Charles H. Hofmayer

Analysis of JNES Seismic Tests on Degraded Piping

Nuclear power plant safety under seismic conditions is an important consideration. The piping systems may have some defects caused by fatigue, stress corrosion cracking, etc., in aged plants. These cracks may not only affect the seismic response, but may also grow and break through causing loss of coolant. Therefore, an evaluation method needs to be developed to predict crack growth behavior under seismic excitation. This paper describes efforts conducted to analyze and better understand a series of degraded pipe tests under seismic loading that was conducted by Japan Nuclear Energy Safety Organization (JNES). A special “cracked-pipe element” (CPE) concept, where the element represented the global moment-rotation response due to the crack, was developed. This approach was developed to simplify the dynamic finite element analysis. In this paper, model validation was conducted by comparisons with a series of pipe tests with circumferential through-wall and surface cracks under different excitation conditions. These analyses showed that reasonably accurate predictions could be made using the ABAQUS connector element to model the complete transition of a circumferential surface crack to a through-wall crack under cyclic dynamic loading. The JNES combined-component test was analyzed in detail. The combined-component test had three crack locations and multiple applied simulated-seismic block loadings. Comparisons were also made between the ABAQUS FE analyses results to the measured displacements in the experiment. Good agreement was obtained and it was confirmed that the simplified modeling is applicable to a seismic analysis for a cracked pipe on the basis of fracture mechanics. Pipe system leakage did occur in the JNES tests. The analytical predictions using the CPE approach did not predict leakage, suggesting that cyclic ductile tearing with large-scale plasticity was not the crack growth mode for the acceleration excitations considered here. Hence, the leakage was caused by low-cycle fatigue with small-scale yielding. The procedure used to make predictions of low-cycle fatigue crack growth with small-scale yielding was based on the Dowling ΔJ procedure, which is an extension of linear-elastic fatigue crack growth methodology into the nonlinear plasticity regime. The predicted moments from the CPE approach were used using a cycle-by-cycle crack growth procedure. The predictions compare quite well with the experimental measurements.Copyright

SEISMIC RESPONSE PREDICTION OF NUPEC'S FIELD MODEL TESTS OF NPP STRUCTURES WITH ADJACENT BUILDING EFFECT.

As part of a verification test program for seismic analysis computer codes for Nuclear Power Plant (NPP) structures, the Nuclear Power Engineering Corporation (NUPEC) of Japan has conducted a series of field model tests to address the dynamic cross interaction (DCI) effect on the seismic response of NPP structures built in close proximity to each other. The program provided field data to study the methodologies commonly associated with seismic analyses considering the DCI effect. As part of a collaborative program between the United States and Japan on seismic issues related to NPP applications, the U.S. Nuclear Regulatory Commission sponsored a program at Brookhaven National Laboratory (BNL) to perform independent seismic analyses which applied common analysis procedures to predict the building response to recorded earthquake events for the test models with DCI effect. In this study, two large-scale DCI test model configurations were analyzed: 1) twin reactor buildings in close proximity and 2) adjacent reactor and turbine buildings. This paper describes the NUPEC DCI test models, the BNL analysis using the SASSI 2000 program, and comparisons between the BNL analysis results and recorded field responses. To account for large variability in the soil properties, the conventional approach of computing seismic responses with the mean, mean plus and minus one-standard deviation soil profiles is adopted in the BNL analysis and the three sets of analysis results were used in the comparisons with the test data. A discussion is also provided in the paper to address 1) the capability of the analysis methods to capture the DCI effect, and 2) the conservatism of the practice for considering soil variability in seismic response analysis for adjacent NPP structures.Copyright

Nonlinear Seismic Correlation Analysis of the JNES/NUPEC Large-Scale Piping System Tests

The Japan Nuclear Energy Safety Organization/Nuclear Power Engineering Corporation (JNES/NUPEC) large-scale piping test program has provided valuable new test data on high level seismic elasto-plastic behavior and failure modes for typical nuclear power plant piping systems. The component and piping system tests demonstrated the strain ratcheting behavior that is expected to occur when a pressurized pipe is subjected to cyclic seismic loading. Under a collaboration agreement between the US and Japan on seismic issues, the US Nuclear Regulatory Commission (NRC)/Brookhaven National Laboratory (BNL) performed a correlation analysis of the large-scale piping system tests using derailed state-of-the-art nonlinear finite element models. Techniques are introduced to develop material models that can closely match the test data. The shaking table motions are examined. The analytical results are assessed in terms of the overall system responses and the strain ratcheting behavior at an elbow. The paper concludes with the insights about the accuracy of the analytical methods for use in performance assessments of highly nonlinear piping systems under large seismic motions.

PROBABILISTIC SITE IDENTIFICATION ANALYSIS USING NUPEC RECORDED FREE FIELD MOTIONS.

THIS PAPER DESCRIBES A PROBABILISTIC SITE IDENTIFICATION ANALYSIS PERFORMED BY BNL, USING THE FREE FIELD EARTHQUAKE MOTIONS RECORDED AT THE NUPEC TEST SITE. THE BNL ANALYSIS WAS INTENDED TO PROVIDE ADEQUATE CHARACTERIZATION OF THE SOIL PROPERTIES FOR THE TEST SITE TO BE USED FOR SSI ANALYSES. THE FREE FIELD DATA WERE PROVIDED BY NUPEC. THE METHODOLOGY EMPLOYED IN THE BNL PROBABILISTIC ANALYSIS OF SITE IDENTIFICATION INCLUDES THE MONTE CARLO PR...

IDENTIFICATION AND ASSESSMENT OF AGING-RELATED DEGRADATION OCCURRENCES IN NUCLEAR POWER PLANTS

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

Evaluation of Simplified Methods for Estimating Shear Capacity Using JNES/NUPEC Low-Rise Concrete Shear Wall Cyclic Test Data.

The simplified methods in current codes for determining the shear capacity of reinforced concrete shear walls had mostly been validated using the test results of single-element shear walls. Recently available JNES/NUPEC test data of reinforced concrete shear walls under multi-directional cyclic loadings provided a unique opportunity to investigate the adequacy of the simplified methods for use in situations with strong interaction effects. A total of 11 test specimens with aspect ratios between 0.47 and 0.87 have been used in the assessment. Two simplified methods from the ACI 349-01 standard [1] and one from the ASCE 43-05 standard [2] have been evaluated. This paper also presents the development of an adjustment factor to consider the aspect ratio and the development of two approaches to consider interaction effects for one of the simplified methods. It concludes with the insights on the applicability of the code methods when interaction effects exist.

Seismic Fragility Analysis of a Condensate Storage Tank With Multiple Uncertain Degradation Scenarios

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are continuing a collaborative effort to achieve a better understanding of the effects of aging on the performance of structures and passive components (SPCs) in nuclear power plants (NPPs). This paper presents a seismic fragility analysis of a condensate storage tank (CST) with multiple degradation scenarios that are treated in a non-perfectly correlated manner. The analysis utilizes a set of optimum Latin Hypercube samples to characterize the deterioration behavior of the fragility capacity as a function of age-related degradations. This study is an addition to the previous study summarized in an ICONE19 paper entitled “Seismic Fragility Analysis of a Degraded Condensate Storage Tank” [1], which considered individual degradation scenarios and multiple degradations occurring in a perfectly correlated manner.Copyright

Structural Design Challenges in Design Certification Applications for New Reactors

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early resolution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific design parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design challenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.Copyright

Assessing Equivalent Viscous Damping Using Piping System Test Results

The specification of damping for nuclear piping systems subject to seismic-induced motions has been the subject of many studies and much controversy. Damping estimation based on test data can be influenced by numerous factors, consequently leading to considerable scatter in damping estimates in the literature. At present, nuclear industry recommendations and nuclear regulatory guidance are not consistent on the treatment of damping for analysis of nuclear piping systems. Therefore, there is still a need to develop a more complete and consistent technical basis for specification of appropriate damping values for use in design and analysis. This paper summarizes the results of recent damping studies conducted at Brookhaven National Laboratory.

Review of Recent Aging-Related Degradation Occurrences of Structures and Passive Components in U.S. Nuclear Power Plants

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are collaborating to develop seismic capability evaluation technology for degraded structures and passive components (SPCs) under a multi-year research agreement. To better understand the status and characteristics of degradation of SPCs in nuclear power plants (NPPs), the first step in this multi-year research effort was to identify and evaluate degradation occurrences of SPCs in U.S. NPPs. This was performed by reviewing recent publicly available information sources to identify and evaluate the characteristics of degradation occurrences and then comparing the information to the observations in the past. Ten categories of SPCs that are applicable to Korean NPPs were identified, comprising of anchorage, concrete, containment, exchanger, filter, piping system, reactor pressure vessel, structural steel, tank, and vessel. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.