Mechanical Collateral Damage Assessment of Reactor Vessel Bottom Mounted Nozzles: Part I—Requirements and Methodology

2011 ◽  
Author(s):  
Tama´s R. Liszkai
Keyword(s):  
Damage Assessment ◽  
Safety Assessment ◽  
Failure Modes ◽  
Response Surfaces ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Collateral Damage ◽  
Original Design ◽  
Pressurized Water ◽  
Inside Radius

A comprehensive work scope including the engineering safety assessments, Non-Destructive Examination (NDE) and repair design, is developed by AREVA NP Inc. for the Reactor Vessel (RV) Incore Monitoring Instrument (IMI) nozzles. The joint Bottom Mounted Nozzle (BMN) Assessment Plan is coordinated under the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The purpose of such coordination is to produce a safety assessment of consistent scope and methodology to address the different IMI nozzle designs in all U.S. Pressurized Water Reactors (PWRs). The IMI nozzles, which are also referred to as the BMNs are installed in the bottom of the reactor vessel RV. For the Babcock & Wilcox (B&W) designed plants the nozzles consist of the original Alloy 600 nozzle material attached to the reactor vessel by a partial penetration Alloy 182 weld. To increase the resistance of the nozzles against flow induced vibration (FIV), the nozzles were modified, which consisted of a thicker, more rigid Alloy 600 nozzle welded to the RV inside radius surface. Recent industry experience indicates that the Alloy 600 BMNs and their Alloy 82/182 weld metal may be more susceptible to primary water stress corrosion cracking (PWSCC) than previously thought. Although the BMNs have been ranked low in susceptibility to PWSCC, they are ranked as having the most severe consequences of failure. Failure of BMNs represents a scenario that would result in a leak or loss of coolant accident (LOCA). Failure of a BMN was not included in the original design basis for the B&W designed plants. This paper describes the mechanical collateral damage analysis of the BMN engineering safety assessment project performed under the sponsorship of PWR Owner’s Group (PWROG) for the seven operating B&W 177-FA PWR units. Failure of a BMN could potentially lead to pipe whip that could impact other IMI pipes. The goal of the mechanical collateral damage assessment is to determine the potential loads on adjacent IMI pipes. First, the IMI piping configurations for all B&W plants were determined. Based on the piping configurations, potential pipe whip pairs were identified and several representative finite element models of the IMI piping were developed. Using the results of the nonlinear transient dynamic pipe whip analyses, response surfaces were developed, which provided the basis for determining loads due to pipe whip at several different locations. The conservative ultimate capacity analysis corresponding to 50% ultimate strain of the materials showed that the maximum ultimate stress ratio of the intact nozzle cross section at the RV outside radius was acceptable. In addition, the fracture mechanics evaluation of the flawed nozzles, at the RV inside radius, showed that the maximum critical half flaw angle was large enough that early detection of leaking BMNs is possible. For other possible failure modes of the piping, such as the jet impingement, asymmetric cavity pressure effects and insulation frame movement, it was shown that the loads obtained from the pipe whip analyses envelop those loads. The description of this work has been divided into two papers. Part I, detailed in this paper, describes the development of the comprehensive collateral damage assessment methodology. Part II, [1], to be also presented at PVP-2011, presents illustrative examples of the pipe whip analyses and application of response surfaces.

Mechanical Collateral Damage Assessment of Reactor Vessel Bottom Mounted Nozzles: Part II—Analysis and Applications

2011 ◽  
Author(s):  
Tama´s R. Liszkai
Keyword(s):  
Damage Assessment ◽  
Safety Assessment ◽  
Failure Modes ◽  
Response Surfaces ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Collateral Damage ◽  
Original Design ◽  
Pressurized Water ◽  
Inside Radius

A comprehensive work scope including the engineering safety assessments, Non-Destructive Examination (NDE) and repair design, is developed by AREVA NP Inc. for the Reactor Vessel (RV) Incore Monitoring Instrument (IMI) nozzles. The joint Bottom Mounted Nozzle (BMN) Assessment Plan is coordinated under the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The purpose of such coordination is to produce a safety assessment of consistent scope and methodology to address the different IMI nozzle designs in all U.S. Pressurized Water Reactors (PWRs). The IMI nozzles, which are also referred to as the BMNs are installed in the bottom of the reactor vessel RV. For the Babcock & Wilcox (B&W) designed plants the nozzles consist of the original Alloy 600 nozzle material attached to the reactor vessel by a partial penetration Alloy 182 weld. To increase the resistance of the nozzles against flow induced vibration (FIV), the nozzles were modified, which consisted of a thicker, more rigid Alloy 600 nozzle welded to the RV inside radius surface. Recent industry experience indicates that the Alloy 600 BMNs and their Alloy 82/182 weld metal may be more susceptible to primary water stress corrosion cracking (PWSCC) than previously thought. Although the BMNs have been ranked low in susceptibility to PWSCC, they are ranked as having the most severe consequences of failure. Failure of BMNs represents a scenario that would result in a leak or loss of coolant accident (LOCA). Failure of a BMN was not included in the original design basis for the B&W designed plants. This paper describes the mechanical collateral damage analysis of the BMN engineering safety assessment project performed under the sponsorship of PWR Owner’s Group (PWROG) for the seven operating B&W 177-FA PWR units. Failure of a BMN could potentially lead to pipe whip that could impact other IMI pipes. The goal of the mechanical collateral damage assessment is to determine the potential loads on adjacent IMI pipes. First, the IMI piping configurations for all B&W plants were determined. Based on the piping configurations, potential pipe whip pairs were identified and several representative finite element models of the IMI piping were developed. Using the results of the nonlinear transient dynamic pipe whip analyses, response surfaces were developed, which provided the basis for determining loads due to pipe whip at several different locations. The conservative ultimate capacity analysis corresponding to 50% ultimate strain of the materials showed that the maximum ultimate stress ratio of the intact nozzle cross section at the RV outside radius was acceptable. In addition, the fracture mechanics evaluation of the flawed nozzles, at the RV inside radius, showed that the maximum critical half flaw angle was large enough that early detection of leaking BMNs is possible. For other possible failure modes of the piping, such as the jet impingement, asymmetric cavity pressure effects and insulation frame movement, it was shown that the loads obtained from the pipe whip analyses envelop those loads. The description of this work has been divided into two papers. Part II detailed in this paper presents illustrative examples of the pipe whip analyses and application of response surfaces. Part I [1], to be also presented at PVP-2011, describes the development of the comprehensive collateral damage assessment methodology.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

2004 ◽  
Author(s):  
Dilip Bhavnani ◽  
James Annett
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Supply System ◽  
Reactor Vessel ◽  
Original Design ◽  
Pressurized Water ◽  
Class 1 ◽  
Current Design ◽  
Successful Design

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

Evaluation of Laser Peening for Mitigation of Primary Water Stress Corrosion Cracking in Pressurized Water Reactors

2019 ◽  
Author(s):  
Stephen Marlette ◽  
Stan Bovid
Keyword(s):  
Surface Stress ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Laser Peening ◽  
Alloy 600 ◽  
Test Results ◽  
Pressurized Water ◽  
Performance Requirements ◽  
Primary Water ◽  
Water Reactors

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Advanced Nondestructive Examination Technologies for Alloy 600 Components

2004 ◽  
Author(s):  
Donald C. Adamonis ◽  
Robert P. Vestovich ◽  
Fred G. Whytsell ◽  
Filippo D’Annucci ◽  
Eric Loehlein ◽  
...  
Keyword(s):  
The United States ◽  
Management Program ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Nondestructive Examination ◽  
The Status ◽  
Primary Water ◽  
Water Reactors

Several pressurized water reactors have experienced primary coolant leaks as a result of degradation in the tubes and J-groove welds of reactor vessel head penetrations. Leakage has been attributed to primary water stress corrosion cracking (PWSCC) of the Alloy 600 nozzle material and Alloy 182/82 weld materials. More recently, other Alloy 600 components including reactor vessel bottom mounted instrumentation nozzles, dissimilar metal pipe welds, hot leg instrument penetrations, and pressurizer heater sleeves have exhibited degradation. Westinghouse has been actively involved in the development of a comprehensive Alloy 600 degradation management program including investigation of root cause, establishing a safety position, and developing inspection and repair/replacement strategies to address Alloy 600 degradation issues. Personnel from Germany, Sweden and the United States have cooperatively developed equipment and nondestructive examination technologies for identification and characterization of degradation that might exist in these Alloy 600 components and attachment welds. These developments represent significant enhancements to technologies and equipment previously available in the industry and are driven by the need to meet new inspection applications and industry requirements which have continued to evolve over the last several years. This paper describes the inspection capabilities available to support Alloy 600 degradation management programs, field experience with those inspection technologies, and the status of ongoing NDE development efforts to enhance future Alloy 600 inspection programs.

Materials Aging Degradation of Reactor Vessel Internals: Part I — Thermal Hydraulics Evaluation

2008 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Norm S. Yee ◽  
Jim R. Smotrel ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Failure Modes ◽  
Operating Experience ◽  
Material Model ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Essential Information ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), is developing Reactor Internals Inspection and Evaluation (I&E) Guidelines mainly to support U.S. license renewed plants. These guidelines are organized around a framework and strategy, [3] and [4], for managing the effects of aging in PWR internals as shown in Figure 1, dependent on a substantial database of material data and supporting results. The key steps include the following: the development of screening criteria, with susceptibility levels for the eight postulated aging mechanisms relevant to reactor internals and their effects [5]; an initial component screening and categorization step, using the susceptibility levels to identify the relative susceptibility of the components; a functionality assessment of degradation for components and assemblies of components; and finally aging management strategy development combining the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. The purpose of this functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation up to 60 years of operation. The stainless steel material model employed in the calculations is an irradiated material-specific constitutive model for use in a finite element analysis [8]. The material model accounts for the effects of plasticity, irradiation assisted stress corrosion cracking (IASCC), irradiation creep-stress relaxation, void swelling, and embrittlement as a function of temperature and fluence [6] and [7]. The study focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, detailed finite element models were developed capable of representing the complex interactions between the components. The goal of this study is to characterize the potential failure modes, spatial and chronological distribution of component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three major physics fields. Radiation calculations of the core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the RV internals. The three major physics fields and their relations are illustrated in Figure 2. This paper focuses on the CFD/CHT aspects of the overall analysis for the B&W designed RV internals and provides information on the state-of-the-art multi-physics approach employed.

Safety Assessment of Long Term Serviced Pressure Vessels: A Case Study of Typical Refining and Chemical Plants in China

2021 ◽  
Author(s):  
Zhiyuan Han ◽  
Guoshan Xie ◽  
Haiyi Jiang ◽  
Xiaowei Li
Keyword(s):  
Pressure Vessel ◽  
Safety Assessment ◽  
Failure Modes ◽  
Pressure Vessels ◽  
Time Dependent ◽  
Chemical Plants ◽  
Safety Specification ◽  
Detailed Assessment

Abstract The safety and risk of the long term serviced pressure vessels, especially which serviced more than 20 years, has become one of the most concerned issues in refining and chemical industry and government safety supervision in China. According to the Chinese pressure vessel safety specification TSG 21-2016 “Supervision Regulation on Safety Technology for Stationary Pressure Vessel”, if necessary, safety assessment should be performed for the pressure vessel which reaches the design service life or exceeds 20 years without a definite design life. However, the safety and risk conditions of most pressure vessels have little changes after long term serviced because their failure modes are time-independent. Thus the key problem is to identify the devices with the time-dependent failure modes and assess them based on the failure modes. This study provided a case study on 16 typical refining and chemical plants including 1870 pressure vessels serviced more than 20 years. The quantitative risk and damage mechanisms were calculated based on API 581, the time-dependent and time-independent failure modes were identified, and the typical pressure vessels were assessed based on API 579. Taking the high pressure hydrogenation plant as an example, this study gave the detailed assessment results and conclusions. The results and suggestions in this study are essential for the safety supervision and extending life of long term serviced pressure vessels in China.

ADS-B: Probabilistic Safety Assessment

2017 ◽  
Vol 70 (4) ◽  
pp. 887-906 ◽  
Author(s):  
Busyairah Syd Ali ◽  
Washington Yotto Ochieng ◽  
Arnab Majumdar
Keyword(s):  
Safety Assessment ◽  
Failure Modes ◽  
Fault Tree Analysis ◽  
Performance Standards ◽  
System Safety ◽  
Probabilistic Safety Assessment ◽  
Assessment Approach ◽  
Primary Means ◽  
Importance Analysis ◽  
Probabilistic Safety

In the effort to quantify Automatic Dependent Surveillance Broadcast (ADS-B) system safety, the authors have identified potential ADS-B failure modes in Syd Ali et al. (2014). Based on the findings, six potential hazards of ADS-B are identified in this paper. The authors then applied the Probabilistic Safety Assessment approach which includes Fault Tree Analysis (FTA) and Importance Analysis methods to quantify the system safety. FTA is applied to measure ADS-B system availability for each identified hazard while Importance Analysis is conducted to identify the most significant failure modes that may lead to the occurrence of the hazards. In addition, risk significance and safety significance of each failure mode are also identified. The result shows that the availability for the ADS-B system as a sole surveillance means is low at 0·898 in comparison to the availability of ADS-B system as supplemental or as primary means of surveillance at 0·95 and 0·999 respectively. The latter availability values are obtained from Minimum Aviation System Performance Standards (MASPS) for Automatic Dependent Surveillance-Broadcast (DO-242A).

Residual Stress Evaluation of Dissimilar Weld Joint Using Reactor Vessel Outlet Nozzle Mock-Up Model (Report-1)

2008 ◽  
Author(s):  
Itaru Muroya ◽  
Youichi Iwamoto ◽  
Naoki Ogawa ◽  
Kiminobu Hojo ◽  
Kazuo Ogawa
Keyword(s):  
Residual Stress ◽  
Stress Distribution ◽  
Welding Process ◽  
Reactor Vessel ◽  
Residual Stress Distribution ◽  
Test Model ◽  
Alloy 600 ◽  
Outlet Nozzle ◽  
Stress Evaluation ◽  
Weld Region

In recent years, the occurrence of primary water stress corrosion cracking (PWSCC) in Alloy 600 weld regions of PWR plants has increased. In order to evaluate the crack propagation of PWSCC, it is required to estimate stress distribution including residual stress and operational stress through the wall thickness of the Alloy 600 weld region. In a national project in Japan for the purpose of establishing residual stress evaluation method, two test models were produced based on a reactor vessel outlet nozzle of Japanese PWR plants. One (Test model A) was produced using the same welding process applied in Japanese PWR plants in order to measure residual stress distribution of the Alloy 132 weld region. The other (Test model B) was produced using the same fabrication process in Japanese PWR plants in order to measure stress distribution change of the Alloy 132 weld region during fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end. For Test model A, residual stress distribution was obtained using FE analysis, and was compared with the measured stress distribution. By comparing results, it was confirmed that the FE analysis result was in good agreement with the measurement result. For mock up test model B, the stress distribution of selected fabrication processes were measured using the Deep Hole Drilling (DHD) method. From these measurement results, it was found that the stress distribution in thickness direction at the center of the Alloy 132 weld line was changed largely during welding process of the safe end to the main coolant pipe.

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