Development of Russian Regulatory Basis for License Renewal/Life Extension of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Pachner ◽  
Inna V. Kaliberda
Keyword(s):  
Power Plant ◽  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Extension ◽  
Safety Standard ◽  
Regulatory Strategy ◽  
Regulatory Guidelines ◽  
License Renewal ◽  
Regulatory Organization

By 2010, fifteen Russian nuclear power plant (NPP) units will reach the end of their scheduled service life of 30 years. An extension of NPP operation beyond the scheduled service life, which is provided for by the Russian safety standard OPB-88/97, requires a renewal of the plant operating license by the regulatory organization Gosatomnadzor (GAN). This paper presents an overview of an IAEA project aimed at assisting GAN in the development of a set of regulatory guidelines for NPP license renewal/life extension. The overview includes a description of the Russian regulatory strategy for life extension of NPPs and IAEA activities to facilitate drafting and application of the regulatory guidelines using international experience.

Hot-Gas-Path Life Extension Options for the V94.2 Gas Turbine

2000 ◽  
Author(s):  
Gerhard Bohrenkämper ◽  
Herbert Bals ◽  
Ursel Wrede ◽  
René Umlauft
Keyword(s):  
Power Plant ◽  
Service Life ◽  
Gas Turbine ◽  
Power Plants ◽  
Combined Cycle ◽  
Life Extension ◽  
Hot Gas ◽  
Maintenance Program ◽  
Path Component ◽  
Strategic Options

Gas turbine and combined cycle power plants are typically designed for a service life of over 30 years. If operated at base load in continuous duty, the gas turbine hot-gas-path components for example in a combined-cycle power plant need repair and replacement according to the maintenance program several times during plant life. Most of the hot components would reach the end of their service life, e.g. 100,000 equivalent operating hours (EOH), after 10 to 12 years. As this is well before the end of the overall plant service life defined in the power plant concept, such plant applications therefore necessitate life extension measures enabling to continuing operation beyond 100,000 EOH. This paper presents strategic options for hot-gas-path component life entension.

Assessment of Remaining Life for Class 1 Piping Components Experiencing Wall Thinning

2004 ◽  
Author(s):  
Rajnish Kumar
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Extension ◽  
Remaining Life ◽  
Original Design ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Class 1 ◽  
Plant Components

Assessment of remaining life of power plant components is important in light of plant life management and life extension studies. This information helps in planning and minimizing plant outages for repairs and refurbishments. Such studies are specifically important for nuclear power plants. Nuclear Safety Solutions Limited (NSS) is involved in conducting such studies for plant operators and utilities. Thickness measurements of certain piping components carrying fluids at high temperature and high pressure have indicated higher than anticipated wall thinning rates. Flow accelerated corrosion (FAC) has been identified as the primary mechanism for this degradation. The effect of FAC was generally not accounted for in the original design of the plants. Carbon steel piping components such as elbows, tees and reducers are prone to FAC. In such cases, it is important to establish the remaining life of the components and assess their adequacy for continued service. Section XI of the ASME Boiler and Pressure Vessel Code is applicable for evaluation of nuclear power plant components in service. This Section of the Code does not specifically deal with wall thinning of the piping components. Code Case N-597 provides guidelines for evaluation for continued service for Class 2 and Class 3 piping components. For Class 1 piping components, this Code Case suggests that the plant owner should develop the methodology and criteria for evaluation. This paper presents methodology and procedure for establishing the remaining life and assessment of Class 1 piping components experiencing wall thinning effects. In this paper, the rules of NB-3600 and NB-3220 and Code Case N-597 have been utilized for assessment of the components for continued service. Details of various considerations, criteria and methodology for assessment of the remaining life and adequacy for continued service are provided.

Fatigue Monitoring: Case Studies in Nuclear Power Plant

2018 ◽  
Author(s):  
G. Bourguigne ◽  
F. Schroeter
Keyword(s):  
Power Plant ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Life Extension ◽  
Monitoring Systems ◽  
Fatigue Monitoring ◽  
Cycle Count ◽  
Critical Components

During design of Class I components in Nuclear Power Plants, cumulative usage factors (CUF) are conservatively calculated to estimate fatigue damage, and results must be below the limits of the applicable codes. Nevertheless, when these results are used to evaluate the possibility of using these components for an extended life, the results are frequently above code limits. Many Nuclear Power Plants have installed commercial fatigue monitoring systems at critical components in order to assess transient severity and cycle count for life extension fatigue calculations among other reasons. Since the commissioning of the system, unexpected operation modes and thermal stratification was discovered and evaluations needed to be done. Findings, interpretations and solving are presented in this paper.

Environmentally-Assisted Fatigue Management of PWR Pressurizer Using Nonlinear Strain-Based Analysis

2013 ◽  
Author(s):  
Timothy Gilman ◽  
Jay Gillis ◽  
Jagannath Hiremagalur ◽  
Scott Rodamaker ◽  
William Weitze ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nonlinear Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nonlinear Strain ◽  
Fatigue Management ◽  
Fatigue Evaluation ◽  
License Renewal

This paper describes the techniques utilized to perform a nonlinear, strain-based, environmentally-assisted fatigue evaluation of a pressurizer vessel in a nuclear power plant. Significant differences between the strain-based and more traditional fatigue analysis results are demonstrated. This paper concludes that leveraging today’s computer power with the use of more detailed, nonlinear analysis is an effective tool for nuclear power plants to meet license renewal commitments related to the management of environmentally-assisted fatigue.

Study on the Interaction Between Safety-Related and Non Safety-Related Items in the Component Cooling Water System Room of the Qinshan Nuclear Power Plant in the Earthquake Condition

2020 ◽  
Author(s):  
Liang Zhang ◽  
Gang Xu ◽  
Yue Wang ◽  
Li Chen ◽  
Shao Chong Zhou
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Water System ◽  
Cooling Water ◽  
Life Extension ◽  
Main Body ◽  
Cooling Water System ◽  
The Impact

Abstract Safety-related items in nuclear power plants are now generally placed separately from the non-safety-related items, but it was not strictly required before. Therefore, it is very important to study whether the non-safety-related items will affect the safety-related items when they are dropped down in an earthquake situation, which determines the safety of a nuclear power plant and its future life extension applications. This research was based on the cooling water system room with the safety and non-safety related items installed together, as an example to study whether the non-safety-related items such as vent pipes and DN50 fire fighting pipes arranged above will damage the DN300 pipes and valves arranged below when earthquakes occur. For the experiments, the relative positions of objects in the room was reproduced by 1: 1. The pressure-holding performance of the pipe was used as a criterion for the damage. The research results of the experiments show that when the 10-meter-long DN50 pipe was dropped from the position of 8 meters height and the 8-meter-long vent dropped from position of 3.6 meters height, they do not affect the integrity of the DN300 valve and pipe below. After the experiment, pressure drop in two hours for the pipe is less than 0.1%. The main body of the valve does not fail neither. The numerical simulation study also shows that there is no failure phenomenon in the simulation as well. Compared with the test results, the impact acceleration and the vent deformation both have the same trend.

scholarly journals ISTD Implementation and Service-Life Monitoring

2014 ◽  
Author(s):  
Glen Palmer
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Federal Regulations ◽  
Testing Program ◽  
Operation And Maintenance ◽  
The Difference ◽  
Program Change

Subsection ISTD of ASME’s Operation and Maintenance of Nuclear Power Plants (OM Code) is the required code for preservice and inservice examination and testing of dynamic restraints (snubbers). This code replaced the inspection requirements of Article IWF-5000, “Inservice Inspection Requirements for Snubbers,” in Section XI, “Inservice Inspection of Nuclear Power Plant Components,” of the ASME Boiler and Pressure Vessel Code after the publication of the 2006 addenda to Section XI, which deleted Article IWF-5000. When the requirements of IWF-5000 were deleted, the requirements for examination and testing of snubbers, as required by Section 50.55a, “Codes and Standards,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR 50.55a) became those specified by Subsection ISTD of the ASME OM Code. Therefore, when nuclear power plant owners prepare their ten-year inservice testing (IST)/inservice inspection (ISI) program updates that incorporate the 2006 (or later) addenda to Section XI, the snubber requirements will be required to be in accordance with those of Subsection ISTD of the latest approved edition and addenda of the ASME OM Code (2004 Edition with Addenda through 2006). This edition of the ASME OM Code is cited in the NRC Rulemaking which was published on June 21, 2011. Because this is a change in requirements, owners should be asking some of the following questions: What is the difference between our existing program requirements and those included in Subsection ISTD of the ASME OM Code? How will this change our existing program or the way the current snubber examination and testing program is implemented? How much effort will be required to implement this program change? This paper will provide some specific guidance for the implementation of the ISTD Code and will identify typical areas where changes may be required to existing snubber examination and testing programs. It will also describe some approaches to satisfy the requirements of ISTD-6000, “Service Life Monitoring,” which might not have been included in the previous requirements under Section XI. Paper published with permission.

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