Key Engineering Materials Vols. 297-300

Abstract: The polycrystallined LaB6-ZrB2 composites with different content of ZrB2 were fabricated by vacuum hot-pressed sintering technique in this paper. The effects of ZrB2 content on the microstructure and mechanical properties were investigated. For the eutectic LaB6-ZrB2 (21wt%) composite, the hardness, flexural strength and fracture toughness is of 93.0 (HRA), 330.0MPa and 3.70MPa·m1/2, respectively. The results showed the hardness and flexural strength of LaB6-ZrB2 polycrystalline have been enhanced with the increasing of ZrB2 content, but the fracture toughness increases first then arrives at peak value, which is corresponding to the ZrB2 content of 21wt%. The microstructure observation revealed an improved densification due to addition of ZrB2. The fracture morphology showed a tendency of the fracturing from intergranular to transgranular manner with increasing the ZrB2 content. The fracture behavior of the composites was analyzed in the paper

Abstract: The purpose of this paper is to evaluate the piping failure frequency based on the piping failure events in Korean pressurized water reactors (PWRs) until the end of 2003. Two types of the piping failure frequencies including the piping damage frequency and the piping rupture frequency are considered in this study. The piping damage frequency for the failed piping system was estimated by using the piping population data such as the weld count or the base metal count. The piping rupture frequency related to the initiating event in a probabilistic safety assessment (PSA) was evaluated by using both the Bayesian approach (Method 1) and the conditional rupture probability approach (Method 2). In the Bayesian approach, two methods using Jeffreys noninformative prior (Method 1-1) and prior distributions based on the results in NUREG/CR-5750 (Method 1-2) were considered. Thirty piping failure events in ASME safety class pipings of Korean PWRs were identified and analyzed in this study. The results showed that the piping damage frequency for the events ranged from 5.42E-3/cr.yr to 2.77E-5/cr.yr. Three kinds of initiating events including the very small LOCA, the feedwater line break, and the flood are evaluated for Korean PWRs. The results for the piping rupture frequency in Korean PWRs were as follows: 1) The mean piping rupture frequency of the very small LOCA event ranged from 3.6E-3/cr.yr to 1.2E-2/cr.yr, the feedwater line break event from 3.6E-3/cr.yr to 2.5E-2/cr.yr, and the flood event from 7.8E-4/cr.yr to 3.6E-3/cr.yr. The mean piping rupture frequencies of the very small LOCA and feedwater line break events were higher than that of the flood event by one order of a magnitude. 2) Method 2 gave conservative results in the very small LOCA and feedwater line break events compared to Method 1-1 or Method 1-2, while Method 1-1 gave conservative results in the flood event. 3) The order of magnitudes in the mean piping rupture frequencies of the very small LOCA, the feedwater line break, and the flood in Korean PWRs were similar to those in the U.S. PWRs.

Abstract: Local failure modes associated with bottom-mounted penetration nozzles are examined as a part of research on sever accident management. Conventional creep rupture studies on reactor vessel lower head during a meltdown accident were based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failures is investigated using data and nozzle materials from Sandia National Laboratory’s Lower Head Failure Experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic-viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIP. It has been concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure with its likelihood significantly greater than previously assumed.

Abstract: Class 1 piping components of a certain old vintage nuclear power plant were designed by ANSI B31.1 code without a detailed fatigue evaluation such as the one required by recent ASME Section III code. These components may undergo fatigue damage when considering the continued operation beyond the design life whilst the inherent fatigue resistances of those may satisfy the corresponding implicit limits. In this paper, the alternative fatigue evaluation has been carried out explicitly for Class 1 piping of old nuclear power plant. At first, four representative nuclear piping systems were selected to check the operational adequacy. After characterization of conservative loading conditions based on design features, a series of finite element analyses have been performed and the cumulative usage factors were calculated to guarantee if the components at each system sustain adequate fatigue resistance. Finally, comparisons were drawn between the implicit fatigue design specifications and alternative explicit fatigue analysis results. Even though there were some exceptions, it was demonstrated that most components satisfied the current explicit fatigue criterion.

Abstract: The cask is used to transport the radioactive materials. It is required to withstand for the thirty minute under the hypothetical fire accident condition of the 800ı. According to development of the computer simulation, finite element analysis is applied to the calculation widely. But finite element method for the hypothetical accident conditions is not established in domestic regulations. In this study, the temperature and thermal stress analysis of KSC-4 cask under 800ı fire condition is conducted using by ANSYS 7.0 code. In order to analyze finite elements, two-dimensional model of KSC-4 cask is used. Symmetric boundary, convection, and radiation condition are applied in the analysis. As the results, maximum temperature and thermal stress of the KSC-4 cask is evaluated.

Abstract: The effects of the microstructural parameters, such as the prior austenite grain size and carbide size, on the cleavage fracture toughness were investigated in the transition region of Mn-Mo-Ni bainitic low alloy steels. Cleavage fracture toughness was evaluated by the ASTM standard E 1921 Master curve method. In order to clarify the effects of each microstructure, the grain size and carbide size of the test materials were independently controlled by modifying the heat treatment process. Firstly, the grain sizes were changed from 25㎛ to 110㎛ without any significant changes in the carbide size and shape. Secondly, the average carbide sizes were changed from 0.20 ㎛ to 0.29㎛ but maintaining the initial grain sizes. As a result, the fracture toughness in the transition region did not show any significant dependency on the austenite grain size, while the carbide size showed a close relation to the fracture toughness. Fracture toughness was decreased with an increase of the average carbide size. From the microscopic observation of the fractured surface, the cleavage initiation distance (CID) from the original crack tip showed no direct relationship to the prior austenite grain sizes but a strong relationship to the carbide sizes. However, the measured cleavage fracture toughness was strongly related to the distance from the crack tip to the cleavage initiation site. From the viewpoint of the weakest link theory, the particle size and their distribution in front of the crack tip is probably more important than the grain size in the transition temperature range where the fracture was controlled by the cleavage crack initiation.

Abstract: The ASME B & PV Code Sec. allows the socket weld for the nuclear piping in spite of the weakness on the weld integrity. Recently, the integrity of the socket weld is regarded as a safety concern in nuclear power plants because many failures and leaks have been reported in the socket weld. OPDE (OECD Piping Failure Data Exchange) database lists 108 socket weld failures among 2,399 nuclear piping failure cases during 1970 to 2001. Eleven failures in the socket weld were also reported in Korean NPPs. Many failure cases showed that the root cause of the failure is the fatigue and the gap requirement for the socket weld given in ASME Code was not satisfied. The purpose of this paper is to evaluate the fatigue crack behavior of a surface crack in the socket weld under fatigue loading condition considering the gap effect. Three-dimensional finite element analysis was performed to estimate the fatigue crack behavior of the surface crack. Three types of loading conditions such as the deflection due to vibration, the pressure transient ranging from P=0 to 15.51MPa, and the thermal transient ranging from T=25oC to 288oC were considered. The results are as follows; 1) The socket weld is susceptible to the vibration where the vibration levels exceed the requirement in the ASME Operation and Maintenance (OM) Code. 2) The effect of pressure or Temperature transient load on the socket weld integrity is not significant. 3) No-gap condition gives very high possibility of the crack initiation at the socket weld under vibration loading condition. 4) For the specific systems having the vibration condition to exceed the requirement in the ASME Code OM and/or the transient loading condition from P=0 and T=25oC to P=15.51MPa and T=288oC, radiographic examination to examine the gap during the construction stage is recommended.

Abstract: The hydrogen concentration limit and critical temperatures for a delayed hydride cracking (DHC) in zirconium alloys have been reanalyzed using Kim’s DHC model that a driving force for DHC is not the stress gradient but the supersaturated hydrogen concentration or ∆C arising from a hysteresis of the terminal solid solubility on a heating and on a cooling. The DHC initiation occurs generally at the temperatures corresponding to the terminal solid solubility for precipititation (TSSP), demonstrating that the supercooling from the terminal solid solubility for dissolution (TSSD) is required to initiate the DHC. The DHC arrest temperatures correspond to the temperatures where the ∆C is reduced to zero. Therefore, we conclude that the ∆C is the driving force for the DHC and that the Kim’s DHC model is feasible.

Abstract: A dissimilar weld zone exists between the pipe and nozzle in a primary reactor cooling system (RCS). Thermal aging is observed in cast stainless steel, CF8M used in a pipe as the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330°C. No effect is observed in low-alloy steel. SA508 cl.3 is used in a nozzle. The artificially accelerated aging specimens are prepared to maintain for a temperature of 430°C for 300, 1800, and 3600hrs, respectively. Then, various mechanical tests such as hardness, tension, impact test, are performed in virgin and aged specimens in order to determine the existence of dissimilar weld zones. The specimens for elastic-plastic fracture toughness tests are prepared for one type, where a notch is created in the heat affected zone of CF8M. From the experiments, it was found that J-integral values decrease as age increases.