Static and cyclic crack growth of low alloy steels in high-oxygen water simulating HDR test conditions

Steel samples of reactor pressure vessel and piping steels from the German HDR programme have been tested in high oxygen water at different temperatures simulating HDR test conditions. The specimens have been exposed to sequences of static and cyclic loading or to purely cyclic loading. During the tests, threshold stress intensity values for stress corrosion cracking and crack growth rates with various cyclic loading parameters were determined. Extensive fracture surface and oxide layer investigations were also performed. Water chemistry parameters such as dissolved oxygen content, pH and conductivity were continuously monitored during the tests. Finally, the measured laboratory water chemistry parameters were compared to those measured in the HDR plant during full scale testing of components and the relevance of the results for normally operating plants is discussed.     

Pipe fracture safety analysis using limit load and J-integral techniques

J-integral fracture toughness tests were performed on full scale pipe specimens to assess the fracture safety performance of two reactor piping alloys. The two alloys investigated were ASTM A106 Grade B carbon steel and circumferentially welded Type 304 stainless steel.The full scale pipe fracture tests were performed on 1.2 m long, circumferentially cracked pipes loaded in four-point bending on a variably compliant test bed. Results of the experiments were analyzed using the limit load approach currently being considered for inclusion in Section XI of the ASME Code. The results were also evaluated using two tearing instability approaches. One approach assumed elastic-perfectly plastic material behavior and the other accounted for material hardening by requiring actual load and displacement data.The limit load analysis provided a good prediction of the maximum load carrying capacity of the pipe specimens in most cases. The results were especially good for the ASTM A106 steel pipes when the materials property data was used to calculate the flow stress. The J-integral tearing instability analysis was shown to accurately describe the ductile tearing instability behavior of the ASTM A106 steel pipe providing material hardening was taken into account.     

国产A508-3钢在模拟AP1000一回路水环境下的疲劳性能研究

反应堆压力容器（RPV）钢在一回路水环境下的疲劳性能是评价其设计寿命的重要参数。本文针对国产A508-3钢开展了模拟AP1000一回路水环境的低周疲劳性能试验研究，获得了321 ℃、155 MPa及01 ppm溶解氧水环境下的疲劳行为数据和断裂机理。研究结果表明，国产A508 3钢峰值应力随应变幅的增大而逐渐增大，疲劳试验过程中试样表现出循环硬化、循环软化和饱和3个阶段；在应变幅由02%逐渐增加至06%的过程中，疲劳周次从105逐渐降低至102；疲劳断口具有疲劳和腐蚀特征，属于典型的腐蚀疲劳断裂。     

Fatigue and fracture behavior of straight pipe with flaws in inner surface

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 AISI type 304L stainless steel were tested under almost the same test conditions by imposing moment loads. Three artificial defects were machined in the inner surface of the test section of the test pipes and the fatigue test was performed until the cracks coalesced and grew through the thickness. Subsequently, a static load was imposed on the test pipe which contained a large crack in the test section.The fatigue test results are compared with an analytical crack growth behavior predicted by the method described in the Section XI of ASME Code, and show slower crack growth than that of the prediction. From the fracture test results, it is found that the test pipes can endure considerably high load.     

SCC testing of pipe materials in BWR environment

A test loop has been installed in Ringhals 1 BWR, including facilities for Constant Elongation Rate Testing (CERT) and Electrochemical Potential (ECP) measurements in primary reactor water at reactor operation temperature. The loop is designed as to minimize transport time for reactor water from the reactor pressure vessel to the specimens being tested. Thus the testing environment is representative of the primary piping systems of BWRs, also with regard to short-lived constituents like hydrogen peroxide.The test program, which is in progress, has covered seven tests during start-up conditions or during power operation with presently current reactor water chemistry. In this presentation only CERT testing results on heavily sensitized austenitic chromium—nickel stainless steel are presented, although many other materials have been tested.Results show sensitized austenitic stainless steel is more prone to intergranular stress corrosion cracking (IGSCC) in actual than in simulated BWR environment and that start-up environment is chemically more aggressive than power operation environment. Reproducibility of the CERT technique as used is excellent.     

The behaviour of WB 35 piping material under dynamic loading in the high speed tensile test

For the determination of the strength-, deformation- and fracture behaviour of the material 17 MnMoV 6 4 (WB 35) which is used for piping components, tensile tests were carried out at different loading rates (monotonic and impact-type) on smooth and notched pipe strip specimens over a temperature range extending from − 30°C to 250°C.For the conduct of the tests a hydraulic high speed tensile machine having a free motion device was used; the velocity of impact was preset at ca. 7 m/s.With impact-type (dynamically) loaded specimens in general higher strength and deformation values were obtained than with monotonic (statically) loaded ones. In all of the specimens having low deformation values which were investigated microfractographically, ductile portions were found adjacent to the notch on the fracture surface.     

Preliminary results of post-irradiation examinations on LiSoR-2 test section

Liquid-solid reaction under irradiation (LiSoR) experiments are aimed at understanding the effects of liquid lead-bismuth eutectic (LBE) corrosion and embrittlement under irradiation on structural materials, which is one of the key items of the materials R&D for the future accelerator-driven system (ADS). The LiSoR setup is basically a LBE loop with a test section irradiated with 72 MeV protons. The second irradiation was conducted for about 34 h and terminated after a leakage of LBE was detected. Post-irradiation examinations (PIE) are being performed on both the tube and tensile specimen in the test section. Optical microscopy, scanning electron microscopy, transmission electron microscopy and microhardness tests have been completed. The results show that a crack formed in the irradiation zone of the tube. In the material in the irradiation zones of both the tube and the tensile specimen dislocation cell structure is well developed, which indicates heavy deformation due to thermal fatigue. The crack should start at the inner surface and propagate to the outer surface. The fracture surfaces of the crack are dominated by a brittle cleavage fracture mode. However, on the surfaces of the tensile specimen, no microcracks are observed.     

Towards an elastic-plastic fracture mechanics predictive capability for reactor piping

Intergranular stress corrosion cracks have been discovered in the recirculation bypass piping and core spray lines of several boiling water reactor (BWR) plants. These cracks initiate in heat-affected zones of girth welds and grow circumferentially by combined stress corrosion and fatigue. Reactor piping is mainly type 304 stainless steel, a material which exhibits high ductility and toughness. A test program described in this paper demonstrates that catastrophic crack growth in these materials is preceded by considerable amounts of stable crack growth accompanied by large plastic deformation. Thus, conventional linear elastic fracture mechanics, which only applies to the initiation of crack growth in materials behaving in a predominantly linear elastic fashion, is inadequate for a failure analysis of reactor piping.This paper is based upon research initiated by a need to develop a realistic failure prediction and a way to delineate leak-before-break conditions for reactor piping. An effective engineering solution for the type of cracks that have been discovered in BWR plants was first developed. This was based upon a simple net section flow stress criterion. Subsequent work to develop an elastic-plastic fracture mechanics methodology has also been pursued. A survey of progress being made is described in this paper. This work is based on the use of finite element models together with experimental results to identify criteria appropriate for the onset of crack extension and for stable crack growth. A number of criteria have been evaluated. However, the optimum fracture criterion has not yet been determined, even for conditions which do not include all of the complications involved in reactor piping.     

国产316LN不锈钢在模拟AP1000一回路水环境中的疲劳行为

一回路水环境下的疲劳性能是核电站主管道设计寿命评估的重要参数。针对国产主管道材料316LN开展了模拟AP1000一回路水环境的低周疲劳试验,分析了疲劳行为和失效机理。研究结果表明:国产316LN峰值应力随应变幅的增大而增大,大应变幅试样在疲劳过程中先后发生了循环硬化、循环软化和失稳,而小应变幅试样在失稳前未发生明显的循环硬化和循环软化;在应变幅由0.2%逐渐增加至1.2%的过程中,疲劳周次从105逐渐降低至102;疲劳断口具有典型的疲劳断口特征,裂纹萌生于试样表面,以穿晶方式垂直于主应力方向扩展,裂纹扩展区具有典型的疲劳辉纹,辉纹上有菱形颗粒状腐蚀产物,环境辅助开裂机制倾向于氢致开裂。     

Effect of test temperature on tensile and fatigue properties of nickel-base heat-resistant alloys

A series of tensile and strain controlled low-cycle fatigue tests were conducted at temperatures ranging from RT to 900°C on a nickel-base heat-resistant alloy, Hastelloy XR-II, which is one of the candidate alloys for applications in the process heating high-temperature gas-cooled reactor (HTGR). Fatigue tests at room temperature and all tensile tests were conducted in air, while fatigue tests at and above 400°C were conducted in the simulated HTGR helium environment. In those tests the effect of test temperature on tensile and fatigue properties was investigated. The ductility minimum point was observed near 600° C, while tensile and fatigue strengths decreased with increasing test temperature. The fatigue lives estimated with the method proposed by Manson were compatible with the experimental results under the given conditions. For the specimens fatigued at and above 700°C, the percentage of the intergranular fracture mode gradually increased with increasing test temperature.     

Precracked pipe under waterhammer action

We numerically simulate a full scale test in several computational steps with the finite element method and compare all calculated data with the experimental findings. First, we compute the deflection under static loading and the spectrum of eigenfrequencies of an integer piping, attached to a nuclear reactor pressure vessel (RPV). Then we consider a sudden pipe break at some distance from the vessel, immediately followed by an undamped closure of a check valve close to the break on the RPV side, and calculate the elastic and plastic transient dynamic response of the integer piping part between the RPV and the break. Finally, we consider a circumferential internal surface crack, fairly close to the vessel; after extensive testing of our fracture mechanics calculation procedure we investigate the stress in the crack region under the waterhammer action.     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.     

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

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.     

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.     

Observations on the effect of post-weld heat treatment on J-resistance curves of sa-106b seamless piping welds

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, J_lc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower J_Ic, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The J_Ic and J-resistance curves obtained by Ontario Hydro and MEA were comparable.     

A research program for dynamic fracture evaluation of Japanese carbon steel pipes

A research program was developed to investigate the dynamic load effect on fracture behavior of Japanese carbon steel STS410 pipe. The program comprises material tests, pipe fracture tests and development of estimation scheme. Material property tests showed that the flow stress was nearly constant or slightly increased with strain rate. Pipe tests showed that fracture load was nearly predicted by the net-section collapse criterion for both quasi-static and dynamic loading. Significant dynamic effect was not observed for STS410 carbon steel piping. Crack growth was well formulated by using J-integral parameter for low cycle fatigue with large scale yielding. Combining the crack growth behavior and unstable fracture criterion, an estimation scheme was newly developed and validated for constant amplitude cyclic loading conditions.     

Investigation of piping failure in the HDR test plant under operational conditions and the influence of oxygen content

The double ended failure of the piping system of the HDR-testing facility is described. The failure occurred in the transition region of a pipe reducer near the pressurizer. Theoretical (FE-analysis) and experimental studies (metallography, fractography) yield the conclusion that one main reason for the failure is strain induced corrosion (stable crack extension in corrosive environment). High strains resulted from internal pressure of 110 bar (operational pressure of the test plant) and from the highly underdimensioned wall thickness of the reducer due to a manufacturing deficiency. Corrosive environment in the loop is artificially prepared by injection of oxygen to 8 ppm. The conductivity of the medium in the loop amounted to 25 μS. Some crack growth and fracture patterns as in HDR-failure surfaces were simulated on CT-specimens in the MPA-autoclave testing facilities.     

Reactor safety research at the large scale facility HDR

The HDR Safety Program includes various full scale experiments to study the component behavior under operating conditions and accidental loads as blowdown, earthquake, aircraft impact, tthermal shock. These tests are conducted at the decomisssioned “superheated steam reactor” (Heiβdampfreaktor HDR) located near Frankfurt/Germany. The construction of the test facility HDR offers many test capabilities relevant to modern commercial nuclear power plants. Operating conditions typical of those for both pressurized- and boiling water reactors can be produced by a electrically-heated test circuit, specially installed for test purposes. The main components of interest include the reactor pressures vessel, core barrel, pipings and safety valves, reactor building and steel containment shell. The construction of the HDR test facility, the main objectives of the program, used calculation methods and summarized results will be described.     

Loading conditions in horizontal feedwater pipes of LWRs influenced by thermal shock and thermal stratification effects

In recent years, a particular form of crack formation has occurred in a number of pressurized and boiling water reactors on the internal surfaces of horizontal lengths of feedwater piping upstream of the steam generators and reactor pressure vessels.The fractographic evaluation and the orientation of the cracks show that these are to be attributed to cyclic stressing in the axial direction. Comparison of the stresses due to thermal shock and thermal stratification reveals that, on account of the associated load cycles, the cracks were in essence caused by thermal stratification. This is also indicated by the orientation of the cracks.The present results of the corrosion tests show that with high oxygen content (450 ppb) and temperature level (210°C) the strain rate at thermal stratification exerts an essential influence on the number of cycles to crack initiation. With the conservative test conditions described, the values may fall below those given in the ASME fatigue design curve. In the case of strain rates that apply to thermal shock, the cycles to crack initiation are on the safe side of the curve.The remedial measures taken by KWU by the installation of the siphon reduce the frequency of stress amplitudes. It can be concluded from corrosion tests simulating these conditions that with the strain rates occurring in this case, the number of cycles to crack initiation are on the safe side according to the ASME fatigue design curve.