Seismic Analysis and Design Upgrade of Condensate Storage Tanks

2016 ◽  
Author(s):  
Daniel Moreno ◽  
John O’Sullivan ◽  
Tsiming Tseng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Analysis ◽  
Seismic Loads ◽  
Storage Tanks ◽  
Analysis And Design ◽  
Pull Out ◽  
Design Basis ◽  
Regulatory Commission

Following the accident at the Fukushima Daiichi Nuclear Power Plant resulting from the March 11, 2011 Great Tohoku Earthquake and subsequent tsunami, there was a general concern regarding the beyond design basis capability of existing nuclear power plants. The Condensate Storage Tanks (CST) in the nuclear power plant were originally designed to withstand an earthquake with a peak ground acceleration (PGA) of 0.3g. The government regulatory commission increased the required PGA to 0.4g, therefore, an upgrade of the design basis for the CSTs was required. Due to the vintage of the existing nuclear power plant, the United States Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46 methodology may be used for seismic upgrade work of the mechanical and electrical equipment. The Condensate Storage Tanks (CST) belong to the mechanical and electric equipment and therefore were required to have seismic upgrade work because of significant deficiencies that were found in the anchorage of the tanks to the concrete foundation. A seismic analysis and design upgrade of the Condensate Storage Tanks (CST) was performed to resolve exceedances for base shear, overturning moment and sloshing due to recently updated seismic loads. A detailed analysis of the CST showed that the as-built anchor chair detail did not provide sufficient margin for the Beyond Design Basis Earthquake Event (BDBEE). The as-built anchor chair detail did not provide the required strength to transfer the shear and pull-out loads from the tank shell to the concrete foundation, i.e. no clear load path was provided for the updated seismic loads. Therefore, as a result of the inadequate anchorage, the main issues to resolve in the CST were tank sliding, shell buckling and sloshing due to earthquake loading. The principal challenges encountered during the analysis, design and construction stages were (1) not allowing to loosen the double nut configuration attaching the anchor bolt to the existing anchor chair, and therefore allowing to (2) remove or replace only a few components of the as-built anchor chair, and (3) the retrofit design had to be implemented while tanks were operable, i.e. filled with fluid. An additional challenge faced during the design of the new anchor chair components was the limit in anchor chair height imposed by the numerous interferences in the CST such as nozzles, reinforcing plates and existing welds. A mitigation strategy is analyzed, designed and successfully implemented for retrofitting the ninety-six anchor chairs and allowing for full development of the anchor bolt shear and pull-out strength.

scholarly journals Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

2010 ◽  
Author(s):  
Joseph Braverman ◽  
Richard Morante ◽  
Charles Hofmayer ◽  
Robert Roche-Rivera ◽  
Jose Pires
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Gas Generation ◽  
Design Basis ◽  
And Performance ◽  
Regulatory Commission ◽  
Technical Basis

Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 [1] and US NRC Standard Review Plan, Section 3.8) [2]; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 [3] and 10 CFR 50 [1]); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 [3] as well as SECY 90–016 [4], SECY 93–087 [5], and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

Seismic Analysis of an Intake Tunnel of Different Tunnellining Programs for a Nuclear Power Plant

2013 ◽  
Vol 353-356 ◽  
pp. 2084-2091
Author(s):  
Gui Zhen Wang ◽  
Ying Min Li ◽  
Zhi Peng Lu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Analysis ◽  
Time History ◽  
Internal Force ◽  
Response Analysis ◽  
Analysis And Design ◽  
Static And Dynamic Analysis ◽  
History Analysis

Developing an analysis program and simulating the radiation damping effect in the infinite foundation and inputs ground motion rationally. The static and dynamic analysis of an intake tunnel for a nuclear power plant is performed with ANSYS based on some 2-D finite element models considering of structure-foundation interaction by Time-History Analysis. Aiming at different programs of two linings in tunnel, the stress and internal force distribution of the tunnel , the wall rocks and the lining is researched and analyzed, and the diagram of stress and internal force is drawn. The regularity obtained can provide the basis for the seismic response analysis and design of tunnels.

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

Study for Application of Safety Soft Control and Information System in Highly-Integrated Control Rooms of CPR1000 NPP

2013 ◽  
Author(s):  
Xuegang Zhang ◽  
Wei Liu ◽  
Hai Chang ◽  
Jianbo Wen ◽  
Yiqian Wu ◽  
...  
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Integrated Control ◽  
Continuous Control ◽  
Design Basis ◽  
Control Rooms ◽  
Soft Control

For most of the newly built nuclear power plants, the computerized main control rooms (MCR) are adopted. The soft control, the typical feature of computerized Human-Interface System (HIS) in the computerized main control room and mediated by software rather than by direct physical connections, is comprised of safety and non-safety control interface which provides the operators with manual control for component-level, and allows both continuous control of plant process and discrete control of components in nuclear power plant. The safety soft control and information system (SSCIS) is used to give the safety commands to and check the immediate response of the safety process. This paper describes the application of the system design basis, functionality, communication, operation faceplate and system modes for SSCIS which is firstly introduced in CPR1000 nuclear power plant. The design criteria and basic design features of SSCIS is developed to be as the design basis of the design implementation. The ISG-04 ‘Highly-Integrated Control Rooms-Communications issues (HICRc)’ provides acceptable methods for addressing SSCIS communications in digital I&C system design. The NUREG0700 ‘Human-System Interface Design Review Guidelines’ is applied as reference for human factor engineering requirement in the SSCIS design. And the SSCIS design has also fully considered the possible customer usual practice.

scholarly journals Seismic Vulnerability Assessment of Site-Vicinity Infrastructure for Supporting the Accident Management of a Nuclear Power Plant

2017 ◽  
Vol 2017 ◽  
pp. 1-7 ◽  
Author(s):  
T. J. Katona ◽  
A. Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Disaster Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Public Utility ◽  
Qualitative Assessment ◽  
Design Basis ◽  
Accident Management ◽  
Plant Site

Nuclear power plants shall be designed to resist the effects of large earthquakes. The design basis earthquake affects large area around the plant site and can cause serious consequences that will affect the logistical support of the emergency actions at the plant, influence the psychological condition of the plant personnel, and determine the workload of the country’s disaster management personnel. In this paper the main qualitative findings of a study are presented that have been performed for the case of a hypothetical 10−4/a probability design basis earthquake for the Paks Nuclear Power Plant, Hungary. The study covers the qualitative assessment of the postearthquake conditions at the settlements around the plant site including quantitative evaluation of the condition of dwellings. The main goal of the recent phase of the study was to identify public utility vulnerabilities that define the outside support conditions of the nuclear power plant accident management. The results of the study can be used for the planning of logistical support of the plant accident management staff. The study also contributes to better understanding of the working conditions of the disaster management services in the region around the nuclear power plant.

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