Earthquake Resistant Design Code and Validation of Local Metal Loss Procedure Based on the Experimental Data Collected in Japan

2008 ◽  
Author(s):  
Atsushi Ohno ◽  
Takayasu Tahara
Keyword(s):  
High Pressure ◽  
Nuclear Power ◽  
Safety Margin ◽  
The United States ◽  
Metal Loss ◽  
Assessment Procedure ◽  
Loss Assessment ◽  
Power Stations ◽  
Earthquake Resistant Design ◽  
Earthquake Load

Fitness-For-Service (FFS) assessments are performed to evaluate the components damaged in service to determine whether it is possible to continue their use. FFS assessment codes were recently standardized, and they are being used in many companies in Europe and the United States. In Japan, the regulation permits the use of FFS codes in nuclear power stations, but not yet in petroleum and petrochemical industries. The PAJ/JPCA FFS task group that consists of the members of petroleum or petrochemical companies has been studying and investigating one of the FFS Codes, API579-1/ASME FFS-1 [1], in an attempt to include it in the high pressure gas safety law [2], which regulates the pressure equipment operating at pressures greater than 1 MPa. We have now completed the adaptation of the FFS code for Japan, and it is in the process of being assessed by the authorities. It is required that the code is modified slightly because Japanese authorities and people are particularly nervous to matters regarding earthquake safety. This paper focuses on cylindrical equipment regulated by the high-pressure gas safety law. The margin for earthquake load of the actual equipment is shown, and the local metal loss assessment procedure according to API579-1/ASME FFS-1 is verified by using experimental burst test data with pressure and/or bending stress in order to determine whether or not the FFS code provides a sufficient safety margin for safe operations in Japan.

Development of Fitness-for-Service Standard for Pressure Equipment With Metal Loss Based on Reliability

2015 ◽  
Author(s):  
Takuyo Kaida ◽  
Shinsuke Sakai
Keyword(s):  
High Pressure ◽  
Structural Integrity ◽  
Limit State ◽  
Probabilistic Approach ◽  
Metal Loss ◽  
Assessment Procedure ◽  
Rational Decision ◽  
Loss Assessment

Concern about probabilistic approach for Fitness-For-Service (FFS) assessment has been growing over the last several years. The FFS assessment based on reliability helps to make a rational decision as to whether to run or repair the equipment. High Pressure Institute of Japan (HPI) formed a committee to develop a HPI FFS standard that can be used for pressure equipment with metal loss. This new standard provides an assessment procedure to evaluate structural integrity of components with metal loss based on reliability. This paper introduces the assessment procedure which is standardized and under preparation for publication, and the technical backgrounds. The standard provides information about limit state of pressure equipment, probabilistic properties of basic variables and target reliability. Probabilistic approach can be applied easily to metal loss assessment by using the standard.

Japanese Code for Assessment Procedure for Crack-like Flaws in Pressure Equipment

2002 ◽  
Author(s):  
Shinji Konosu ◽  
Takayasu Tahara ◽  
Hideo Kobayashi
Keyword(s):  
High Pressure ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Plate Thickness ◽  
Pressure Vessels ◽  
Material Strength ◽  
Assessment Procedure ◽  
Absorbed Energy ◽  
Industrial Facilities ◽  
Characterization Procedure

There are numerous instances in which in-service flaws due to various kinds of damage and deterioration are found in equipment as a result of in-service inspections. The proper evaluation of such flaws is extremely important. Fitness-for-Service (FFS) codes, such as ASME B&PV Code Sec. XI and JSME S NA1 for nuclear power generation facilities and BS 7910 and API-RP579 for general industrial facilities, are available. In light of such circumstances, the High Pressure Institute of Japan (HPI) has prescribed its code “Assessment procedure for crack-like flaws in pressure equipment” for conducting quantitative safety evaluations of flaws detected in common industrial pressure components such as pressure vessels, piping, storage tanks, and so on designed and fabricated in accordance with Japanese codes and regulations such as JIS B8265 and High Pressure Gas Safety Law. The FFS code consists of Level 1 assessment (whereby assessment can be conducted without extensive knowledge of fracture mechanics) and Level 2 assessment (which enables more detailed fracture mechanics analyses and is currently being studied). The allowable flaw size is specified in accordance with the plate thickness. The required impact absorbed energies based on material strength, whether or not PWHT has been done and the orientation of the flaw in relation to the weld seam, are also specified. An approximated equation of stress intensity factor for an embedded flaw near the surface has been derived. The re-characterization procedure for assessing an embedded flaw has been clarified. The flaw can be judged to be acceptable if its size is less than that of an allowable flaw and the equipment is to be used at temperatures exceeding the temperature (MAT) at which the material absorbed energy meets the required impact absorbed energy.

Local planning v. national policy: urban growth near nuclear power stations in the United States

1986 ◽  
Vol 57 (3) ◽  
pp. 225 ◽  
Author(s):  
Michael R. Greenberg ◽  
Donald A. Krueckeberg ◽  
Michael Kaltman ◽  
William Metz ◽  
Charles Wilhelm
Keyword(s):  
United States ◽  
Urban Growth ◽  
Nuclear Power ◽  
National Policy ◽  
The United States ◽  
Local Planning ◽  
Power Stations

scholarly journals Material Matters

MRS Bulletin ◽  
1998 ◽  
Vol 23 (3) ◽  
pp. 6-16 ◽  
Author(s):  
W. Stoll
Keyword(s):  
Soviet Union ◽  
Former Soviet Union ◽  
Nuclear Power ◽  
The United States ◽  
Weapons Of Mass Destruction ◽  
Spent Fuel ◽  
Power Stations ◽  
The Soviet Union ◽  
Water Reactors ◽  
Enriched Uranium

The following article is based on a talk for Symposium X presented by Wolfgang Stoll, Chief Scientific Advisor and Consultant in Siemens, Germany, at the 1996 MRS Fall Meeting.Since 1941 when Glenn Seaborg first isolated plutonium in milligram quantities, the total amount converted through neutron capture in U-238 has increased worldwide to about 1,200 tons and continues to grow about 70 tons/year. What was fissioned in situ in operating nuclear power stations is roughly equivalent to 5 billion tons of black coal, while the fission energy contained in those 1,200 tons unloaded in spent fuel is equivalent to another 2 billion tons of coal. About 260 of these 1,200 tons are ready to release their energy in about 4 kg-portions each in microseconds which is equivalent to 10,000 tons of coal. Most people believe this release of energy poses a major threat of the worldwide arsenal of weapons of mass destruction (WMD). The about 20-fold overkill stored in worldwide WMD is considered superfluous after the crumbling of the Soviet Union. Options are sought to dispose of this surplus in a safe, speedy, and controllable manner. While for highly enriched uranium (HEU) (the other nuclear weapons material) dilution into low-enriched uranium and utilization in current light water reactors (LWR) poses market adaptation problems only, and while the worldwide consensus on the elimination of chemical and biological WMD is still in an initial phase, the decision of both the United States (US) and the former Soviet Union (FSU) to remove most of the plutonium out of weapons looks as if it was a firm political decision.

Safe Operation and FFS Assessment for Pressure Equipment: Updated With Study for Local Metal Loss Assessment

2006 ◽  
Author(s):  
Atsushi Ohno ◽  
Yoshiaki Uno ◽  
Takayasu Tahara
Keyword(s):  
United States ◽  
Finite Element Analysis ◽  
Finite Element ◽  
Metal Loss ◽  
Assessment Procedure ◽  
Loss Assessment ◽  
Safe Operation ◽  
Element Analysis ◽  
Assessment Procedures ◽  
Design Margin

Recently, Codes and Standards for FFS assessment has been developed and applied in United States and other countries such as API RP579 as a series of maintenance procedures for pressure equipment. Activities developing FFS assessment procedures in conjunction with new safe inspection standards are also progressing in Japan. In order to prove applicability of the FFS procedure for assessment of damaged pressure equipment, it is also important to validate how much of inservice safe margin is derived from the FFS assessment procedures in compared with design margin of pressure equipment. Local metal loss assessment procedure specified by API RP579 is studied using finite element analysis and discussed how much of in-service safe margin will be sufficient as standardized FFS assessment procedure.

LBB Analysis on Nuclear-Class Pipes of Floating Nuclear Power Stations

2018 ◽  
Author(s):  
Fan Bai ◽  
Yong Liu ◽  
Xingsheng Lao ◽  
Qi Xiao ◽  
Zhenxing Zhao ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Oil And Gas ◽  
Critical Length ◽  
Unstable Fracture ◽  
Assessment Procedure ◽  
Gas Drilling ◽  
Power Stations ◽  
Offshore Oil And Gas ◽  
Pressure Swing

Floating nuclear power stations are vessels with nuclear reactors, designed to power offshore oil and gas drilling, island development and remote areas. The safety of the facilities is an important issue. The Leak-Before-Break (LBB) assessment is essential to the design and evaluation of nuclear power plants against a sudden double-ended guillotine break of pipes. This paper describes the LBB assessment procedure applicable to the nuclear-class pipe of the floating nuclear power station. The loads considered in the analysis include variation of temperature/pressure, swing and underwater impact. Circumferential cracks are postulated at the dangerous positions of the pipe. The fatigue propagation of the surface crack is predicted based on the fracture mechanics and the finite element method to determine the time that pipe leaking happens. The critical length of the through-wall crack is calculated by the J integral–tearing modulus approach, and then compared with the minimum detectable crack length for the purpose that unstable fracture would not be happened before the leakage detected. According to the results of the analysis it could conclude that the pipe fulfils the LBB requirements.

scholarly journals Risk-Informed Safety Margin Characterization

2009 ◽  
Author(s):  
Stephen M. Hess ◽  
Nam Dinh ◽  
John P. Gaertner ◽  
Ronaldo Szilard
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Safety Margin ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Plant Design ◽  
Time Horizons ◽  
Safety Margins

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.

Fitness-For-Service Assessment of Pressure Equipments With Local Metal Loss Subjected to Seismic Loading

2013 ◽  
Author(s):  
Takashi Yamamoto ◽  
Takuyo Kaida ◽  
Satoshi Nagata ◽  
Hirokazu Tsuji
Keyword(s):  
Internal Pressure ◽  
Ambient Temperature ◽  
Safety Margin ◽  
Seismic Loading ◽  
Metal Loss ◽  
Seismic Load ◽  
Assessment Procedure ◽  
Safety Margins ◽  
Seismic Design Code ◽  
Bending Load

The concerns about Fitness-For-Service (FFS) assessment technique for pressure equipments with local metal loss have been growing from some characteristic damages, for example, many examples of the corrosion under insulation (CUI) of pressure equipment, have been reported from petroleum and petrochemical industries. FFS assessment procedure for the pressure equipments for metal loss has been validated by the results of various burst tests and FEM simulations for internal pressure loads. There has, however, been little study to validate FFS assessment for pressure equipments subjected to seismic load. This paper suggests an FFS assessment procedure for pressure equipments with local metal loss subjected to both internal pressure and seismic loading. To ensure consistency to High Pressure Gas Law in Japan, allowable stress is based on the Japanese seismic design code. Developed stress on local metal loss from both internal pressure and seismic load is evaluated in accordance with API 579/ASME FFS-1. The authors have verified safety margin and reliability of this method to toward to practical application. In order to verify, some cyclic bending load testings and finite element analysis were implemented under the conditions of ambient temperature and 300 degree C. The results of these validations show that the safety margins against low cycle fatigue are the range of 2.6 to 4.6. In addition, the test results at 300 degree C showed higher safety margin than that in ambient temperature.

Core Protection Calculator System: Past, Present, and Future

2010 ◽  
Author(s):  
Peter L. Hung
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Plant Protection ◽  
Power Level ◽  
Safety Margin ◽  
Operating Experience ◽  
Nucleate Boiling ◽  
The United States ◽  
Reactor Power ◽  
Safe Condition

The Core Protection Calculator System (CPCS) was the first implementation of digital computers in a nuclear power plant safety protection system. The system was based on first principles to calculate the specified acceptable fuel design limit (SAFDL) online. This approach provides the theoretical optimum safety margin. The first-of-its-kind system was installed in the United States at Arkansas Nuclear One Unit 2 (ANO-2) in 1980. Extensive efforts were made by Combustion Engineering and U.S. Nuclear Regulatory Commission (NRC) staff to gain licensing approval of the CPCS. Based on accumulated operating experience, numerous improvements were made to enhance the performance of the CPCS. The CPCS software provided the flexibility to readily accommodate these design changes. Currently, CPCS is implemented in 21 nuclear power plants in operation or under construction in the U.S.A. and Asia. The next generation CPCS will focus on optimizing the plant protection by improving the SAFDL calculation. By taking advantage of the advances in digital computer technology, the comprehensive safety analysis code will be used online. A more detailed core power map using the incore detector signals will be used as the basis of the departure from nucleate boiling ratio (DNBR) and local power density (LPD) calculation. A quick power reduction will provide adequate margin for most of the design basis events. For these events, CPCS will initiate a reactor power cutback as opposed to a reactor trip, which will maintain the plant at a safe condition with a reduced power level.

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