steel pipes previously submitted to fatigue loading, before macroscopic cracking, and help to increase the reliability of these structures.

Material property evaluations of bimetallic welds, stainless steel saw fusion lines, and materials affected by dynamic strain aging

Energy Technology Data Exchange (ETDEWEB)

Rudland, D.; Scott, P.; Marschall, C.; Wilkowski, G. [Battelle Memorial Institute, Columbus, OH (United States)

1997-04-01

Pipe fracture analyses can often reasonably predict the behavior of flawed piping. However, there are material applications with uncertainties in fracture behavior. This paper summarizes work on three such cases. First, the fracture behavior of bimetallic welds are discussed. The purpose of the study was to determine if current fracture analyses can predict the response of pipe with flaws in bimetallic welds. The weld joined sections of A516 Grade 70 carbon steel to F316 stainless steel. The crack was along the carbon steel base metal to Inconel 182 weld metal fusion line. Material properties from tensile and C(T) specimens were used to predict large pipe response. The major conclusion from the work is that fracture behavior of the weld could be evaluated with reasonable accuracy using properties of the carbon steel pipe and conventional J-estimation analyses. However, results may not be generally true for all bimetallic welds. Second, the toughness of austenitic steel submerged-arc weld (SAW) fusion lines is discussed. During large-scale pipe tests with flaws in the center of the SAW, the crack tended to grow into the fusion line. The fracture toughness of the base metal, the SAW, and the fusion line were determined and compared. The major conclusion reached is that although the fusion line had a higher initiation toughness than the weld metal, the fusion-line J-R curve reached a steady-state value while the SAW J-R curve increased. Last, carbon steel fracture experiments containing circumferential flaws with periods of unstable crack jumps during steady ductile tearing are discussed. These instabilities are believed to be due to dynamic strain aging (DSA). The paper discusses DSA, a screening criteria developed to predict DSA, and the ability of the current J-based methodologies to assess the effect of these crack instabilities. The effect of loading rate on the strength and toughness of several different carbon steel pipes at LWR temperatures is also discussed.

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

Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report. [BWR

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

Acoustic emission reviling and danger level evaluation of stress corrosion cracking in stainless steel pipes

International Nuclear Information System (INIS)

Muravin, Gregory; Muravin, Boris; Lezvinsky, Luidmila

2000-01-01

Breakdowns and catastrophic damage occurring during the operation of nuclear power stations pipelines cause substantial economic and social loss annually throughout the world. Stress corrosion, vibration, fatigue, erosion, water shock, dynamic load, construction defects/errors are the main causes of pipes failures. For these reasons and in view of the age of nuclear power station pipes, there is an increased interest in finding means to prevent potential pipe failures. Nevertheless, statistical data of pipe failures continues to show significant numbers of accidents mainly due to stress corrosion cracking (about 65-80% of total number). To this end, a complex of investigations was carried out for the reliable AE diagnosis of pipes undergone stress corrosion cracking. These include: finding AE indications (fingerprints) of flaws developing in the metal in original condition as well as in metal subjected to stress corrosion; preparing AE criteria for evaluating the danger level of defects. (author)

Structural evaluation report of piping and support structure for design-changed hot-water layer system

International Nuclear Information System (INIS)

Ryu, Jeong Soo

1998-05-01

After hot-water layer system had been installed, the verification tests to reduce the radiation level at the top of reactor pool were performed many times. The major goal of this report is to assess the structural integrity on the piping and the support structures of design-changed hot-water layer system. The piping stress analysis was performed by using ADLPIPE program for the pump suction line and the pump discharge line subjected to dead weight, pressure, thermal expansion and seismic loadings. The stress analysis of the support structure was carried out using the reaction forces obtained from the piping stress analysis. The results of structural evaluation for the pipings and the support structures showed that the structural acceptance criteria were satisfied, in compliance with ASME, subsection ND for the piping and subsection NF for the support structures. Therefore based on the results of the analysis and the design, the structural integrity on the piping and the support structures of design-changed hot-water system was proved. (author). 9 refs., 9 tabs., 14 figs

The effects of cyclic and dynamic loading on the fracture resistance of nuclear piping steels. Technical report, October 1992--April 1996

Energy Technology Data Exchange (ETDEWEB)

Rudland, D.L.; Brust, F.; Wilkowski, G.M.

1996-12-01

This report presents the results of the material property evaluation efforts performed within Task 3 of the IPIRG-2 Program. Several related investigations were conducted. (1) Quasi-static, cyclic-load compact tension specimen experiments were conducted using parameters similar to those used in IPIRG-1 experiments on 6-inch nominal diameter through-wall-cracked pipes. These experiments were conducted on a TP304 base metal, an A106 Grade B base metal, and their respective submerged-arc welds. The results showed that when using a constant cyclic displacement increment, the compact tension experiments could predict the through-wall-cracked pipe crack initiation toughness, but a different control procedure is needed to reproduce the pipe cyclic crack growth in the compact tension tests. (2) Analyses conducted showed that for 6-inch diameter pipe, the quasi-static, monotonic J-R curve can be used in making cyclic pipe moment predictions; however, sensitivity analyses suggest that the maximum moments decrease slightly from cyclic toughness degradation as the pipe diameter increases. (3) Dynamic stress-strain and compact tension tests were conducted to expand on the existing dynamic database. Results from dynamic moment predictions suggest that the dynamic compact tension J-R and the quasi-static stress-strain curves are the appropriate material properties to use in making dynamic pipe moment predictions.

The effects of cyclic and dynamic loading on the fracture resistance of nuclear piping steels. Technical report, October 1992--April 1996

International Nuclear Information System (INIS)

Rudland, D.L.; Brust, F.; Wilkowski, G.M.

1996-12-01

This report presents the results of the material property evaluation efforts performed within Task 3 of the IPIRG-2 Program. Several related investigations were conducted. (1) Quasi-static, cyclic-load compact tension specimen experiments were conducted using parameters similar to those used in IPIRG-1 experiments on 6-inch nominal diameter through-wall-cracked pipes. These experiments were conducted on a TP304 base metal, an A106 Grade B base metal, and their respective submerged-arc welds. The results showed that when using a constant cyclic displacement increment, the compact tension experiments could predict the through-wall-cracked pipe crack initiation toughness, but a different control procedure is needed to reproduce the pipe cyclic crack growth in the compact tension tests. (2) Analyses conducted showed that for 6-inch diameter pipe, the quasi-static, monotonic J-R curve can be used in making cyclic pipe moment predictions; however, sensitivity analyses suggest that the maximum moments decrease slightly from cyclic toughness degradation as the pipe diameter increases. (3) Dynamic stress-strain and compact tension tests were conducted to expand on the existing dynamic database. Results from dynamic moment predictions suggest that the dynamic compact tension J-R and the quasi-static stress-strain curves are the appropriate material properties to use in making dynamic pipe moment predictions

Effect of vibration loading on the fatigue life of part-through notched pipe

International Nuclear Information System (INIS)

Mittal, Rahul; Singh, P.K.; Pukazhendi, D.M.; Bhasin, V.; Vaze, K.K.; Ghosh, A.K.

2011-01-01

A systematic experimental and analytical study has been carried out to investigate the effect of vibration loading on the fatigue life of the piping components. Three Point bend (TPB) specimens machined from the actual pipe have been used for the evaluation of Paris constants by carrying out the experiments under vibration + cyclic and cyclic loading as per the ASTM Standard E647. These constants have been used for the prediction of the fatigue life of the pipe having part-through notch of a/t = 0.25 and aspect ratio (2c/a) of 10. Predicted results have shown the reduction in fatigue life of the notched pipe subjected to vibration + cyclic loading by 50% compared to that of cyclic loading. Predicted results have been validated by carrying out the full-scale pipe (with part-through notch) tests. Notched pipes were subjected to loading conditions such that the initial stress-intensity factor remains same as that of TPB specimen. Experimental results of the full-scale pipe tests under vibration + cyclic loading has shown the reduction in fatigue life by 70% compared to that of cyclic loading. Fractographic examination of the fracture surface of the tested specimens subjected to vibration + cyclic loading have shown higher presence of brittle phases such as martensite (in the form of isolated planar facets) and secondary micro cracks. This could be the reason for the reduction of fatigue life in pipe subjected to vibration + cyclic loading. - Highlights: → Vibration loading affects fatigue crack growth rate. → Crack initiation life depends on crack tip radius. → Crack initiation life depends on the characteristic distance. → Characteristic distance depends on the loading conditions. → Vibration + cyclic load gives lower fatigue life.

Clinical evaluation of stress fractures using bone scintigraphy

International Nuclear Information System (INIS)

Furuta, Atsuhiko; Tanohata, Kazunori; Otake, Toru; Hashizume, Toshiyuki; Kobayashi, Yozi; Nakazima, Hiroyuki.

1984-01-01

Clinical evaluation of stress fractures were performed in 58 athletes using bone scintigraphy with sup(99m)Tc-MDP. Stress fractures of the tibia were most often seen in the males with running type sports. They occurred more often in the proximal tibia and on the right side. Stress fractures of the fibula were most often seen in females with jumping type sports, such as volley ball. They occurred more often in the distal fibula and on the right side. Tarsal bone fractures were seen most often rugby players. Metatarsal fractures occurred in the third fourth and fifth metatarsals. No lesion was seen in the first and second metatarsals. We feel that stress fractures of the femur can be differentiated from osteosarcoma by small loculated radionuclide accumulation as well as symptome, course and tomographic and CT finding. Bilateral involvement was seen in two cases in patellae and calcanei. Most of the other fractures were seen on the right side. Negative radiographs were seen in 36% of the patients and occurred most commonly in the tarsal bones excluding calcaneus. Bone scintigrams were positive in all cases and were most useful in fractures of the tarsal bones excluding calcaneus. (author)

Clinical evaluation of stress fractures using bone scintigraphy

Energy Technology Data Exchange (ETDEWEB)

Furuta, Atsuhiko; Tanohata, Kazunori; Otake, Toru; Hashizume, Toshiyuki (Kanto Rosai Hospital, Kawasaki, Kanagawa (Japan)); Kobayashi, Yozi; Nakazima, Hiroyuki

1984-05-01

Clinical evaluation of stress fractures were performed in 58 athletes using bone scintigraphy with sup(99m)Tc-MDP. Stress fractures of the tibia were most often seen in the males with running type sports. They occurred more often in the proximal tibia and on the right side. Stress fractures of the fibula were most often seen in females with jumping type sports, such as volley ball. They occurred more often in the distal fibula and on the right side. Tarsal bone fractures were seen most often rugby players. Metatarsal fractures occurred in the third fourth and fifth metatarsals. No lesion was seen in the first and second metatarsals. We feel that stress fractures of the femur can be differentiated from osteosarcoma by small loculated radionuclide accumulation as well as symptoms, course and tomographic and CT findings. Bilateral involvement was seen in two cases in patellae and calcanei. Most of the other fractures were seen on the right side. Negative radiographs were seen in 36% of the patients and occurred most commonly in the tarsal bones excluding calcaneus. Bone scintigrams were positive in all cases and were most useful in fractures of the tarsal bones excluding calcaneus.

Numerical modelling of coupled fluid, heat, and solute transport in deformable fractured rock

International Nuclear Information System (INIS)

Chan, T.; Reid, J.A.K.

1987-01-01

This paper reports on a three-dimensional (3D) finite-element code, MOTIF (model of transport in fractured/porous media), developed to model the coupled processes of groundwater flow, heat transport, brine transport, and one-species radionuclide transport in geological media. Three types of elements are available: a 3D continuum element, a planar fracture element that can be oriented in any arbitrary direction in 3D space or pipe flow in 3D space, and a line element for simulating fracture flow in 2D space or pipe flow in 3D space. As a quality-assurance measure, the MOTIF code was verified by comparison of its results with analytical solutions and other published numerical solutions

Rupture hardware minimization in pressurized water reactor piping

International Nuclear Information System (INIS)

Mukherjee, S.K.; Ski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.F.; Quinones, D.F.; Server, W.L.

1989-01-01

For much of the high-energy piping in light reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied a Beaver Valley Power Station- Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferrutic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elemination

Fracture network evaluation program (FraNEP): A software for analyzing 2D fracture trace-line maps

Science.gov (United States)

Zeeb, Conny; Gomez-Rivas, Enrique; Bons, Paul D.; Virgo, Simon; Blum, Philipp

2013-10-01

Fractures, such as joints, faults and veins, strongly influence the transport of fluids through rocks by either enhancing or inhibiting flow. Techniques used for the automatic detection of lineaments from satellite images and aerial photographs, LIDAR technologies and borehole televiewers significantly enhanced data acquisition. The analysis of such data is often performed manually or with different analysis software. Here we present a novel program for the analysis of 2D fracture networks called FraNEP (Fracture Network Evaluation Program). The program was developed using Visual Basic for Applications in Microsoft Excel™ and combines features from different existing software and characterization techniques. The main novelty of FraNEP is the possibility to analyse trace-line maps of fracture networks applying the (1) scanline sampling, (2) window sampling or (3) circular scanline and window method, without the need of switching programs. Additionally, binning problems are avoided by using cumulative distributions, rather than probability density functions. FraNEP is a time-efficient tool for the characterisation of fracture network parameters, such as density, intensity and mean length. Furthermore, fracture strikes can be visualized using rose diagrams and a fitting routine evaluates the distribution of fracture lengths. As an example of its application, we use FraNEP to analyse a case study of lineament data from a satellite image of the Oman Mountains.

Compilation of references, data sources and analysis methods for LMFBR primary piping system components

International Nuclear Information System (INIS)

Reich, M.; Esztergar, E.P.; Ellison, E.G.; Erdogan, F.; Gray, T.G.F.; Wells, C.W.

1977-03-01

A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions

Pipe-to-pipe impact tests

Energy Technology Data Exchange (ETDEWEB)

Bampton, M C.C.; Alzheimer, J M; Friley, J R; Simonen, F A

1985-11-01

Existing licensing criteria express what damage shall be assumed for various pipe sizes as a consequence of a postulated break in a high energy system. The criteria are contained in Section 3.6.2 of the Standard Review Plan, and the purpose of the program described with this paper is to evaluate the impact criteria by means of a combined experimental and analytical approach. A series of tests has been completed. Evaluation of the test showed a deficiency in the range of test parameters. These deficiencies are being remedied by a second series of tests and a more powerful impact machine. A parallel analysis capability has been developed. This capability has been used to predict the damage for the first test series. The quality of predictions has been improved by tests that establish post-crush and bending relationships. Two outputs are expected from this project: data that may, or may not, necessitate changes to the criteria after appropriate value impact evaluations and an analytic capability for rapidly evaluating the potential for pipe whip damage after a postulated break. These outputs are to be contained in a value-impact document and a program final report. (orig.).

Development of piping strain sensor for stress evaluation

International Nuclear Information System (INIS)

Takahama, Tsunemichi; Nishimura, Kazuma; Ninomiya, Seiichiro; Matsumoto, Yoshihiro; Harada, Yutaka

2014-01-01

In a small diameter piping, stresses are generated due to internal fluid or pump vibrations especially around the welding parts. Authors have successfully developed a pipe strain sensor which is able to measure such stresses. Unlike conventional methods using strain gages and adhesive bond, the sensor can measure the strain without putting adhesive bond on the piping surface. However, the strain sensor can provide measurements with a level of accuracy equivalent to that of conventional method using strain gages and adhesive bond. Accordingly, the strain sensor can significantly reduce the working time without any loss of the measurement accuracy. (author)

Evaluation of Trenchless Installation Technology for Radioactive Wastewater Piping Applications

International Nuclear Information System (INIS)

Robinson, Sharon M.; Jubin, Robert Thomas; Patton, Bradley D.; Sullivan, Nicholas M.; Bugbee, Kathy P.

2009-01-01

The U.S. Department of Energy (DOE) Office of Environmental Management (EM) cleanup mission at Oak Ridge National Laboratory (ORNL) includes dispositioning facilities, contaminated legacy materials/waste, and contamination sources and remediation of soil under facilities, groundwater, and surface water to support final Records of Decision (RODs). The Integrated Facilities Disposition Project (IFDP) is a roughly $15B project for completion of the EM mission at Oak Ridge, with a project duration of up to 35 years. The IFDP Mission Need Statement - Critical Decision-0 (CD-0) - was approved by DOE in July 2007, and the IFDP Alternative Selection and Cost Range - Critical Decision-1 (CD-1) - was approved in November 2008. The IFDP scope includes reconfiguration of waste collection and treatment systems as needed to complete the IFDP remediation and decontamination and decommissioning (D and D) missions in a safe and cost-effective manner while maintaining compliance with all governing regulations and bodies and preserving the support of continuing operations at ORNL. A step in the CD-1 approval process included an external technical review (ETR) of technical approaches proposed in the CD-1 document related to the facility reconfiguration for the ORNL radioactive waste and liquid low-level waste management systems. The ETR team recommended that the IFDP team consider the use of trenchless technologies for installing pipelines underground in and around contaminated sites as part of the alternatives evaluations required in support of the CD-2 process. The team specifically recommended evaluating trenchless technologies for installing new pipes in existing underground pipelines as an alternative to conventional open trench installation methods. Potential benefits could include reduction in project costs, less costly underground piping, fewer disruptions of ongoing and surface activities, and lower risk for workers. While trenchless technologies have been used extensively in

Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

International Nuclear Information System (INIS)

Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

1989-01-01

A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

Anterior process calcaneal fractures: a systematic evaluation of associated conditions

Energy Technology Data Exchange (ETDEWEB)

Petrover, David [NYU Hospital for Joint Disease, Radiology Department, New York, NY (United States); Hopital Beaujon, Service de Radiologie, Paris (France); Schweitzer, Mark E. [NYU Hospital for Joint Disease, Radiology Department, New York, NY (United States); Laredo, J.D. [Hopital Lariboisiere, Service de Radiologie, Paris (France)

2007-07-15

The objective was to evaluate the association, by MRI, of anterior calcaneal process fractures with tarsal coalitions, ankle sprains, and bifurcate ligament abnormalities. A retrospective review of 1,479 foot and ankle MR images was performed, over a period of 5 years, for isolated anterior process fractures of the calcaneus. Fifteen 1.5-T MR examinations were systematically evaluated by two radiologists in consensus. Marrow edema patterns, presence of a calcaneonavicular coalition, as well as bifurcate and anterior talofibular ligaments, were evaluated. There were 15 fractures of the anterior calcaneal process with an incidence of 1%. The average patient age was 51 years (range 25-82). Twelve patients were women and 3 were men. The majority of the fractures (14 out of 15) presented as an edema pattern on T2-weighted images, either diffuse (9 out of 15), or vertical (5 out of 15). One case did not show marrow edema, but rather a hypointense line. Nine patients (60%) demonstrated calcaneonavicular coalition and anterior calcaneal process fracture. In 6 patients (50%) the anterior talofibular ligament (ATFL) was thickened. Three patients did not have axial images, and were classified as non-conclusive for the ATFL evaluation. The bifurcate ligament was thickened with hyperintense signal demonstrating a sprain in 9 out of 13 (69%). Only 2 patients (16.5%) had an anterior calcaneal process fracture without any associated abnormality. We believe that there is a probable association of anterior process fractures and calcaneonavicular coalitions. We also feel, based on our results and the prior literature that there is likely also an association with both ATFL injuries and bifurcate ligament injuries. (orig.)

Anterior process calcaneal fractures: a systematic evaluation of associated conditions

International Nuclear Information System (INIS)

Petrover, David; Schweitzer, Mark E.; Laredo, J.D.

2007-01-01

The objective was to evaluate the association, by MRI, of anterior calcaneal process fractures with tarsal coalitions, ankle sprains, and bifurcate ligament abnormalities. A retrospective review of 1,479 foot and ankle MR images was performed, over a period of 5 years, for isolated anterior process fractures of the calcaneus. Fifteen 1.5-T MR examinations were systematically evaluated by two radiologists in consensus. Marrow edema patterns, presence of a calcaneonavicular coalition, as well as bifurcate and anterior talofibular ligaments, were evaluated. There were 15 fractures of the anterior calcaneal process with an incidence of 1%. The average patient age was 51 years (range 25-82). Twelve patients were women and 3 were men. The majority of the fractures (14 out of 15) presented as an edema pattern on T2-weighted images, either diffuse (9 out of 15), or vertical (5 out of 15). One case did not show marrow edema, but rather a hypointense line. Nine patients (60%) demonstrated calcaneonavicular coalition and anterior calcaneal process fracture. In 6 patients (50%) the anterior talofibular ligament (ATFL) was thickened. Three patients did not have axial images, and were classified as non-conclusive for the ATFL evaluation. The bifurcate ligament was thickened with hyperintense signal demonstrating a sprain in 9 out of 13 (69%). Only 2 patients (16.5%) had an anterior calcaneal process fracture without any associated abnormality. We believe that there is a probable association of anterior process fractures and calcaneonavicular coalitions. We also feel, based on our results and the prior literature that there is likely also an association with both ATFL injuries and bifurcate ligament injuries. (orig.)

Promethus Hot Leg Piping Concept

International Nuclear Information System (INIS)

AM Girbik; PA Dilorenzo

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

Comparative study of approaches to estimate pipe break frequencies

Energy Technology Data Exchange (ETDEWEB)

Simola, K.; Pulkkinen, U.; Talja, H.; Saarenheimo, A.; Karjalainen-Roikonen, P. [VTT Industrial Systems (Finland)

2002-12-01

The report describes the comparative study of two approaches to estimate pipe leak and rupture frequencies for piping. One method is based on a probabilistic fracture mechanistic (PFM) model while the other one is based on statistical estimation of rupture frequencies from a large database. In order to be able to compare the approaches and their results, the rupture frequencies of some selected welds have been estimated using both of these methods. This paper highlights the differences both in methods, input data, need and use of plant specific information and need of expert judgement. The study focuses on one specific degradation mechanism, namely the intergranular stress corrosion cracking (IGSCC). This is the major degradation mechanism in old stainless steel piping in BWR environment, and its growth is influenced by material properties, stresses and water chemistry. (au)

Proximal Humerus Fractures: Evaluation and Management in the Elderly Patient

Science.gov (United States)

Grawe, Brian

2018-01-01

Introduction: Proximal humerus fractures are common in the elderly. The evaluation and management of these injuries is often controversial. The purpose of this study is to review recent evidence and provide updated recommendations for treating proximal humerus fractures in the elderly. Methods: A literature review of peer-reviewed publications related to the evaluation and management of proximal humerus fractures in the elderly was performed. There was a focus on randomized controlled trials and systematic reviews published within the last 5 years. Results: The incidence of proximal humerus fractures is increasing. It is a common osteoporotic fracture. Bone density is a predictor of reduction quality and can be readily assessed with anteroposterior views of the shoulder. Social independence is a predictor of outcome, whereas age is not. Many fractures are minimally displaced and respond acceptably to nonoperative management. Displaced and severe fractures are most frequently treated operatively with intramedullary nails, locking plates, percutaneous techniques, or arthroplasty. Discussion: Evidence from randomized controlled trials and systematic reviews is insufficient to recommend a treatment; however, most techniques have acceptable or good outcomes. Evaluation should include an assessment of the patient’s bone quality, social independence, and surgical risk factors. With internal fixation, special attention should be paid to medial comminution, varus angulation, and restoration of the calcar. With arthroplasty, attention should be paid to anatomic restoration of the tuberosities and proper placement of the prosthesis. Conclusion: A majority of minimally displaced fractures can be treated conservatively with early physical therapy. Treatment for displaced fractures should consider the patient’s level of independence, bone quality, and surgical risk factors. Fixation with percutaneous techniques, intramedullary nails, locking plates, and arthroplasty are all

Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

Method to evaluate stress and deformation of small-sized buried pipe induced by wheel loads. Shokokei maisetsukan no rinkaju ni okeru oryoku henkei hyokaho

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Y.; Kataoka, T.; Kokusho, T.; Yoshida, Y. (Central Research Institute of Electric Power Industry, Tokyo (Japan))

1994-03-21

In order to establish a practical method to evaluate safety of buried pipes, which were generally used as cable-protection pipes, at shallow depth and under vehicle loads, a series of experiments and analyses were performed. Based on the results, a simplified method to evaluate stress and deformation of the buried pipes in pavement were proposed. In the experiments, hard PVC pipes, light steel conduit pipes, and corrugated hard PE pipes in nominal diameters from 75 to 200 mm were applied as specimens for representative flexible pipes, and field load tests in actual size as well as laboratory load tests using soil vessels were carried out. The calculated results by the proposed method were compared to the experimental results, finding that the calculated values gave a satisfactory agreement with measured values which were obtained by the field tests. As a result, it was confirmed that the practical method to evaluate circumferential stresses at the top of pipes and flat-deflection ratios of small-sized buried Pipes induced by wheel loads under various conditions were presented. 20 refs., 17 figs., 6 tabs.

Flow induced vibrations of piping system (Vibration sources - Mechanical response of the pipes)

International Nuclear Information System (INIS)

Gibert, R.J.; Axisa, F.; Villard, B.

1978-01-01

In order to design the supports of piping system, an estimation of the vibration induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary. To evaluate the power spectra of all the main sources generated by the flow. These sources are located at the singular points of the circuit (enlargements, bends, valves, etc. ...). To calculate the modal parameters of fluid containing pipes. This paper presents: a methodical study of the most current singularities. Inter-correlation spectra of local pressure fluctuation downstream from the singularity and correlation spectra of associated acoustical sources have been measured. A theory of noise generation by unsteady flow in internal acoustics has been developed. All these results are very useful for evaluating the source characteristics in most practical pipes. A comparison between the calculation and the results of an experimental test has shown a good agreement

Evaluation of Bamboo Porous Pipe as Line Source Emitter in Trickle ...

African Journals Online (AJOL)

This paper attempts to evaluate the use of bamboo as porous pipe (line source) emitter in trickle irrigation at the Cross River University of Technology Teaching and Research Farm Obubra. Two sets of bamboo laterals: opened and plugged ends were used for the trial. The experiment was conducted using four different ...

evaluation of bamboo porous pipe as line source emitter in trickle ...

African Journals Online (AJOL)

CHRISTY

This paper attempts to evaluate the use of bamboo as porous pipe (line source) emitter in trickle irrigation at the Cross River University of Technology Teaching and Research Farm Obubra. Two sets of bamboo laterals: opened and plugged ends were used for the trial. The experiment was conducted using four different ...

Effect of pipe rupture loads inside containment in the break exclusionary piping outside containment

International Nuclear Information System (INIS)

Weiss, G.

1987-01-01

The plant design for protection against piping failures outside containment should make sure that fluid system piping in containment penetration areas are designed to meet the break exclusionary provisions contained in the BTP MEB 3-1. According to these provisions, following a piping failure (main steam line) inside containment, the part of the flued head connected to the piping outside containment, should not exceed the ASME Code stress limits for the appropriate load combinations. A finite element analysis has been performed to evaluate the stress level in this area. (orig./HP)

Performance evaluation of citric ion-stabilized magnetic fluid heat pipe

Energy Technology Data Exchange (ETDEWEB)

Jeyadevan, B. [Graduate School of Environmental Studies, Department of Geoscience and Technology, Tohoku University, Aramaki, Aoba 01, Aoba-ku, Sendai 980-7589 (Japan)]. E-mail: jeya@mail.kankyo.tohoku.ac.jp; Koganezawa, H. [Graduate School of Environmental Studies, Department of Geoscience and Technology, Tohoku University, Aramaki, Aoba 01, Aoba-ku, Sendai 980-7589 (Japan); Nakatsuka, K. [Graduate School of Environmental Studies, Department of Geoscience and Technology, Tohoku University, Aramaki, Aoba 01, Aoba-ku, Sendai 980-7589 (Japan)

2005-03-15

The performance of heat pipe (HP) using citric ion-stabilized magnetic fluid (CMF) as working fluid (WF) was evaluated. The heat transferred was influenced by the application of magnetic field and was enhanced by a maximum of 30% compared to the field-free case. Furthermore, under the optimum magnetic field configuration, the heat transferred by CMF HP was 10% higher than that with water as WF.

Stability of cracked pipe under seismic/dynamic displacement-controlled stresses. Subtask 1.2 final report

International Nuclear Information System (INIS)

Kramer, G.; Veith, P.; Marschall, C.

1997-06-01

Results of displacement-controlled pipe fracture experiments, analyses, and material characterization efforts performed within the International Piping Integrity Research Group, IPIRG, Program Subtask 1.2 are discussed. Effects of dynamic versus quasi-static and monotonic versus cyclic loading were evaluated for ductile tearing of two materials, A106 Grade B ferritic steel and TP304 austenitic steel. Twelve through-wall-cracked pipe experiments were conducted on 6-inch diameter Schedule 120 pipe at 288 C (550 F). The results indicated dynamic loading at seismic strain rates marginally increased the load-carrying capacity of austenitic steel. The ferritic steel tested was sensitive to dynamic strain-aging, and consequently, its load-carrying capacity decreased at dynamic strain rates. Two parameters were found to affect the apparent ductile crack growth resistance during cyclic loading, load ratio (R) and incremental plastic displacement that occurs in a cycle. Cyclic (R = 0) loading had minimal effect on ductile tearing for both materials. However, fully reversed loading decreased the load-carrying capacity and toughness for both materials. The incremental plastic displacement can be as important as the load ratio; however, it is harder to quantify from design stress reports. Large plastic displacements will minimize the effect of negative load ratios

Proposal of failure criterion applicable to finite element analysis results for wall-thinned pipes under bending load

Energy Technology Data Exchange (ETDEWEB)

Meshii, Toshiyuki, E-mail: meshii@u-fukui.ac.jp [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui (Japan); Ito, Yoshiaki [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui (Japan)

2012-01-15

Highlights: Black-Right-Pointing-Pointer Limit bending load (LBL) of wall-thinned pipe by large strain FEA was considered. Black-Right-Pointing-Pointer Net section yield load had sufficient margin to LBL. Black-Right-Pointing-Pointer LBL for collapse was the load when volume with nominal thickness yielded. Black-Right-Pointing-Pointer LBL for cracking was the load when flawed section stress exceeded tensile strength. Black-Right-Pointing-Pointer Failure criterion considering above was named Domain Collapse Criterion. - Abstract: In this work, a failure criterion applicable to large strain Finite Element Analysis (FEA) results was proposed in order to predict both the fracture mode (collapse or cracking) and the limit bending load of wall-thinned straight pipes. This work was motivated from the recent experimental results of ; that is, fracture mode is not always collapse, and the fracture mode affects the limit bending load. The key finding in comparing their test results and a detailed large strain FEA results was that the Mises stress distribution at the limit bending load of a flawed cylinder was similar to that of a flawless cylinder; specifically, in case of collapse, the Mises stress exceeded the true yield stress of a material for the whole 'volume' of a cylinder with a nominal wall thickness. Based on this finding, a failure criterion applicable to large strain FEA results of wall-thinned straight pipes under a bending load that can predict both fracture mode and limit bending load was proposed and was named the Domain Collapse Criterion (DCC). DCC predicts the limit bending load as the lower value of either the M{sub c}{sup FEA}, which is the load at which the Mises stress exceeds the true yield strength of a straight pipe for the whole 'volume' with a nominal wall thickness (fracture mode: collapse), or the M{sub c}{sup FEAb}, which is the load at which the Mises stress in a section of the flaw ligament exceeds the true tensile stress

Pipe Wall Thinning Evaluation through the Arrival Time Delay of A0 Lamb Wave Using Magnetostrictive Patch Transducers

Energy Technology Data Exchange (ETDEWEB)

Cho, Seung Hyun; Kwon, Hyu Sang; Ahn, Bong Young; Lee, Seung Seok [Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

2008-12-15

Guided wave technology is advantageous for fast inspection of pipe wall thinning since the guided wave propagates long distance. In this investigation, the method to evaluate gradual wall thinning in a pipe based on the arrival time delay with magnetostrictive patch transducers is presented. Low frequency A0 Lamb waves were generated and measured by the present transducer and it was applied to arrival time delay measurement experiments on a test pipe having gradual wall thinnings artificially manufactured. From experiments, consistent results that wall thinning increases the arrival time delay of A0 waves were obtained. Consequently, the feasibility of the magnetostrictive patch transducers to evaluate wall thinning was verified

Pipe Wall Thinning Evaluation through the Arrival Time Delay of A0 Lamb Wave Using Magnetostrictive Patch Transducers

International Nuclear Information System (INIS)

Cho, Seung Hyun; Kwon, Hyu Sang; Ahn, Bong Young; Lee, Seung Seok

2008-01-01

Guided wave technology is advantageous for fast inspection of pipe wall thinning since the guided wave propagates long distance. In this investigation, the method to evaluate gradual wall thinning in a pipe based on the arrival time delay with magnetostrictive patch transducers is presented. Low frequency A0 Lamb waves were generated and measured by the present transducer and it was applied to arrival time delay measurement experiments on a test pipe having gradual wall thinnings artificially manufactured. From experiments, consistent results that wall thinning increases the arrival time delay of A0 waves were obtained. Consequently, the feasibility of the magnetostrictive patch transducers to evaluate wall thinning was verified

Subprogram Calculating The Distance Between Pipe And Plane For Automatic Piping System Design

International Nuclear Information System (INIS)

Satmoko, Ari

2001-01-01

DISTLNPL subprogram was created using Auto LISP software. This subprogram is planned to complete CAPD (Computer Aided Piping Design) software being developed. The CAPD works under the following method: suggesting piping system line and evaluating whether any obstacle allows the proposed line to be constructed. DISTLNPL is able to compute the distance between pipe and any equipment having plane dimension such as wall, platform, floors, and so on. The pipe is modeled by using a line representing its axis, and the equipment is modeled using a plane limited by some lines. The obtained distance between line and plane gives information whether the pipe crosses the equipment. In the case of crashing, the subprogram will suggest an alternative point to be passed by piping system. So far, DISTLNPL has not been able to be accessed by CAPD yet. However, this subprogram promises good prospect in modeling wall, platform, and floors

View of industry on the impact of pipe break criteria

International Nuclear Information System (INIS)

Bernsen, S.A.

1983-01-01

Historically, large pipe breaks in the types of materials used and under operating conditions similar to those in light water reactor service have not occurred. Nevertheless, the non-mechanistic assumption of a double ended pipe break of the early sixties, selected for loss of coolant accident analysis purposes, has become a mechanistic criterion for the design and arrangement of high pressure piping systems and their associated supports and enclosures in today's nuclear plants. While it seems reasonable and appropriate to continue to design the Emergency Core Cooling Systems for a range of loss of coolant accidents up to and including those that approximate the area of the largest pipe connected to the reactor vessel and to use this break in determining the loading and temperature rise rate for containment structures and equipment qualification, it no longer seems reasonable to provide precisely engineered break protection for a limited number of potential pipe break locations. This observation is gaining increasing support, particularly as engineering judgment and historical perspectives are being supplemented by both deterministic and probabilistic studies that indicate the potential for large instantaneous breaks in nuclear grade piping systems is virtually incredible. Fracture mechanics analyses support leak-before-break assumptions with wide margins and probabilistic studies indicate potentials for double-ended pipe breaks in the range of less than one in a billion years

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

Energy Technology Data Exchange (ETDEWEB)

1985-04-01

This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.

Computed tomographic evaluation of comminuted middle phalangeal fractures in the horse

International Nuclear Information System (INIS)

Rose, P.L.; Seeherman, H.; O'Callaghan, M.

1997-01-01

Comminuted fractures of the middle phalanx have been well described in the horse. Choice of treatment, surgical planning and prognosis have traditionally been based upon evaluation of radiographs. However, the complex nature of comminuted fractures makes radiographic interpretation difficult. Computed tomography (CT) allows the production of cross-sectional images with spatial separation of structures which are superimposed on survey radiographs. This allows accurate assessment of the number and direction of fracture lines within the bone. In this paper we report the use of CT in the evaluation of 6 comminuted middle phalangeal fractures. Computed tomography is potentially useful in deciding the type of treatment, surgical planning and determining the prognosis

Degradation of impact fracture during accelerated aging of weld metal on microalloyed steel

International Nuclear Information System (INIS)

Vargas-Arista, B.; Hallen, J. M.; Albiter, A.; Angeles-Chavez, C.

2008-01-01

The effect of accelerated aging on the toughness and fracture of the longitudinal weld metal on an API5L-X52 line pipe steel was evaluated by Charpy V-notch impact test, fracture analysis and transmission electron microscopy. Aging was performed at 250 degree centigrade for 100 to 1000 h. The impact results indicated a significant reduction in the fracture energy and impact toughness as a function of aging time, which were achieved by the scanning electron microscope fractography that showed a decrease in the vol fraction of microvoids by Charpy ductile failure with the aging time, which favored the brittle fracture by transgranular cleavage. The minimum vol fraction of microvoids was reached at 500 h due to the peak aged. The microstructural analysis indicated the precipitation of transgranular iron nano carbides in the aged specimens, which was related to the deterioration of toughness and change in the ductile to brittle behavior. (Author) 15 refs

Pipe rupture hardware minimization in pressurized water reactor system

International Nuclear Information System (INIS)

Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.

1987-01-01

For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination

Integrity assessment of the cold leg piping system in a PWR

International Nuclear Information System (INIS)

Mayfield, M.E.; Leis, B.N.

1981-01-01

The purpose of this paper is to examine the integrity of a nuclear piping system, designed in accordance with Section III, in the context of a damage tolerance analysis procedure. Such a procedure directly addresses the defects and cyclic loadings that are responsible for the above noted exceptions. The analysis and results reported here are for a fatigue life analysis of the Cold Leg piping in a PWR. This piping system is particularly important from a safety standpoint since a large break is a possible initiator of a core meltdown accident. The analysis employs LEFM concepts to determine the time between the initial start-up and (1) formation of a leak, (2) detection of the leak, and (3) the final fracture of the piping. Both longitudinal and circumferential defects are considered. The defects are assumed to propagate from the pipe I.D. in a self-similar manner. Inputs to the analysis were derived from information supplied by plant operators and vendors, published data, and 'expert opinions'. The life was computed using a linear damage accumulation. (orig./GL)

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

Development of crack shape: LBB methodology for cracked pipes

Energy Technology Data Exchange (ETDEWEB)

Moulin, D.; Chapuliot, S.; Drubay, B. [Commissariat a l Energie Atomique, Gif sur Yvette (France)

1997-04-01

For structures like vessels or pipes containing a fluid, the Leak-Before-Break (LBB) assessment requires to demonstrate that it is possible, during the lifetime of the component, to detect a rate of leakage due to a possible defect, the growth of which would result in a leak before-break of the component. This LBB assessment could be an important contribution to the overall structural integrity argument for many components. The aim of this paper is to review some practices used for LBB assessment and to describe how some new R & D results have been used to provide a simplified approach of fracture mechanics analysis and especially the evaluation of crack shape and size during the lifetime of the component.

Soil-pipe interaction modeling for pipe behavior prediction with super learning based methods

Science.gov (United States)

Shi, Fang; Peng, Xiang; Liu, Huan; Hu, Yafei; Liu, Zheng; Li, Eric

2018-03-01

Underground pipelines are subject to severe distress from the surrounding expansive soil. To investigate the structural response of water mains to varying soil movements, field data, including pipe wall strains in situ soil water content, soil pressure and temperature, was collected. The research on monitoring data analysis has been reported, but the relationship between soil properties and pipe deformation has not been well-interpreted. To characterize the relationship between soil property and pipe deformation, this paper presents a super learning based approach combining feature selection algorithms to predict the water mains structural behavior in different soil environments. Furthermore, automatic variable selection method, e.i. recursive feature elimination algorithm, were used to identify the critical predictors contributing to the pipe deformations. To investigate the adaptability of super learning to different predictive models, this research employed super learning based methods to three different datasets. The predictive performance was evaluated by R-squared, root-mean-square error and mean absolute error. Based on the prediction performance evaluation, the superiority of super learning was validated and demonstrated by predicting three types of pipe deformations accurately. In addition, a comprehensive understand of the water mains working environments becomes possible.

Risk based service life prediction of underground cast iron pipes subjected to corrosion

International Nuclear Information System (INIS)

Li, C.Q.; Mahmoodian, M.

2013-01-01

Aging and deterioration of underground cast iron pipes is inevitable after their long time in service, with corrosion being the most predominant mechanism for pipe failures. Although considerable research has been undertaken in the past few decades, more is on the effects of corrosion on structural capacity of pipes than that on the prediction of their service life. This paper presents a methodology to quantitatively assess the risk of pipe collapse and predict its remaining service life using a time-dependent reliability theory. The concept of stress intensity in fracture mechanics is employed to establish the failure criterion of pipe collapse. An empirical model is derived for maximum pit growth of corrosion from the available data based on mathematical regressions. An example is provided to illustrate the application of the proposed method. It is found in the paper that the risk of pipe collapse increases with an increase in the diameter of the pipe for both external and internal corrosion. It is also found that the tougher the pipe is, the smaller the risk of its collapse. The paper concludes that a time-dependent reliability method is a very useful tool to predict the risk of pipe collapse and its remaining service life. The proposed method can help the water industry develop rehabilitation or replacement strategy for existing pipe networks with a view for better management of the pipe asset

Life Management of high energy piping girth welds

International Nuclear Information System (INIS)

Cohn, M.J.; Paterson, S.R.

1994-01-01

Life management of high energy piping systems is a synergistic process that combines the collective results from nondestructive examination (NDE), stress analysis, metallurgical replication, and fracture mechanics evaluations. To achieve conclusions with high confidence and reliability, the methodology requires that: (1) all weldments must be appropriately examined to establish initial baseline data and indicate both fabrication and inservice damage, (2) as-found stress analyses must be consistent with field estimated displacements to select sites of maximum in-service damage, and (3) metallurgical replicas must be taken at high stress sites determined as an outcome of a life exhaustion evaluation. The multidiscipline tasks are effectively managed by developing a rational framework incorporating all of the above requirements. Analytical algorithms that estimate linear and nonlinear degradation effects are included in the remaining life predictions. High damage locations are selected based on a life exhaustion calculation. Localized remaining life and reexamination intervals for failures governed by creep and fatigue damage are estimated by metallurgical replica cavitation damage and the service life. Stress-based creep damage identified in metallurgical replicas should correspond to the predicted high damage locations from the life exhaustion calculation. The methodology minimizes future reexamination locations while providing high certainty of monitoring lead-the-fleet damage. This life management program recognizes the aging process in plant equipment. It establishes a continuous process of examinations, evaluations, and decisions to track degradation of the piping system life cycle. Potential problems are identified long before failures occur. Corrective action is taken during scheduled outages to maintain the required level of plant performance

Evaluation of Fracture Resistance in Root Canal-Treated Teeth ...

African Journals Online (AJOL)

2018-06-11

Jun 11, 2018 ... Objective: This study aimed to evaluate the effects of different coronal restoration techniques on fracture ... Therefore, the application of horizontal fiber posts in. MOD cavities has .... Table 1: Median and 25% and 75% quartile values of the groups n ..... restorations on fracture resistance and failure mode of.

Application of LBB to a nozzle-pipe interface

Energy Technology Data Exchange (ETDEWEB)

Yu, Y.J.; Sohn, G.H.; Kim, Y.J. [and others

1997-04-01

Typical LBB (Leak-Before-Break) analysis is performed for the highest stress location for each different type of material in the high energy pipe line. In most cases, the highest stress occurs at the nozzle and pipe interface location at the terminal end. The standard finite element analysis approach to calculate J-Integral values at the crack tip utilizes symmetry conditions when modeling near the nozzle as well as away from the nozzle region to minimize the model size and simplify the calculation of J-integral values at the crack tip. A factor of two is typically applied to the J-integral value to account for symmetric conditions. This simplified analysis can lead to conservative results especially for small diameter pipes where the asymmetry of the nozzle-pipe interface is ignored. The stiffness of the residual piping system and non-symmetries of geometry along with different material for the nozzle, safe end and pipe are usually omitted in current LBB methodology. In this paper, the effects of non-symmetries due to geometry and material at the pipe-nozzle interface are presented. Various LBB analyses are performed for a small diameter piping system to evaluate the effect a nozzle has on the J-integral calculation, crack opening area and crack stability. In addition, material differences between the nozzle and pipe are evaluated. Comparison is made between a pipe model and a nozzle-pipe interface model, and a LBB PED (Piping Evaluation Diagram) curve is developed to summarize the results for use by piping designers.

Development of testing system for the thermo-mechanical fatigue crack analysis of nuclear power plant pipes

International Nuclear Information System (INIS)

Lee, Ho Jin; Kim, Maan Won; Lee, Bong Sang

2003-12-01

Fatigue crack growth analysis plays an important role in the structural integrity assessment or the service life calculation of the nuclear power plant pipes. To obtain the material properties as a basic data to achieve an accurate crack growth analysis, a lot of tests and numerical crack growth simulations have been done for decades. The BS 7910 or the ASME Boiler and Pressure Vessel Code Section XI, generally used to evaluate crack growth behavior, were made under the based on simple stress states or at the evaluated isothermal temperature. It is well known that the ASME code could sometimes give so conservative results in some cases of which the cracked components are experiencing with cyclic thermal shock. In this report, we suggested a method for the life assessment of a crack embedded in nuclear power plant pipes under the thermal-mechanical fatigue loads. We here use the numerical method to get the temperature history for thermal- mechanical fatigue crack growth test. And then we can calculate the remaining life time of the pipe by using the fracture mechanics and the test results together. For this purpose, we constructed a thermal-mechanical fatigue crack growth testing system. We also gave a lot of review about recent researches in the experimental field of thermal-mechanical fatigue analysis

Research program plan: piping. Volume 3

International Nuclear Information System (INIS)

Vagins, M.; Strosnider, J.

1985-07-01

Regulatory issues related to piping can be divided into the three areas of pipe cracking, postulated design basis pipe breaks, and design of piping for seismic and other dynamic loads. The first two of these issues are in the domain of the Materials Engineering Branch (MEBR), while the last of the three issues is the responsibility of the Mechanical/Structural Engineering Branch. This volume of the MEBR Research Plan defines the critical aspects of the pipe cracking and postulated design basis pipe break issues and identifies those research efforts and results necessary for their resolution. In general, the objectives of the MERB Piping Research Program are to provide experimentally validated analytic techniques and appropriate material properties characterization methods and data to support regulatory activities related to evaluating and ensuring piping integrity

Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report. [BWR

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

1982-04-01

This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 5. Summary - Piping Review Committee conclusions and recommendations

International Nuclear Information System (INIS)

1985-04-01

This document summarizes a comprehensive review of NRC requirements for Nuclear Piping by the US NRC Piping Review Committee. Four topical areas, addressed in greater detail in Volumes 1 through 4 of this report, are included: (1) Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants; (2) Evaluation of Seismic Design; (3) Evaluation of Potential for Pipe Breaks; and (4) Evaluation of Other Dynamic Loads and Load Combinations. This volume summarizes the major issues, reviews the interfaces, and presents the Committee's conclusions and recommendations for updating NRC requirements on these issues. This report also suggests research or other work that may be required to respond to issues not amenable to resolution at this time

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.

Fracture analysis procedure for cast austenitic stainless steel pipe with an axial crack

International Nuclear Information System (INIS)

Kamaya, Masayuki

2012-01-01

Since the ductility of cast austenitic stainless steel pipes decreases due to thermal aging embrittlement after long term operation, not only plastic collapse failure but also unstable ductile crack propagation (elastic-plastic failure) should be taken into account for the structural integrity assessment of cracked pipes. In the fitness-for-service code of the Japan Society of Mechanical Engineers (JSME), Z-factor is used to incorporate the reduction in failure load due to elastic-plastic failure. However, the JSME code does not provide the Z-factor for axial cracks. In this study, Z-factor for axial cracks in aged cast austenitic stainless steel pipes was derived. Then, a comparison was made for the elastic-plastic failure load obtained from different analysis procedures. It was shown that the obtained Z-factor could derive reasonable elastic-plastic failure loads, although the failure loads were more conservative than those obtained by the two-parameter method. (author)

Analysis of the failure performance of internally pressurized piping with surface flaws

International Nuclear Information System (INIS)

Iorio, A.F; Crespi, J.C.

1987-01-01

Due to frequent failures an Atucha I PHWR moderator circuit branch piping, made of stainless steel type AISI 347 (DIN 1.4550), studies have been made, involving the application of several fracture mechanics criteria, in order to determine the conditions of leak-before-break (L.BB) and the critical crack length of the piping. These studies lead to the conclusions that, for a straight pipe of outer diameter of 219 mm and 16 mm wall thickness, with a circumferential flaw and the principal stress being in the bending, the L.BB criteria are satisfied, being the critical crack length of the order of 400 mm. A better mechanical finishing and heat treatment was suggested in order to improve the resistance to crack initiation. (Author)

Evaluation of residual stresses for the multipass welds of 316L stainless steel pipe

International Nuclear Information System (INIS)

Kim, S. H.; Joo, Y. S.; Lee, J. H.

2003-01-01

It is necessary to evaluate the influence of the residual stress and distortion in the design and fabrication of welded structure and the sound welded structure can be maintained by this consideration. Multipass welds of the 316L stainless steel have been widely employed in the pipes of Liquid Metal Reactor. In this study, the residual stresses in the 316L stainless steel pipe welds were calculated by the finite element method using ANSYS code. Also, the residual stresses both on the surface and in the interior of the thickness were measured by HRPD(High Resolution Powder Diffractometer) instrumented in HANARO Reactor. The residual stresses were measured for each 18 points in small(t/d=0.075) and large pipe specimens (t/d=0.034). The experimental and calculated results were compared and the characteristics of the distribution of the residual stress discussed

A benchmark on computational simulation of a CT fracture experiment

International Nuclear Information System (INIS)

Franco, C.; Brochard, J.; Ignaccolo, S.; Eripret, C.

1992-01-01

For a better understanding of the fracture behavior of cracked welds in piping, FRAMATOME, EDF and CEA have launched an important analytical research program. This program is mainly based on the analysis of the effects of the geometrical parameters (the crack size and the welded joint dimensions) and the yield strength ratio on the fracture behavior of several cracked configurations. Two approaches have been selected for the fracture analyses: on one hand, the global approach based on the concept of crack driving force J and on the other hand, a local approach of ductile fracture. In this approach the crack initiation and growth are modelized by the nucleation, growth and coalescence of cavities in front of the crack tip. The model selected in this study estimates only the growth of the cavities using the RICE and TRACEY relationship. The present study deals with a benchmark on computational simulation of CT fracture experiments using three computer codes : ALIBABA developed by EDF the CEA's code CASTEM 2000 and the FRAMATOME's code SYSTUS. The paper is split into three parts. At first, the authors present the experimental procedure for high temperature toughness testing of two CT specimens taken from a welded pipe, characteristic of pressurized water reactor primary piping. Secondly, considerations are outlined about the Finite Element analysis and the application procedure. A detailed description is given on boundary and loading conditions, on the mesh characteristics, on the numerical scheme involved and on the void growth computation. Finally, the comparisons between numerical and experimental results are presented up to the crack initiation, the tearing process being not taken into account in the present study. The variations of J and of the local variables used to estimate the damage around the crack tip (triaxiality and hydrostatic stresses, plastic deformations, void growth ...) are computed as a function of the increasing load

Clinical diagnostic evaluation for scaphoid fractures: a systematic review and meta-analysis

NARCIS (Netherlands)

Mallee, Wouter H.; Henny, Erik P.; van Dijk, C. Niek; Kamminga, Sjoerd P.; van Enst, Wynanda A.; Kloen, Peter

2014-01-01

To provide an overview of available clinical evaluation tests for scaphoid fractures and to compare their diagnostic accuracies. PWe performed a systematic review of all studies assessing diagnostic characteristics of clinical evaluation in scaphoid fractures by searching MEDLINE, EMBASE, Cochrane,

Formation evaluation of fractured basement, Cambay Basin, India

International Nuclear Information System (INIS)

Gupta, Saurabh Datta; Farooqui, M Y; Chatterjee, Rima

2012-01-01

Unconventional reservoirs such as fractured basalts, shale gas and tight sand are currently playing an important role in producing a significant amount of hydrocarbon. The Deccan Trap basaltic rocks form the basement of the Cambay Basin, India, and hold commercially producible hydrocarbon. In this study two wells drilled through fractured basalts are chosen for evaluating the lithology, porosity and oil saturation of the reservoir sections. Well logs, such as gamma ray, high resolution resistivity, litho density, compensated neutron and elemental capture spectroscopy, have been used in cross-plotting techniques for lithology and mineral identification. Formation micro imagery log data have been analysed to quantify the fractures and porosity in the fractured reservoirs for a well in the south Ahmedabad block of the Cambay Basin. The results of the analysis of two wells are presented and discussed and they are found to be in good agreement with geological and production data. (paper)

Pipe damping studies

International Nuclear Information System (INIS)

Ware, A.G.

1986-01-01

The Idaho National Engineering Laboratory (INEL) is conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. This paper describes four tasks in the program that were undertaken in FY-86. In the first task, tests were conducted on a 5-in. INEL laboratory piping system and data were analyzed from a 6-in. laboratory system at the ANCO Engineers facility to investigate the parameters influencing damping in the seismic frequency range. Further tests were conducted on 3- and 5-in. INEL laboratory piping systems as the second task to determine damping values representative of vibrations in the 33 to 100 Hz range, typical of hydrodynamic transients. In the third task a statistical evaluation of the available damping data was conduted to determine probability distributions suitable for use in probabilistic risk assessments (PRAs), and the final task evaluated damping data at high strain levels

Evaluation of aluminum drill-pipe material and design

Energy Technology Data Exchange (ETDEWEB)

Placido, Joao C. [PETROBRAS, Rio de Janeiro, RJ (Brazil); Lourenco, Marcelo I.; Netto, Theodoro Antoun [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE)

2008-07-01

Experimental program and numerical analyses were carried out to investigate the fatigue mechanisms of aluminum drill pipes designed and manufactured in compliance with ISO 15546. The main objective is to improve the fatigue performance of these components by selecting the appropriate aluminum alloy and by enhancing the mechanical design of the threaded steel connector. This paper presents the experimental test program and numerical analyses conducted on a drill-pipe of different materials (Al-Cu-Mg and Al-Zn-Mg system aluminum alloys) and geometry. Material mechanical properties, including S-N curve, were determined through small-scale tests on specimens cut from actual drill pipes. Full-scale experiments were also performed in laboratory. A finite element model of the drill pipe, including the tool-joint region, was developed. The model simulates, through different load steps, the tool-joint hot assembly, and then reproduces the physical experiments numerically in order to obtain the actual stress distribution. Good correlation between full-scale and small-scale fatigue tests was obtained by adjusting the strain/stress levels monitored in the full-scale tests in light of the numerical simulations and performing fatigue life calculations via multiaxial fatigue models. The weak points of the current practice design are highlighted for further development. (author)

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

2002-03-15

This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

The characteristic investigation on narrow-gap TIG weld joint of heavy wall austenitic stainless steel pipe

International Nuclear Information System (INIS)

Shim, Deog Nam; Jung, In Cheol

2003-01-01

Although Gas Tungsten Arc Welding (GTAW or TIG welding) is considered as high quality and precision welding process, it also has demerit of low melting rate. Narrow-gap TIG welding which has narrow joint width reduces the groove volume remarkably, so it could be shorten the welding time and decrease the overall shrinkage in heavy wall pipe welding. Generally narrow-gap TIG welding is used as orbital welding process, it is important to select the optimum conditions for the automatic control welding. This paper looks at the application and metallurgical properties on narrow-gap TIG welding joint of heavy wall large austenitic stainless steel pipe to determine the deposition efficiency, the resultant shrinkage and fracture toughness. The fracture toughness depends slightly on the welding heat input

Fabrication and evaluation of chemically vapor deposited tungsten heat pipe.

Science.gov (United States)

Bacigalupi, R. J.

1972-01-01

A network of lithium-filled tungsten heat pipes is being considered as a method of heat extraction from high temperature nuclear reactors. The need for material purity and shape versatility in these applications dictates the use of chemically vapor deposited (CVD) tungsten. Adaptability of CVD tungsten to complex heat pipe designs is shown. Deposition and welding techniques are described. Operation of two lithium-filled CVD tungsten heat pipes above 1800 K is discussed.

Evaluation of fracture toughness for metal/ceramics composite materials by means of miniaturized specimen technique

International Nuclear Information System (INIS)

Saito, Masahiro; Takahashi, Hideaki; Jeong, Hee-Don; Kawasaki, Akira; Watanabe, Ryuzo

1991-01-01

In order to evaluate fracture strength for Y 2 O 3 -ZrO 2 , 3 mol% Y 2 O 3 -ZrO 2 (PSZ)/SUS 304 composite materials, Macor as a machinable ceramics and comercially available ceramics (SiC, Si 3 N 4 , PSZ, Al 2 O 3 ), fracture toughness tests were carried out by use of RCT or bending specimens. On the other hand, the fracture strength of these materials was evaluated and inspected the correlation between fracture toughness and fracture stress of small punch (SP) or modified small punch (MSP) test data to predict the fracture toughness value by using miniaturized specimens. Characteristic of the MSP testing method is the ability to evaluate elastic modulus (Young's modulus), fracture strength, yield strength, fracture strain, and fracture energy, etc., with high accuracy and good reproducibility for brittle materials. For a series of metal/ ceramics composites which from ductile to brittle, this paper clarified clear the applicable range for SP and MSP testing methods, which suggested that the simultaneous use of SP and MSP test methods can evaluate the fracture strength of metal/ ceramics composites. (author)

Evaluation of isolation valve leakage in alternate charging piping

International Nuclear Information System (INIS)

Strauch, P.L.; Roarty, D.H.; Brice-Nash, R.L.

1995-01-01

The chemical and volume control system (CVCS) alternate charging flow path at an operating pressurized water reactor (PWR) plant was determined to be susceptible to adverse stresses from isolation valve leakage. Isolation valve leakage had resulted in pipe cracks at several nuclear units worldwide, as described in United States Nuclear Regulatory Commission Bulletin 88-08 and its supplements. To provide for continuing assurance that cracks would not initiate over the plant life, the operators considered performing fatigue evaluation to demonstrate structural integrity of the system. This evaluation included heat transfer, stress and fatigue analysis, using methods described in Electric Power Research Institute Report ''Thermal Stratification, Cycling, and Striping (TASCS),'' March 1994. The evaluation concluded that the fatigue usage would be less than 1.0 under worst case isolation valve leakage conditions, and therefore a significant investment in permanent temperature monitoring was avoided

Laboratory experiments on heat-drive two-phase flows in natural and artificial rock fractures

International Nuclear Information System (INIS)

Kneafsey, Timothy J.; Pruess, Karsten

1998-01-01

Water flow in partially saturated fractures under thermal drive may lead to fast flow along preferential localized pathways and heat pipe conditions. At the potential high-level nuclear waste repository at Yucca Mountain, water flowing in fast pathways may ultimately contact waste packages and transport radionuclides to the accessible environment. Sixteen experiments were conducted to visualize heat-driven liquid flow in fracture models that included (1) assemblies of roughened glass plates, (2) epoxy replicas of rock fractures, and (3) a fractured specimen of Topopah Spring tuff. Continuous rivulet flow was observed for high liquid flow rates, intermittent rivulet flow and drop flow for intermediate flow rates, and film flow for lower flow rates and wide apertures. Heat pipe conditions (vapor-liquid counterflow with phase change) were identified in five of the seven experiments in which spatially resolved thermal monitoring was performed but not when vapor-liquid counterflow was hindered by very narrow apertures and when an inadequate working fluid volume was used

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

International Nuclear Information System (INIS)

Hong, S.Y.; Yeater, M.L.

1985-01-01

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

Investigation and evaluation of the cause about the loss of split pin which was set to pipe support system of the primary cooling system of Monju

International Nuclear Information System (INIS)

Ichikawa, Shoichi; Kawanago, Sho; Nishio, Ryuichi; Wakimoto, Fumitsugu; Fujimura, Tomofumi; Kobayashi, Takanori; Sakamoto, Tsutomu

2015-07-01

The loss of the retaining split pins (four pieces) for clevis pin was confirmed at the inspection of the pipe supports in the Monju prototype fast-breeder reactor (hereinafter referred to Monju) in May, 2014. The split pins (two pieces) of ROD RESTRAINT and CONSTANT HANGER were fallen off. The split pins (two pieces) of MECHANICAL SNUBBER were broken at both ends of them. As a result of investigation, a dimple pattern was observed in a fracture surface of broken split pin. This observation result showed that fracture morphology is ductile fracture. A reproduction test, whether split pin was broken by loading the external force to the clevis pin, also gave the same fracture morphology. As the result of all cause investigation, the reason of the broken split pins is that the split pins were loaded shearing stress by the external force loaded to the clevis pin axial direction. The result of the cause investigation and a recurrence prevention measure of this trouble was reported. (author)

Heat pipe development

Science.gov (United States)

Bienart, W. B.

1973-01-01

The objective of this program was to investigate analytically and experimentally the performance of heat pipes with composite wicks--specifically, those having pedestal arteries and screwthread circumferential grooves. An analytical model was developed to describe the effects of screwthreads and screen secondary wicks on the transport capability of the artery. The model describes the hydrodynamics of the circumferential flow in triangular grooves with azimuthally varying capillary menisci and liquid cross-sections. Normalized results were obtained which give the influence of evaporator heat flux on the axial heat transport capability of the arterial wick. In order to evaluate the priming behavior of composite wicks under actual load conditions, an 'inverted' glass heat pipe was designed and constructed. The results obtained from the analysis and from the tests with the glass heat pipe were applied to the OAO-C Level 5 heat pipe, and an improved correlation between predicted and measured evaporator and transport performance were obtained.

Generation of plastic influence functions for J-integral and crack opening displacement of thin-walled pipes with a short circumferential through-wall crack

International Nuclear Information System (INIS)

Park, Jeong Soon; Choi, Young Hwan; Im, Seyoung

2014-01-01

Fracture mechanics parameters such as the J-integral and crack opening displacement (COD), are necessary for Leak-Before-Break (LBB) evaluation. The famous two estimation methods, the GE/EPRI and the Reference Stress Method (RSM), have their applicability limit with regard to the ratio of a pipe mean radius to thickness (R m /t). In order to extend their applicability limit to a thin walled pipe, several finite element analyses are performed for the J-integral and COD, and then new plastic influence functions are developed for thin-walled pipes with a short circumferential through-wall crack. With the newly generated plastic influence functions, the GE/EPRI and the RSM give closer results with those obtained from detailed finite element analyses. In addition, C*-integral and COD rate are estimated by using the new plastic influence functions and they are well matched with elastic–creep finite element analysis results under the power-law creep condition. Since the LBB concept can be applied to a piping system in a Korean Sodium-cooled Fast Reactor (SFR) which is designed to have thin-walled pipes and to operate in high temperature enough to cause creep, this paper can be applied for the LBB assessment of thin-walled pipes with a short through-wall crack in the SFR

Report of examination of the ruptured pipe at the Hamaoka Nuclear Power Station Unit-1

International Nuclear Information System (INIS)

2001-12-01

In order to investigate root cause of the pipe rupture, which took place at the Hamaoka Nuclear Power Station Unit-1 of Chubu Electric Power Company on November 7, 2001, a task force was established within the Nuclear and Industrial Safety Agency (NISA) and initiated a detailed investigation of the ruptured pipe. The Japan Atomic Energy Research Institute (JAERI) was asked from the Ministry of Education, Culture, Sports, Science and Technology (MEXT) in response to the request from NISA to cooperate as an independent neutral organization with NISA and perform an examination of the ruptured pipe independently from Chubu Electric Power Company. JAERI accepted the request by considering the fact that JAERI is an integrated research institution for nuclear research and development, a prime research institution for nuclear safety research, a research institution with experience of root-cause investigation of various nuclear incidents and accidents of domestic as well as overseas, and a research institution provided with advanced examination facilities necessary for examination of the ruptured pipe. The JAERI examination group was formed at the Tokai Research Establishment and conducted detailed and thorough examination of the pieces taken from the ruptured pipe primarily in the Reactor Fuel Examination Facility (RFEF) with the use of tools such as scanning electron microscopes and other equipments. Purpose of examination was to provide technical information in order to identify causes of the pipe rupture through examination of the pieces taken from the ruptured region of the pipe. The following findings and conclusion were made as the result of the present examination. (1) Wall thickness of the pipe was significantly reduced in the ruptured region. (2) Dimple pattern resulting from ductile fracture by shearing was observed in the fracture surfaces of nearly all of the pieces and no indication of fatigue crack growth was found. (3) Microstructure showed a typical carbon

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)

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 4. Evaluation of other loads and load combinations

International Nuclear Information System (INIS)

1984-12-01

Six topical areas were covered by the Task Group on Other Dynamic Loads and Load Combinations as described below: Event Combinations - dealing with the potential simultaneous occurrence of earthquakes, pipe ruptures, and water hammer events in the piping design basis; Response Combinations - dealing with multiply supported piping with independent inputs, the sequence of combinations between spacial and modal components of response, and the treatment of high frequency modes in combination with low frequency modal responses; Stress Limits/Dynamic Allowables - dealing with inelastic allowables for piping and strain rate effects; Water Hammer Loadings - dealing with code and design specifications for these loadings and procedures for identifying potential water hammer that could affect safety; Relief Valve Opening and Closing Loads - dealing with the adequacy of analytical tools for predicting the effects of these events and, in addition, with estimating effective cycles for fatigue evaluations; and Piping Vibration Loads - dealing with evaluation procedures for estimating other than seismic vibratory loads, the need to consider reciprocating and rotary equipment vibratory loads, and high frequency vibratory loads. NRC staff recommendations or regulatory changes and additional study appear in this report

Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

International Nuclear Information System (INIS)

Lowry, W.

1994-01-01

The objective for the development of the Pipe Explorer trademark radiological characterization system is to achieve a cost effective, low risk means of characterizing gamma radioactivity on the inside surface of pipes. The unique feature of this inspection system is the use of a pneumatically inflated impermeable membrane which transports the detector into the pipe as it inverts. The membrane's internal air pressure tows the detector and tether through the pipe. This mechanism isolates the detector and its cabling from the contaminated surface, yet allows measurement of radioactive emissions which can readily penetrate the thin plastic membrane material (such as gamma and high energy beta emissions). In Phase 1, an initial survey of DOE facilities was conducted to determine the physical and radiological characteristics of piping systems. The inverting membrane deployment system was designed and extensively tested in the laboratory. A range of membrane materials was tested to evaluate their ruggedness and deployment characteristics. Two different sizes of gamma scintillation detectors were procured and tested with calibrated sources. Radiation transport modeling evaluated the measurement system's sensitivity to detector position relative to the contaminated surface, the distribution of the contamination, background gamma levels, and gamma source energy levels. In the culmination of Phase 1, a field demonstration was conducted at the Idaho National Engineering Laboratory's Idaho Chemical Processing Plant. The project is currently in transition from Phase 1 to Phase 2, where more extensive demonstrations will occur at several sites. Results to date are discussed

Relevance of adjacent joint imaging in the evaluation of ankle fractures.

Science.gov (United States)

Antoci, Valentin; Patel, Shaun P; Weaver, Michael J; Kwon, John Y

2016-10-01

Routinely obtaining adjacent joint radiographs when evaluating patients with ankle fractures may be of limited clinical utility and an unnecessary burden, particularly in the absence of clinical suspicion for concomitant injuries. One thousand, three hundred and seventy patients who sustained ankle fractures over a 5-year period presenting to two level 1 trauma centers were identified. Medical records were retrospectively reviewed for demographics, physical examination findings, and radiographic information. Analyses included descriptive statistics along with sensitivity and predictive value calculations for the presence of adjacent joint fracture. Adjacent joint imaging (n=1045 radiographs) of either the knee or foot was obtained in 873 patients (63.7%). Of those, 75/761 patients (9.9%) demonstrated additional fractures proximal to the ankle joint, most commonly of the proximal fibula. Twenty-two of 284 (7.7%) demonstrated additional fractures distal to the ankle joint, most commonly of the metatarsals. Tenderness to palpation demonstrated sensitivities of 0.92 and 0.77 and positive predictive values of 0.94 and 0.89 for the presence of proximal and distal fractures, respectively. Additionally, 19/22 (86.4%) of patients sustaining foot fractures had their injury detectable on initial ankle X-rays. Overall, only 5.5% (75/1370) of patients sustained fractures proximal to the ankle and only 0.2% (3/1370) of patients had additional foot fractures not evident on initial ankle X-rays. The addition of adjacent joint imaging for the evaluation of patients sustaining ankle fractures is low yield. As such, patient history, physical examination, and clinical suspicion should direct the need for additional X-rays. Level IV. Copyright © 2016 Elsevier Ltd. All rights reserved.