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.

Materials Aging Degradation of Reactor Vessel Internals: Part III—Structural Evaluation and Local Finite Element Models

2009 ◽  
Author(s):  
Matthew D. Snyder ◽  
Tama´s R. Liszkai ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Structural Analysis ◽  
Constitutive Model ◽  
Reactor Vessel ◽  
Finite Element Models ◽  
Local Models ◽  
Bolted Connections ◽  
The Core ◽  
Potential Failure ◽  
Reactor Internals

Pressurized water reactor (PWR) internals components can experience material aging and degradation due to irradiation. The purpose of the functionality analysis is to provide a best-estimate evaluation of the reactor internals core barrel assembly for materials degradation to see if the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis [1] representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor internals. In order to evaluate the potential failure mechanisms of the core barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox-type (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the core provide essential information on radiation dose and heat rates of the internals. The computational fluid dynamics domain (CFD) allows evaluation of the internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the internals components and bolted connections is the third major physics field involved, which facilitates the development of operating stress fields within the internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections. Therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [2] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, to be presented at PVP 2009 [2] describes global structural finite element models. Part III, presented in this paper, presents a description of local models of bolted connections, results, and conclusions.

Materials Aging Degradation of Reactor Vessel Internals: Part II—Structural Evaluation and Global Finite Element Models

2009 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Matthew Snyder ◽  
Anne Demma
Keyword(s):  
Finite Element ◽  
Structural Analysis ◽  
Constitutive Model ◽  
Reactor Vessel ◽  
Finite Element Models ◽  
Local Models ◽  
Bolted Connections ◽  
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), developed PWR Internals Inspection and Evaluation (I&E) Guidelines mainly to support license renewal of U.S. plants [2]. The functionality analysis of reactor internals components and assemblies was one of the tools used to develop these guidelines. The purpose of the functionality analysis is to provide a best estimate evaluation of the reactor internals core barrel assembly for materials degradation and to assess whether the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement [3], 4, [5]. This constitutive model is a function of temperature and fluence. The analysis 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, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential 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 analytical calculations. 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 analytical calculation, which facilitates the development of operating stress fields within the RV internals. Structural analysis was performed as two parts. First, a global structural model of the core barrel assembly was used to represent the interaction of components of the core barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections, therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [6] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, detailed in this paper, describes global structural finite element models. Part III, to be also presented at PVP-2009 [7], presents a description of local models of bolted connections, results, and conclusions.

Application of Aging Management Strategies for Reactor Vessel Internals and Core Support Structures

2010 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Matthew Snyder ◽  
Steve Fyfitch ◽  
Hongqing Xu ◽  
Hasan Charkas
Keyword(s):  
Degradation Mechanism ◽  
Management Strategies ◽  
Operating Experience ◽  
Management Program ◽  
Reactor Vessel ◽  
Lessons Learned ◽  
Support Structures ◽  
Acceptance Criteria ◽  
Pressurized Water ◽  
Reactor Internals

The Materials Reliability Program (MRP) Reactor Internals Focus Group (RI-FG) developed Pressurized Water Reactor (PWR) Internals Inspection and Evaluation (I&E) Guidelines under the sponsorship of the Electric Power Research Institute (EPRI). The I&E guidelines summarized in MRP-227 [1], provide a generic basis for U.S. utilities to develop their Aging Management Program (AMP) for managing the long-term aging degradation of PWR reactor internals including the existing and extended license periods. A number of internals structural bolts in the Babcock & Wilcox (B&W) design PWRs are fabricated from high-strength alloys such as Alloy A-286 or Alloy X-750. The materials in general, and bolts in particular, are known to be susceptible to stress corrosion cracking (SCC) based on past operating experience. The Upper and Lower Core Barrel (UCB and LCB) bolts have a core support function and have been generically categorized as Primary components for inspection in the I&E Guidelines. The remaining Alloy A-286 and Alloy X-750 structural bolts are in the Expansion category. Per 10CFR54, all U.S. PWRs are required to establish a unit-specific AMP for the extended license period in accordance with the ten elements of an effective AMP outlined in the Generic Aging Lessons Learned (GALL, NUREG-1801 Rev. 01, [2]) report published by the U.S. Nuclear Regulatory Commission (NRC). The goal of this paper is to provide an overview of the work performed by AREVA NP Inc. to support the development of the MRP I&E guidelines and unit-specific AMP for UCB and LCB bolts. A review of Alloy A-286 and Alloy X-750 bolts in the B&W design PWR is provided including the degradation mechanism, operating and inspection experience, replacement, and autoclave and in-reactor test results. The latest UT inspection technique used to characterize the extent of flaws is also discussed. Acceptance criteria for evaluating degraded conditions in UCB and LCB bolts were developed in accordance with the requirements of the ASME Section III, Subsection-NG core support structures requirements. In addition to Code compliance, special limits were established to limit the change in the core support structure stiffness. The acceptance criteria enable utilities to rapidly disposition UT inspection findings during an outage within 48 hours. In order to support the objectives of an efficient AMP for the UCB and LCB bolts, three-dimensional finite element models were prepared capable of evaluating all potential failure scenarios. These models enable accurate representation of flange flexibility and redistribution of loads due to deficient bolts. Prior to an outage, hypothetical patterns of bolt failures could be evaluated to support pre-outage planning and contingency preparation. During an outage, these models are used to disposition inspection results and help operability assessment of continued operation, and re-inspection requirement to ensure continued safety and integrity of the reactor vessel internals. Based on the existing work performed, future improvement and expansion of analytical capability is outlined in the last section of this paper. In conclusion, AREVA NP Inc. has demonstrated an effective use of a multi-disciplined approach using structural analyses, operating experience, material evaluations, and non-destructive examination (NDE) to fulfill both the development and implementation of unit-specific aging management commitments as required by MRP-227 for the current and extended license periods.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Aging Management Strategies for Pressurized Water Reactor Vessel Internals

2004 ◽  
Author(s):  
Timothy J. Griesbach ◽  
Robert E. Nickell ◽  
H. T. Tang ◽  
Jeff D. Gilreath
Keyword(s):  
Reactor Vessel ◽  
Performance Criteria ◽  
Pressurized Water Reactor ◽  
Aging Effects ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
License Renewal ◽  
Technical Basis ◽  
Reactor Internals

Management of materials aging effects, such as loss of material, reduction in fracture toughness, or cracking, depends upon the demonstrated capability to detect, evaluate, and potentially correct conditions that could affect function of the internals during the license renewal term. License renewal applicants in their submittals to NRC have identified the general elements of aging management programs for Pressurized Water Reactor (PWR) internals, including the use of inservice inspection and monitoring with the possibility of enhancement or augmentation if a relevant condition is discovered. As plants near the license renewal term, plant-specific aging management programs will be implemented focusing on those regions most susceptible to aging degradation. A framework for the implementation of an aging management program is proposed in this paper. This proposed framework is based on current available research results and state of knowledge and utilizes inspections and flaw tolerance evaluations to manage the degradation issues. The important elements of this framework include: • The screening of components for susceptibility to the aging mechanisms, • Performing functionality analyses of the components with representative material toughness properties under PWR conditions, • Evaluating flaw tolerance of lead components or regions of greatest susceptibility to cracking, loss of toughness, or swelling, and • Using focused inspections to demonstrate no loss of integrity in the lead components or regions of the vessel internals. The EPRI Material Reliability Program (MRP) Reactor Internals Issue Task Group (RI-ITG) is actively working to develop the data and methods to quantify an understanding of aging and potential degradation of reactor vessel internals, to develop materials/components performance criteria, and to provide utilities tools for extending plant operations. Under this MRP Program, the technical basis for the framework will be documented. Then, based on that technical basis, PWR internals inspection and flaw evaluation guidelines will be developed for plants to manage reactor internals aging and associated potential degradation.

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.

Finite Element Analysis of Freely Rotating Tires

1987 ◽  
Vol 15 (2) ◽  
pp. 134-158 ◽  
Author(s):  
N-T. Tseng
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Failure Modes ◽  
Deformable Body ◽  
Equations Of Motion ◽  
Material Model ◽  
Angular Speed ◽  
Element Analysis ◽  
Automobile Tire ◽  
Truck Tire

Abstract Axisymmetric analysis of an inflated tire rotating with constant angular speed can be used to simulate two loading conditions of a tire during its service life: (1) a freely rotating tire on an automobile that is stuck in snow or mud and (2) the top region of a rolling loaded tire, where footprint loading has little influence on the distribution of its stresses and strains. The equations of motion for a freely rotating deformable body with constant angular speed have been derived and implemented into a finite element code developed in-house. The rotation of a thin disk was used to check the validity of the implemented formulation and coding. Excellent agreement between the numerical and the analytical results was obtained. A cast tire, a radial automobile tire, and a radial truck tire, were then analyzed by the new finite element procedure. The tires were inflated and rotated at speeds up to 241 km/h (150 mph). The elastomers in these tires were simulated by incompressible elements for which the nonlinear mechanical properties were described by the Mooney-Rivlin model. Each ply was simulated by its equivalent orthotropic material model. The finite element predictions agreed well with the available experimental measurements. Significant changes in interply shear strain at the belt edge, the bead load, and the crown cord loads of plies were observed in the finite element analysis. These phenomena suggest three possible failure modes in freely rotating tires, i.e. belt edge separation, bead breakage, and belt failure at crown region.

Structural Integrity Assessment to Justify Increase in Upper Operating Temperature Limit of AGR Dome

2017 ◽  
Author(s):  
Antony M. Hurst
Keyword(s):  
High Temperature ◽  
Failure Modes ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Manganese Steel ◽  
Outlet Temperature ◽  
Original Design ◽  
The Core ◽  
Safety Case ◽  
Potential Failure

The UK advanced gas cooled reactors (AGRs) use a graphite core with carbon dioxide gas as the primary coolant. There is a diaphragm above the core which separates re-entrant gas at lower temperature and higher pressure from that leaving the channel guide tubes at reactor outlet temperature. This diaphragm is known as the hot box dome. The dome is perforated to facilitate the passage of fuel and control rods into the core. The dome is fabricated in carbon-manganese steel and incorporates a number of full penetration welds which are post-weld heat treated (PWHT). The dome’s upper surface is insulated to protect it from gas at high temperature, intended to maintain the dome at a temperature of below 380°C. Since dome failure could conceivably result in gas by-passing and, hence, failing to adequately cool the core the original safety case claims that gross failure of the dome is incredible. More recently potential failure modes of the dome have been reviewed and various dome weld failure scenarios have been analysed and assessed to demonstrate a tolerance to the consequences of complete failure of certain welds. On this basis the dome could be shown to satisfy a lesser classification of high integrity, although no claim to reclassify the region has been made. Through-life temperature monitoring is carried out to demonstrate that the peak dome temperature remains below 380°C. This has shown evidence of rising temperatures, believed to arise from a reduction in the effectiveness of the upper surface insulation, an effect that was acknowledged by the original design. Work to investigate this effect has developed the understanding of the dome thermal environment which is far more complex than previously thought. The hottest parts of the dome are far smaller and more localised than previously thought, and lie immediately above the monitored locations. In order to support a case to operate for an extended life, it is now proposed that the upper temperature limit could safely be increased to 390°C. Structural integrity analyses and assessments have been carried out to support the proposed increase to 390°C and include a demonstration of the absence of a cliff-edge effect by assessing cases with the hottest parts of the dome at temperatures of 400 and 410°C. The work seeks to demonstrate adequate margins of safety against all potential failure modes. ASME III code assessment against primary stress limits has been used to guard against failure by plastic collapse and/or creep rupture. Creep-fatigue initiation assessments have been used to demonstrate margins against the formation of defects using the EDF Energy high temperature assessment procedure, R5. This has enabled the consideration of potential defects to be confined to those that might have formed during welding or PWHT and have been missed by extensive pre-service inspections. Notwithstanding the low likelihood that any such defects exist, with high confidence it may be postulated that any that do would be located in welds and be of limited size. Defect tolerance assessments have been carried out, including the calculation of limiting defect sizes in accordance with the EDF Energy R6 procedure and the growth of postulated defects by creep and fatigue using R5. Other failure and degradations mechanisms have been considered and eliminated as a potential threat by drawing on reviews of relevant operating experience on other reactors with similar materials and environments, and material property data from long term tests. This paper describes how the multi-facetted programme of work, which proposes a modification to an existing safety case, has been devised to explicitly address all conceivable modes of failure and demonstrate a robust argument against each one.

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.

scholarly journals Static Behaviors and Applications of Buckling Monitoring Members with Rigid Ends

Processes ◽  
2021 ◽  
Vol 9 (5) ◽  
pp. 836
Author(s):  
Wudang Ying ◽  
Changgen Deng ◽  
Chenhui Zhang
Keyword(s):  
Finite Element ◽  
Bearing Capacity ◽  
Failure Modes ◽  
Slenderness Ratio ◽  
Ultimate Bearing Capacity ◽  
Spatial Structures ◽  
Core Tube ◽  
The Core ◽  
Parametric Studies ◽  
Compression Member

The buckling of compression members may lead to the progressive collapse of spatial structures. Based on the sleeved compression member, the buckling monitoring member is introduced to control the buckling of compression member and raise buckling alert by sensing contact between the core tube and the restraining tube. Considering the rigid connection among the members in spatial structures, the buckling monitoring member with rigid ends needs to be further analyzed. An experimental test was conducted and finite element analyses were performed with calibrated finite element models. The results indicated that the ultimate bearing capacity and post-ultimate bearing capacity of the core tube were enhanced due to the restraint from the restraining tube. The contact was successfully sensed by pressure sensor, revealing that it sensed the buckling of the core tube. Parametric studies were conducted, indicating that the core protrusion, core slenderness ratio, the gap between the core tube and restraining tube, and the flexural rigidity ratio are the key parameters affecting the bearing capacity and the failure modes of the buckling monitoring member, and some key values of parameters were proposed to obtain good bearing capacity. Based on the parametric studies, the failure modes of buckling-monitoring members are summarized as global buckling and local buckling. The stress distribution and deformation mode of buckling monitoring members are presented in the non-contact, point-contact, line-contact, reverse-contact and ultimate bearing state. The buckling monitoring member is applied in a reticulated shell by substituting the buckling members. It can effectively improve the ultimate bearing capacity of reticulated shell.

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