Using Limit Analysis for Seismic Evaluation of Component Located in Nuclear Power Plants

2002 ◽  
Author(s):  
Petr Zeman
Keyword(s):  
Limit Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Static Load ◽  
Seismic Load ◽  
Plastic Behavior ◽  
Seismic Evaluation ◽  
Perfectly Plastic ◽  
American Society

Using limit analysis for evaluation of the seismic resistance of the components located in NPPs is compared with the standard evaluation method. This comparison is based on the procedure specified in American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III. Subsection NC, version 1992 standard. The limit analysis uses perfectly plastic behavior of the material. The seismic load is restricted when using limit analysis to the pseudo-static load. The possibility of building of more realistic non-linear model including contacts is another advantage of limit analysis. Using limit analysis is the way to move the evaluation method closer to the real collapse load and to reduce conservatism.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Evaluation Method for Seismic Fatigue Damage of Plant Pipeline

2019 ◽  
Vol 141 (2) ◽  
Author(s):  
Fumio Inada ◽  
Michiya Sakai ◽  
Ryo Morita ◽  
Ichiro Tamura ◽  
Shin-ichi Matsuura ◽  
...  
Keyword(s):  
Fatigue Damage ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Response Spectrum ◽  
Damage Mechanism ◽  
Peak Frequency ◽  
Cumulative Damage ◽  
Cumulative Absolute Velocity ◽  
Seismic Index

Although acceleration and cumulative absolute velocity (CAV) are used as seismic indexes, their relationship with the damage mechanism is not yet understood. In this paper, a simplified evaluation method for seismic fatigue damage, which can be used as a seismic index for screening, is derived from the stress amplitude obtained from CAV for one cycle in accordance with the velocity criterion in ASME Operation and Maintenance of Nuclear Power Plants 2012, and the linear cumulative damage due to fatigue can be obtained from the linear cumulative damage rule. To verify the performance of the method, the vibration response of a cantilever pipe is calculated for four earthquake waves, and the cumulative fatigue damage is evaluated using the rain flow method. The result is in good agreement with the value obtained by the method based on the relative response. When the response spectrum obtained by the evaluation method is considered, the value obtained by the evaluation method has a peak at the peak frequency of the ground motion, and the value decreases with increasing natural frequency above the peak frequency. A higher peak frequency of the base leads to a higher value obtained by the evaluation method.

INTERCOMPARISON ON INTERNAL DOSE ASSESSMENT FOR NUCLEAR POWER PLANTS IN KOREA

2019 ◽  
Vol 186 (4) ◽  
pp. 524-529
Author(s):  
Si Young Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Dose Assessment ◽  
Assessment Procedure ◽  
Internal Dose ◽  
Nuclear Facilities ◽  
International Intercomparison ◽  
Selection Of

Abstract The intercomparison test is a quality assurance activity performed for internal dose assessment. In Korea, the intercomparison test on internal dose assessment was carried out for nuclear facilities in May 2018. The test involved four nuclear facilities in Korea, and seven exposure scenarios were applied. These scenarios cover the intake of 131I, a uranium mixture, 60Co and tritium under various conditions. This paper only reviews the participant results of three scenarios pertinent to the operation of nuclear power plants and adopts the statistical evaluation method, used in international intercomparison tests, to determine the significance values of the results. Although no outliers were established in the test, improvements in the internal dose assessment procedure were derived. These included the selection of intake time, selection of lung absorption type according to the chemical form and consideration of the contribution of previous intake.

Contribution to the Design of Concrete Columns Reinforced With Externally Fiber Polymers

2009 ◽  
Author(s):  
Ferran Prats Bella ◽  
Ramo´n Gonza´lez-Drigo ◽  
Adrina Bachiller San˜a
Keyword(s):  
Structural Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Pushover Analysis ◽  
Concrete Columns ◽  
Seismic Evaluation ◽  
Modal Pushover Analysis ◽  
Nuclear Domain ◽  
Concrete Building ◽  
The Subject

In the design of seismic category 1 buildings in nuclear power plants (NPP) or, outside the nuclear domain, in the conventional structural design of buildings, the seismic evaluation of these buildings may be done. In the occurrence of an earthquake in a NPP or in the case of changing the use of a conventional building, the seismic levels are modified. Then a new analysis need to be performed. This paper focusses on the situation where reinforcing the concrete building is needed and it also analyses how an extern reinforcement performed using polymers can be carried out to fulfill the new seismic requeriments. We present two main results: a) the resulting momentum-curvature diagrams obtained reinforcing standard segments embraced with polymers; b) the evaluation of the structure capacity on the basis of the modified diagrams. Finally, a modal pushover analysis is selected to perform the seismic evaluation of two types of concrete columns, those having a polymer reinforcement and those without it. This paper presents the basis of the subject in a theoretical form.

Security Risk Analysis of Nuclear Power Plants Based on Fuzzy Comprehensive Evaluation

2011 ◽  
Vol 130-134 ◽  
pp. 3708-3711
Author(s):  
Chuan Sheng Xie ◽  
Sheng Ping Hua ◽  
Da Peng Dong ◽  
Xiao Xi Jia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Comprehensive Evaluation ◽  
Risk Index ◽  
Fuzzy Comprehensive Evaluation ◽  
Security Risk ◽  
Fuzzy Comprehensive Evaluation Method ◽  
Comprehensive Evaluation Method

A fuzzy comprehensive evaluation method for nuclear power plants is introduced in this article. First, a risk index system is established of which these indicators will be explained accordingly latter. Then, an evaluation set is constructed, and the weight of each index and corresponding membership is determined according to suggestions of experts and membership function to make evaluation level by level ，until a final comprehensive evaluation is obtained. This method is not very objective but simple and available.

Plastic Limit Loads for Through-Wall Cracked Pipes Using Finite Element Limit Analysis

2006 ◽  
Vol 321-323 ◽  
pp. 724-728
Author(s):  
Nam Su Huh ◽  
Yoon Suk Chang ◽  
Young Jin Kim
Keyword(s):  
Finite Element ◽  
Limit Analysis ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Limit Load ◽  
Plastic Behavior ◽  
Plastic Limit ◽  
Integrity Assessment ◽  
Perfectly Plastic ◽  
Elastic Perfectly Plastic

The present paper provides plastic limit load solutions for axial and circumferential through-wall cracked pipes based on detailed three-dimensional (3-D) finite element (FE) limit analysis using elastic-perfectly plastic behavior. As a loading condition, both single and combined loadings are considered. Being based on detailed 3-D FE limit analysis, the present solutions are believed to be valuable information for structural integrity assessment of cracked pipes.

Crack Growth Evaluation for Cracked Carbon and Stainless Steel Pipes Under Large Seismic Cyclic Loading

2015 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Kunio Onizawa
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Cyclic Response ◽  
Term Operation ◽  
Steel Pipes ◽  
Wave Forms ◽  
Growth Evaluation ◽  
Large Earthquakes

Japanese nuclear power plants have recently experienced several large earthquakes beyond the previous design basis ground motion. In addition, cracks resulting from long-term operation have been detected in piping lines. Therefore, it is very important to establish a crack growth evaluation method for cracked pipes that are subjected to large seismic cyclic response loading. In our previous study, we proposed an evaluation method for crack growth during large earthquakes through experimental study using small specimens. In the present study, crack growth tests were conducted on pipes with a circumferential through-wall crack, considering large seismic cyclic response loading with complex wave forms. The predicted crack growth values are in good agreement with the experimental results for both stainless and carbon steel pipe specimens and the applicability of the proposed method was confirmed.

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