Jinsuo Nie

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

On the Correct Application of the 100-40-40 Rule for Combining Responses due to Three Directions of Earthquake Loading

The 100-40-40 rule is often used with the response spectrum analysis method to determine the maximum seismic responses from structural responses resulting from the three spatial earthquake components. This rule has been referenced in several recent Design Certification applications of nuclear power plants, and appears to be gaining in popularity. However, this rule is described differently in ASCE 4-98 and Regulatory Guide 1.92, consequently causing confusion on correct implementation of this rule in practice. The square root of the sum of the squares method is another acceptable spatial combination method and was used to justify the adequacy of the 100-40-40 rule during the development of the Regulatory Guide 1.92. The 100-40-40 rule, when applied correctly, is almost always conservative compared to the SRSS method, and is only slightly unconservative in rare cases. The purpose of this paper is to describe in detail the proper application of the 100-40-40 rule, as prescribed in ASCE 4-98 and in Regulatory Guide 1.92, and to clarify the confusion caused by the two different formats of this rule.

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.

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.

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.

An Approach for Assessing Structural Uplifting Using Blast Motions.

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.