bluuebunny/crossref_metadata_2025 huggingface dataset:
A Study of the Key Parameters Affecting the Design of Optimized Weld Overlays
Authors: Killian Doug , Mahmoud Samer , Xu Heqin , Noronha Silvester , Nana Ashok
Published: May 2012

The potential for primary water stress corrosion cracking (PWSCC) of large diameter austenitic nickel alloy components and their associated welds presents a particular problem for the nuclear industry due to a limited number of available options for mitigating or repairing large bore pressure boundary components such as reactor vessel, reactor coolant pump, and steam generator inlet or outlet nozzles. While a full structural weld overlay (FSWOL), as governed by ASME Code Case N-740, is commonly used to mitigate and repair small (4″) to medium (10″) bore piping assemblies employing Alloy 82/182 dissimilar metal welds, the large amount of weld metal that would be have to be deposited on large components (and the associated impact on outage schedule) makes this an unattractive strategy for managing the degradation of Alloy 600 type materials. An alternative design option, specifically developed for the mitigation and repair of large bore (30″) components, utilizes a thinner weld overlay whose thickness has been optimized to achieve a specific level of stress on the inside surface of the PWSCC susceptible material. According to ASME Code Case N-754, inside surface stresses should be limited to 10 ksi during the design phase of an optimized weld overlay (OWOL) in order to minimize the initiation or consequences of primary water stress corrosion cracking. With the increased inspection requirements of Code Case N-754 and the corresponding smaller crack growth design flaw size, and along with the reduced weld volume of an OWOL, as compared to a FSWOL, an optimized weld overlay is often the preferred technique for mitigating or repairing large bore piping components. This paper investigates the influence of various parameters on the effectiveness of an optimized weld overlay in satisfying its principle design objective, to reduce the inside surface stresses in PWSCC susceptible materials to no more than 10 ksi. Inherent design parameters are the thickness of the underlying pipe or weld, and the depth of any recorded or postulated weld repairs in the pre-overlay configuration of the welded joint. Explicit design parameters include the thickness of the overlay, the number of weld layers used to form the overlay, and the length of the overlay. Finite element analysis is used to calculate residual and operating stresses in a representative large bore reactor vessel coolant nozzle dissimilar metal weld for various combinations of design parameters. The overall objective of this study is to identify the key parameters influencing inside surface stresses, and thereby provide screening criteria for use in determining the applicability of the optimized weld overlay as a viable PWSCC mitigation or repair option for large bore primary pressure boundary components.
Fatigue Cycle Monitoring
Authors: Woods Kirby , Thomas Kenneth
Published: May 2012

The past and/or current approach to cycle count is a deviation from the current license basis (CLB)/design stress report, CENC-1150 Analytical Report, as applied in the fatigue evaluations for pressure and thermal cycles. These departures are relative to design projection versus actual operating experience in relation to “Cyclic Life” and the intent of ASME III, Section N-415 component suitability requirements for power operations of 40 years and subsequent life extension beyond to 60 years. In general, the practice of counting scram events does not accurately represent the “Cyclic Life” and is overly conservative, resulting in inaccurate projections of component reliability. The Nuclear Regulatory Commission (NRC) requires licensees to keep an account of the number of transient occurrences to ensure that design assumptions used in the fatigue evaluation required by ASME Section III are not exceeded. In the mid 1970s, typical plant technical specifications required “the number of the transients which are comparable to or more severe than the transients evaluated in the stress report Code fatigue usage calculations will be recorded in a log book.” In the mid 1980s this transitioned into counting of transients, “this program provides controls to track the Final Safety Analysis Report, Section III-3.5, cyclic and transient occurrences to ensure that components are maintained within the design limits” (NUREG-1434, Section 5.5.5). In addition, the evaluation of Time-Limiting Age Analysis per Title 10 Code of Federal Regulation Part 54, Section 54.21(c). Cycle counting is used to summarize lengthy, irregular load-versus-time histories by providing the number of times cycles of various amplitudes occur. The definition of a cycle varies with the method of cycle counting. In fatigue analysis, a cycle is the load variation from valley-to-peak-to-valley. Cycle counts can be made for time histories of force, stress, strain, torque, acceleration, deflection or other loading parameters of interest. In general, the operational transients plants experience do not approach the magnitude or severity of the bounding events used in these analysis. Thus counting every transient as equal to a design event is unnecessarily conservative. However, no clear specific guidance for Owners on how to perform cycle counting has been promulgated. This paper provides guidance for formulating a fatigue management program that takes advantage of the original analytical design stress reports.
Fatigue and SCC Lifetime Enhancement of Steels by Means of Laser Peening
Authors: Hackel Lloyd A. , Dane C. Brent , Rankin Jon , Harris Fritz , Truong Chanh , Marley Serena
Published: May 2012

Laser peening is a technology employed to induce deep levels of plasticity in materials resulting in compressive residual stress to depths of 5mm (0.2 inches) or deeper enabling increased fatigue strength and lifetimes and greatly enhanced resistance to stress corrosion cracking. An advanced in Nd:glass laser technology has enabled high rate, cost effective industrial deployment of the technology that is now being widely used in aerospace and electric power generation applications. Examples of deep levels of compressive stress and resulting improvements in fatigue life and resistance to stress corrosion cracking will be shown. An advanced beam delivery technology incorporating a robotic arm coupled with a video detection/beam alignment system is enabling rapid setup and application in field applications. The technology has been approved and deployed to prevent early fatigue cracking of key structural components on the F-22 fighter jet. It is also being applied to oil and natural gas applications and has the potential to serve a broad base of field applications including oil and gas lines, on-board ship SCC, nuclear power plant SCC, and fatigue and corrosion applications in upstream exploration and downstream oil refining.
Dynamics of a Free-Clamped Cylinder in Confined Axial Flow
Authors: Rinaldi Stephanie , Pai¨doussis Michael P.
Published: May 2012

The dynamics of a flexible cantilevered cylinder in confined axial flow is studied theoretically and experimentally, in the case where the flow is directed from the free end towards the clamped end. First, the dynamics is described, as observed in specially conducted experiments with air flow; the system developed small-amplitude first-mode vibrations at low flow velocities, which could be flutter, and at higher flows the oscillatory behaviour was succeeded by a static divergence. A simple, linear theoretical model is also developed, and the theoretically predicted behaviour is compared to the experimental one. The model captures the essentials of the observed behaviour, but requires improvement before quantitative prediction can be considered to be adequate.
Definition of “Small Gram Quantity Contents” for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations
Authors: Sitaraman Shiva , Kim Soon , Biswas Debdas , Hafner Ronald , Anderson Brian
Published: May 2012

This paper presents a compendium of allowable masses for a variety of gamma and neutron emitting isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded radioactive material transportation packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term “Small Gram Quantity” (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. Two sets of mass limit results are presented: (1) mass limits calculated with a “voided sphere” model, and (2) mass limits calculated with the unshielded radioactive material transportation packaging Model 9977-96.
Implementation of ASME Code, Section XI, Code Case N-770, on Alternative Examination Requirements for Class 1 Butt Welds Fabricated With Alloy 82/182
Authors: Sullivan Edmund J. , Anderson Michael T.
Published: May 2012

In May 2010, the U.S. Nuclear Regulatory Commission (NRC) issued a proposed notice of rulemaking (75 FR 24324) [1] that includes a new section to its rules to require licensees to implement ASME Code Case N–770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without the Application of Listed Mitigation Activities, Section XI, Division 1,” [2] with 15 conditions. Code Case N-770 contains baseline and inservice inspection (ISI) requirements for unmitigated Alloy 82/182 butt welds and preservice and ISI requirements for mitigated Alloy 82/182 butt welds. The NRC stated that application of ASME Code Case N-770 is necessary because the inspections currently required by the ASME Code, Section XI, were not written to address stress corrosion cracking of Alloy 82/182 butt welds, and the safety consequences of inadequate inspections can be significant. The NRC expects to issue the final rule incorporating this Code Case into its regulations toward the middle of 2011. This paper discusses the new examination requirements, the conditions that NRC proposed to impose, and potential areas of concern with implementation of the new Code Case.
Mismatch Effect on Fracture Driving Force in Mismatched Girth Welded Pipes
Authors: Castelluccio Gustavo M. , Cravero S. , Bravo R. , Ernst H.
Published: May 2012

Elasto-Plastic Fracture Mechanics (EPFM) is a useful tool for analyzing the structural integrity of components. However, EPFM has originally been developed for homogeneous materials and there are some concerns when it is applied to inhomogeneous materials. In the case of welds, the material fracture toughness and the applied fracture mechanics parameter on the structural member (J-integral, CTOD) should be adequately estimated. Furthermore, the mechanic mismatch influences on the local constraint may increase the risk of unstable failure. Hence, to study the effects of weld mismatch and crack locations on fracture behavior, single edge notch under tension (SE(T)) specimens and girth welded pipes under bending containing circumferential cracks were studied by means of finite elements simulations. Different weld widths and locations of cracks over the weld are considered. A study of the opening stresses ahead the crack tip developed in mismatched SE(T) specimens and cracked pipes allows the determination of the most critical combination of weld width and crack location in terms of applied J-integral and crack tip constraint level.
In-Service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution Per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock
Authors: Sullivan Edmund J. , Anderson Michael T. , Norris Wallace
Published: May 2012

The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking of a reactor pressure vessel (RPV) due to a pressurized thermal shock (PTS) event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” The §50.61a rule, which is optional, requires licensees to analyze the results from periodic volumetric examinations required by the American Society of Mechanical Engineers (ASME) Code. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory has been working on a program to assess the ability of current inservice inspection ultrasonic testing (UT) techniques, as qualified through the ASME Code to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, “Spirit of Appendix VIII reactor vessel examination,” a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.
Instrumented Indentation Testing to Evaluate High-Temperature Material Properties
Authors: Park Chan-Pyoung , Kim Kug-Hwan , Kang Seung-Kyun , Jo Won-Je , Kwon Dongil
Published: May 2012

Mechanical properties must be evaluated at high temperatures to predict high-temperature deformation and fracture behavior, since high-temperature properties differ greatly from those at room temperature. A high-temperature uniaxial tensile test, a representative high-temperature test, is generally used, but it has the limitation of obtaining merely the average material properties. Recently an advanced method for evaluating tensile properties has been developed: the instrumented indentation test (IIT), which simultaneously applies a load and measures displacement. Here we use instrumented indentation testing to evaluate the flow properties (yield strength, ultimate tensile strength, etc.) of heat-resistant steel at high temperature. The contact-area determination algorithm and representative stress-representative strain approach are applied for high temperatures. We compare our experimental results to those of conventional high-temperature uniaxial tensile testing to assess the high-temperature performance of the instumented indentation test.
Initiation Stress Threshold Irradiation Assisted Stress Corrosion Cracking Criterion Assessment for Core Internals in PWR Environment
Authors: Tanguy Benoit , Pokor Ce´dric , Stern Anthony , Bossis Philippe
Published: May 2012

Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material.
Probability of Initial Failure for Spring Operated Relief Valves
Authors: Bukowski Julia V. , Gross Robert E. , Goble William M.
Published: May 2012

We present clear and convincing evidence that, for new spring operated relief valves (SORV) that are not proof tested by the user shortly before installation, there is a non-trivial probability that the SORV will be installed in the fail-to-open (stuck shut) failure mode. Using the results of over 4800 new ASME Boiler and Pressure Vessel Code Section VIII SORV proof tests, we estimate the probability of initial failure (PIF) due to manufacturer/assembly anomalies, as well as PIF due to in-storage aging of SORV based on their material composition. We indicate how PIF can be reduced by various preinstallation activities that may be undertaken by the user. We show how to compute values of PIF to be used in calculating the average probability of fail danger (PFDavg) (as required by IEC61508 and similar safety standards in order to determine a safety integrity level (SIL)) which accounts for both the SORV material composition and the pre-installation activities undertaken. For four typical SORV of different material compositions we show how pre-installation activities influence the achievable SIL. We discuss the implication of these findings for estimating PIF for used (previously installed) SORV. We close with recommendations to further address PIF.
Effects of Graphite Surface Roughness on Bypass Flow Computations for an HTGR
Authors: Tung Yu-Hsin , Johnson Richard W. , Sato Hiroyuki
Published: May 2012

Bypass flow in a prismatic high temperature gas reactor (HTGR) occurs between graphite blocks as they sit side by side in the core. Bypass flow is not intentionally designed to occur in the reactor, but is present because of tolerances in manufacture, imperfect installation and expansion and shrinkage of the blocks from heating and irradiation. It is desired to increase the knowledge of the effects of such flow; it has been suggested that it may be 20% of the total helium coolant flow [INL report 2007, INL/EXT-07-13289]. Computational fluid dynamic (CFD) simulations can provide estimates of the scale and impacts of bypass flow. Previous CFD calculations have examined the effects of bypass gap width, level and distribution of heat generation and effects of shrinkage. The present contribution examines the effects of graphite surface roughness on the bypass flow for different relative roughness factors on three gap widths. Such calculations should be validated using specific bypass flow measurements. While such experiments are currently underway for the specific reference prismatic HTGR design for the next generation nuclear plant (NGNP) program of the U.S. Dept. of Energy, the data are not yet available. To enhance confidence in the present calculations, wall shear stress and heat transfer results for several turbulence models and their associated wall treatments are first compared for flow in a single tube that is representative of a coolant channel in the prismatic HTGR core. The results are compared to published correlations for wall shear stress and Nusselt number in turbulent pipe flow. Turbulence models that perform well are then used to make bypass flow calculations in a symmetric one-twelfth sector of a prismatic block that includes bypass flow. The comparison of shear stress and Nusselt number results with published correlations constitutes a partial validation of the CFD model. Calculations are also compared to ones made previously using a different CFD code. Results indicate that increasing surface roughness increases the maximum fuel and helium temperatures as do increases in gap width. However, maximum coolant temperature variation due to increased gap width is not changed by surface roughness.
An Investigation on the Reduction Schemes of Environmental Fatigue Usage Factor for Pressurizer Surge Line
Authors: Kang Sun-yeh , Jo Won-ho , Song Min-sup , Yoon Ki-seok , Choi Taek-sang , Yang Jun-seog
Published: May 2012

For plant life extension, it is the regulatory requirement to assess reactor coolant environmental impacts on critical components of the nuclear power plant including at least those mentioned in NUREG/CR-6260[2]. The pressurizer surge line is the most easy-to-fail component in view of LWR (Light Water Reactor) environments when it comes to meeting the current ASME code limit of the fatigue evaluation. Cumulative Usage Factor (CUF) value could be increased to a maximum of 15.35 times due to the environmental effects, which makes it easy to exceed the allowable fatigue limit (1.0). This paper discusses the process of the environmental correction factor calculation described in NUREG/CR-5704[4], and five proposed schemes for reducing the environmental CUF value to the ASME code limit or below. This paper concludes that the proposed schemes are effective in lowering the environmental CUF value of the pressurizer surge line.
Hydraulic Qualification and Testing of Nuclear Equipments and Installations: Technical Considerations by Application of European Harmonized Standard and Different Codes; KTA, ASME, RCC-M, Etc.
Authors: Bezdikian Georges
Published: May 2012

For new Nuclear Power plants construction in European Community, the evolutions between several codes and standards need precise attentions, especially for Hydro-test qualification and testing, considering that there were several manufacturers and vendors from different countries participating to the construction of a Nuclear Power Plant in One European country. It is identified that there are some differences on hydro-testing conditions considering: Harmonized standards; RCC-M code, ASME code, KTA code, Etc. This paper presents a comparison between different aspects of hydro-tests qualification and testing.
EDF Ageing Management: Status and Ongoing Developments
Authors: Faidy Claude
Published: May 2012

During the past 15 years many works have been done on Ageing Management Program (AMP) of Safety classed components in EDF. The paper will describe all the different aspects concerning these programs, and in particular the EDF step by step procedure and the major results. To-day, EDF is starting the 3rd ten-year shutdown of all these 3-loop plants (34 plants). During the associated Safety Review, a specific task is devoted to ageing effects and control of all the safety concerned components. A large list of components has been reviewed: mechanical, civil engineering, instrumentation and control, cables, non metallic components. Few non safety but important in term of availability components are considered, like turbine or some balance of plant components. A general review of results and difficulties for 40 and 60 years are presented in the paper. The second part is devoted to a short review of our procedure with IAEA guidelines. The third part will shortly describe the associated R&D program for metallic components and knowledge management associated to AMP. The major conclusions are clearly supporting the needs of international basic procedure and harmonization on the major topics.
Modeling of In-Situ Hydride Growth in Zr-2.5%Nb
Authors: Xu Feng , Metzger Don
Published: May 2012

The presence of Zr hydrides can greatly reduce the ductility and fracture toughness in a pressure vessel made of Zr alloy. Understanding how the hydrides form and grow is critical to flaw assessment for the pressure tubes. The process of hydride growth in Zr-2.5%Nb has been monitored in-situ in high-energy synchrotron X-ray radiation. The C-shaped specimen with a V-notch was held under constant load at a temperature where hydride formation was ensured. The development of hydride size and elastic strains in both the hydrides and the Zr matrix was recorded. Afterwards, the hydride morphology was characterized by Scanning Electronic Microscopy. A finite element program in combination with a process zone model and a diffusion model has been used to interpret the experimental data for better understanding the hydride growth process. The hydride length and morphology are well predicted, given that the information of hydride size and hydrostatic stress is properly updated and exchanged between the process zone and diffusion models. The effect of creep has been included in the modeling but found relatively small compared to that of hydride volumetric expansion. The elastic strains in Zr are well reproduced except that disagreement with the experiment is found in the hydrided region. This analysis provides further evidence that the process zone and diffusion models can be used to predict the hydride size and morphology development. Further modeling at a micro-structural level is needed for improving predictions of the stress/strain state in the hydrides, which is essential to the development of a sound hydride crack initiation model.
RCC-M, RSE-M and RCC-MRx: A Consistent Set of Mechanical Components Codes and Standards
Authors: Faidy Claude
Published: May 2012

During the past 25 years EDF, AREVA and CEA developed a complete set of Codes and Standards for Nuclear Power Plants inside the French Association for Nuclear Codes and Standards (AFCEN). The list of these Codes are: RCC-M (Design of PWR Mechanical Components), RSE-M (Surveillance and Operation of PWR Mechanical Components), RCC-E (Electrical Components), RCC-G (Civil Engineering), RCC-I (Fire Protection), RCC-C (Design of PWR fuel) and in connection with CEA RCC-MR (Design of HTR Mechanical Components) and RCC-Mx (Design of Experimental Reactor Mechanical Components). This paper will present major Codes for Mechanical Components and the future developments of these Codes. For all these Codes, revisions are generally done every year and are associated to: - needs for new project, - recent R&D results, - regular user requests, - field experience. For all these different disciplines, modernization and State of the Art justification have to be justified. In parallel to that all these Codes have to be used in different international projects, and need some particular developments (as specific appendices) to assure usability under different national regulations and practices.
An Improved Shell Theory Applied for Failure Analysis of Pressure Vessels
Authors: Liu R. , Zhao J. , Wu X. J.
Published: May 2012

Crack-face closure occurs physically at the compressive edges when a shell is subjected to bending loads. However, in traditional shell theories, crack closure effects are not concerned when evaluating the stress intensity factor (SIF). In reality, crack closure effects influence significantly the SIF. This article presents the theoretical and numerical analyses of crack-face closure effects on the stress intensity factor of shells under bending. The theoretical formulation is based on the shallow shell theory of Delale and Erdogan, incorporating the effects of crack-face closure, which are modeled by a line contact at the compressive edges of the crack faces. It is shown that due to curvature effects crack closure in shells may not occur on the entire length of the crack, depending on the nature of the bending loading and the geometry of the shell. To validate the theoretical solution finite element analysis (FEA) is also performed; the two results agree well. As an example, the stress intensity factor for a pressurized cylinder containing an axial crack is determined based on the improved shell theory which takes into account the effects of crack-face closure.
French Nuclear Reactor Pressure Vessel Integrity Assessment and Life Management Strategy
Authors: Churier H. , Balard G. , Meister E. , Clemendot F. , Todeschini P. , Pages C.
Published: May 2012

The structural integrity of the RPV is an essential issue for the plant safety. At the design stage, the demonstration is required with material properties at end of life, to ensure the adequacy of the design with the expected operating transients in all conditions. During operation, the integrity assessment is updated every ten years with new existing knowledge and feedback of operating experience, in particular in service aged material data coming from Irradiation Surveillance Program ISP, fluence evaluations taking into account the effective in service core arrangements in each vessel, in service detected flaws plus a postulated subclad crack whose detection cannot be guaranteed by the qualified ISI program. The final assessment showed that the regulatory criteria are met until the fourth decennial outage for 900 MW RPV. The analysis is performed in accordance with French regulations (use of safety coefficients) and follows a deterministic approach in which the input parameters and uncertainties are taken into account conservatively. For the future demonstration beyond 40 years, a multidisciplinary effort is committed to improving knowledge in order to reduce uncertainties in data and in methods. This extensive program involves in particular: - Thermohydraulic analysis and description of transients: temperatures and heat exchange coefficients; - Mechanical analysis: warm pre-stress effect and crack arrest. In addition, a complementary study using a probabilistic approach to rationalize the level of conservatism of input data is launched. In this report, the French deterministic approach and the main results for 40 years duration are presented and the new developments for the future.
Degradation Susceptibility Metrics as the Bases for Bayesian Reliability Models of Aging Passive Components and Long-Term Reactor Risk
Authors: Unwin Stephen D. , Lowry Peter P. , Toyooka Michael Y. , Ford Benjamin E.
Published: May 2012

Conventional probabilistic risk assessments (PRAs) are not well-suited to addressing long-term reactor operations. Since passive structures, systems and components are among those for which refurbishment or replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant. Yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA does present challenges. First, service data required to quantify component reliability models are sparse, and this problem is exacerbated by the greater data demands of age-dependent reliability models. A compounding factor is that there can be numerous potential degradation mechanisms associated with the materials, design, and operating environment of a given component. This deepens the data problem since the risk-informed management of materials degradation and component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates the metrics of materials degradation susceptibility being developed under the Nuclear Regulatory Commission’s Proactive Materials Degradation Assessment Program with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by the predicted long-term operational risk.
Statement of the French Safety Authority Concerning the Behaviour of 900 MWe PWR RPV Between 30 and 40 Years in Operation
Authors: Monin L.
Published: May 2012

The French 900 MWe reactor pressure vessel (RPV) integrity demonstration was reviewed in 2010 by the French Nuclear Safety Authority (ASN), together with its technical support organisation, the French Institute for Radiation Protection and Nuclear Safety (IRSN). The purpose was to decide whether these 3-loop PWR RPV’s are able to remain in service for ten years after their third decennial outage (VD3). ASN and IRSN examined the demonstration developed by Electricite´ De France (EDF) in order to ensure the compliance of in-service RPV behaviour with regulatory requirements and to check the validity of the calculations and hypotheses that were used for that purpose. The analysis aimed notably at ensuring that the results of every calculation step were conservative and that regulatory safety margins were applied. The calculations made by EDF showed that regulatory criteria were met until the provisional deadline of their fourth decennial outage (VD4). Thus ASN considered that the serviceability of all 900 MWe PWR RPV’s is guaranteed throughout the decade following the VD3 of the reactors, considering that EDF respects ASN demands with a view to comfort safety margins.
Residual Stress Measurements Through the Thickness of the Dissimilar Weld Pipe Using Neutron Diffraction
Authors: Woo Wanchuck , Em Vyacheslav , Lee Ho-Jin , Seong Baek-Seok , Park Kwang Soo , Byeon Jin Gwi , Hubbard Camden R.
Published: May 2012

The distribution of residual stresses was determined in an overlay dissimilar joining pipe weld using neutron diffraction. The specimen was dissimilarly welded between the bcc ferrite steel (SA508) and fcc austenite (SA182) steel with the Ni-based welding consumable (Alloy 182). The weld pipe simulates the nozzle joint component of the nuclear power plants with the dimension of about 130-mm diameter, 500-mm length, and 21-mm thickness. A total of 13 positions were measured from 2 mm to 20 mm underneath the pipe outer wall with 1∼2 mm steps along the weld centerline. The neutron beam gauge volume provides 1-mm spatial resolution along the thickness direction of the weld pipe. The macroscopic ‘stress-free’ lattice spacing (do) was also measured using 2-mm width comb teeth coupons taken from the identical location of the bulk weld neutron-diffraction measurements. The result shows that the hoop stress component developed tension of about 100 MPa and compression of −600 MPa near the outer and inner wall surface of the overlay pipe weld, respectively.
Practical Application of RBI and Experiential Analysis to Sulfur Recovery Unit Inspection
Authors: Wang Jifeng , Xue Xiaolong , Qian Yaozhou , Yu Houde , Tang Xiaoying
Published: May 2012

According to the management and scheduled inspection of pressure vessels and piping in sulfur recovery unit, it described the evaluation of the failure probability and consequences of pressure vessels and piping by the Risk Based Inspection (RBI). This article focuses on the practical application process of RBI to the Sulfur Recovery Unit. Together with the production crafts and loop medium in pressure vessels and pipes, we could dig out high risk equipment and its’ reasons, therefore, suggest possible precautions. Further precautions will be provided for equipment manufacturers to assure production safety through personal experiential analysis and comparison with the similar product from the aspect of failure mechanism and position. It is aimed to explore the combined method of RBI and personal experiential analysis and it’s practical application.
Healable and Repeatable Adhesively Bonded Joint
Authors: Ji Gefu , Ouyang Zhenyu , Li Guoqiang , Ibekwe Samuel , Pang Su-Seng
Published: May 2012

The adhesively bonded structure has to be replaced after the crack initiation and propagation. In a previous study, a biomimic two-step self-healing scheme (close-then-heal) by mimicking human skin has been proposed for self-healing structural-length scale damage. The adhesively bonded joint are prepared and to invest its feasibility and repeatability by fabricating a composite adhesive bonded joint with thermoplastic particles dispersed in a most commonly used epoxy based adhesive material. The fractured specimens were healed per the close-then-heal mechanism and tested again to fracture. This fracture-healing test lasted for 3 cycles.
Effects of Adhesive Thickness on Global and Local Mixed Mode I/II Interfacial Fracture of Bonded Steel Joints
Authors: Ji Gefu , Ouyang Zhenyu , Li Guoqiang , Xu Wei , Jerro Dwayne , Pang Su-Seng
Published: May 2012

Interfacial toughness and interfacial strength, as two critical parameters in an interfacial traction-separation law, have important effect on the fracture behaviors of bonded joints. In this work, the global and local tests are employed to investigate the effect of adhesive thickness on interfacial energy release rate, interfacial strength, and shapes of the interfacial traction-separation laws. Basically, the measured laws in this work reflect the equivalent and lumped interfacial fracture behaviors which include the cohesive fracture, damage and plasticity. The experimentally determined interfacial traction-separation laws may provide valuable baseline data for the parameter calibrations in numerical models. The current experimental results may also facilitate the understanding of adhesive thickness-dependent interface fracture of bonded joints.
Evaluation of a Novel Microwave Based NDT Inspection Method for Polyethylene Joints
Authors: Murphy Ken , Lowe David
Published: May 2012

Exova is continuing in the long tradition of Pipeline Developments Limited (PDL) in the UK with research, development, testing, quality assurance (QA) and failure analysis of modern plastic pipe materials and joints. Along with this ongoing work Exova has assessed various non destructive testing (NDT) techniques in an attempt to further improve the reliability of Polyethylene (PE) joints in the field. In the past, reliable inspection of PE pipe joints using NDT has proven to be difficult, but this has generally been based upon the use of existing metallic inspection methods, including radiography and ultrasound. However neither radiography nor ultrasound can reliably detect key issues that are known to affect PE joint quality such as fine particulate contamination and cold fusion in butt welds and misalignment and contamination in electrofusion joints. Advances in technology have seen the development of a new non destructive testing (NDT) technique using low voltage microwave imaging. Microwave imaging has been developed in North America by Evisive Inc to inspect modern dielectric materials such as PE. Comprehensive laboratory testing indicated that this technique could be used reliably and repeatedly to detect issues related to poor PE joint fabrication in butt welds and electrofusion joints. Issues which have been identified include fine particle contamination, lack of fusion and misalignment. More recently field trials have been carried out on various PE joints, to establish a correlation between the microwave NDT results obtained and mechanical testing results.
Structural Analysis of Mechanical Modules for Modular Designed Nuclear Plants
Authors: Sherfey Steven K.
Published: May 2012

Most nuclear power plants being designed and constructed in the world today utilize advanced light water reactors with improved economics and safety; they are referred to by the US Department of Energy as Generation III+ Nuclear Plants. Overall, the Generation III+ power plants are expected to be safer and more affordable than those presently in operation. The offsite construction of structural and mechanical modules is a key element of the Generation III+ plant design; this feature significantly reduces the amount of onsite laborers and compresses the construction schedule. Each mechanical module must be structurally qualified to support its attached components for transportation, lifting, and operation scenarios. Qualification of the modules is very complicated because of the applicable codes and criteria as well as the diversity of components that may be attached to them. The purpose of this paper is to provide analysis instruction and to recommend special modeling techniques for structurally analyzing mechanical modules; the recommendations provided in this paper should not be taken as absolute rules but rather as guidelines to be altered, as needed, in order to more accurately simulate specific plant requirements. A mechanical module may be classified as a large gang hanger which supports many system components; a module may have dozens of pipe supports attached to it as well as tanks, piping, valves, pumps, conduit, ductwork, and cable trays. Mechanical modules are a fundamental aspect of the Generation III+ plant, and therefore must be properly analyzed and qualified. Due to the practically infinite possible arrangements of structural members and components, special modeling techniques are often required for considering all the possible loadings that may exist for the transportation, lifting, and operation scenarios. A structural analysis computer program such as GTStrudl or StaadPro must be used to build an analytical model of the module; the module frame and its attached components must be simulated in the model. Loadings such as dead weight, live load, thermal expansion, earthquake, wind, pressure, and flow transients are commonly applied to modules and their components. The analysis and qualification of the mechanical module frame must address structural member stresses, weld stresses, connection local stresses, and module support design. Results of the module qualification must be documented and verified according to plant procedures and criteria.
Comparison of German KTA and ASME Nuclear Design Codes for Class 1, 2, 3 Components and Piping
Authors: Hofer Daniel , Schau Henry , Karabaki Hu¨seyin Ertugrul , Hill Ralph
Published: May 2012

This paper compares the design rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Rules for Construction of Nuclear Facility Components, with German nuclear design standards for Class 1, 2, 3 components and piping. The paper is focused on a comparison of the equations for Design by Analysis and on Piping equations. The ASME Section III Code has been used in combination with design specifications for design of German nuclear power plants. Together with manufacturers, inspectors and power plant owners, the German regulatory authority decided to develop their own nuclear design standards. The current versions being used are from 1992 and 1996. New versions of KTA design standards for pressure retaining components (KTA 3201.2 and KTA 3211.2) are currently under development. This comparison will cover the major differences between the design rules for ASME Section III, Div. 1 and KTA standards 3201.2 and 3211.2 as well as code or standard organization by sections, paragraphs, articles and code development.
Predicting Macroscopic Dynamics in Large Distributed Systems: Part II
Authors: Mills K. L. , Filliben J. J. , Cho D.-Y. , Schwartz E. J.
Published: May 2012

Society increasingly depends on large distributed systems, such as the Internet and Web-based service-oriented architectures deployed over the Internet. Such systems constantly evolve as new software components are injected to provide increased functionality, better performance and enhanced security. Unfortunately, designers lack effective methods to predict how new components might influence macroscopic behavior. Lacking effective methods, designers rely on engineering techniques, such as: analysis of critical algorithms at small scale and under limiting assumptions; factor-at-a-time simulations conducted at modest scale; and empirical measurements in small test beds. Such engineering techniques enable designers to characterize selected properties of new components but reveal little about likely dynamics at global scale. In this paper, we outline an approach that can be used to predict macroscopic dynamics when new components are deployed in a large distributed system. Our approach combines two main methods: scale reduction and multidimensional data analysis techniques. Combining these methods, we can search a wide parameter space to identify factors likely to drive global system response and we can predict the resulting macroscopic dynamics of key system behaviors. We demonstrate our approach in the context of the Internet, where researchers, motivated by a desire to increase user performance, have proposed new algorithms to replace the standard congestion control mechanism. Previously, the proposed algorithms were studied in three ways: using analytical models of single data flows, using empirical measurements in test beds where a few data flows compete for bandwidth, and using simulations at modest scale with a few sequentially varied parameters. In contrast, by applying our approach, we simulated configurations covering four-tier network topologies, spanning continental and global distances, comprising routers operating at state-of-the-art speeds and transporting more than 105 simultaneous data flows with varying traffic patterns and temporary spatiotemporal congestion. Our findings identify the main factors influencing macroscopic dynamics of Internet congestion control, and define the specific combination of factors that must hold for users to realize improved performance. We also uncover potential for one proposed algorithm to cause widespread performance degradation. Previous engineering studies of the proposed congestion control algorithms were unable to reveal such essential information.
Cofermentation of Glucose, Xylose, and Cellobiose by the Beetle-Associated Yeast Spathaspora passalidarum
Authors: Long Tanya M. , Su Yi-Kai , Headman Jennifer , Higbee Alan , Willis Laura B. , Jeffries Thomas W.
Published: May 2012

Fermentation of cellulosic and hemicellulosic sugars from biomass could resolve food-versus-fuel conflicts inherent in the bioconversion of grains. However, the inability to coferment glucose and xylose is a major challenge to the economical use of lignocellulose as a feedstock. Simultaneous cofermentation of glucose, xylose, and cellobiose is problematic for most microbes because glucose represses utilization of the other saccharides. Surprisingly, the ascomycetous, beetle-associated yeastSpathaspora passalidarum, which ferments xylose and cellobiose natively, can also coferment these two sugars in the presence of 30 g/liter glucose.S. passalidarumsimultaneously assimilates glucose and xylose aerobically, it simultaneously coferments glucose, cellobiose, and xylose with an ethanol yield of 0.42 g/g, and it has a specific ethanol production rate on xylose more than 3 times that of the corresponding rate on glucose. Moreover, an adapted strain ofS. passalidarumproduced 39 g/liter ethanol with a yield of 0.37 g/g sugars from a hardwood hydrolysate. Metabolome analysis ofS. passalidarumbefore onset and during the fermentations of glucose and xylose showed that the flux of glycolytic intermediates is significantly higher on xylose than on glucose. The high affinity of its xylose reductase activities for NADH and xylose combined with allosteric activation of glycolysis probably accounts in part for its unusual capacities. These features makeS. passalidarumvery attractive for studying regulatory mechanisms enabling bioconversion of lignocellulosic materials by yeasts.