bluuebunny/crossref_metadata_2025 huggingface dataset:
SCC Initiation Model and Its Implementation for Probabilistic Assessment
Authors: Garud Yogendra S.
Published: January 2011

An engineering model was recently developed to account for the inter-dependent influence of applied plus residual stress, material strength, strain-hardening, and cold work on stress corrosion cracking (SCC) susceptibility and time to initiation. Both the experimental observations and the model response suggest that a careful and combined consideration of these factors is essential in a quantitative or predictive assessment of SCC. These findings and the basic model are presented in this paper. Also, the proposed modeling framework of SCC susceptibility distinguishes, in the engineering sense, the influence of prior or additional cold work on the residual macro-stress and the microstructure, in addition to the role of (raised) strength properties. The model presented in this paper was evaluated for typical Alloy 600 and Type 304 stainless steel in high purity water environment. Additional objective of this paper is to demonstrate the utility and implementation of the model in a probabilistic assessment of the initiation of a short crack of engineering significance. It is shown that the basis and framework of the model provide a simple approach to quantify the key uncertainties in SCC initiation leading to results of interest in a probabilistic or risk-based methodology for SCC evaluation. Application of the results and possible implementation of the model are discussed with reference to the SCC of materials in the light water reactor environments.
Analytical Method for the Calculation of J Parameter in Cracked Components: Presentation of the RSE-M/RCC-MR Procedure and Benchmark Proposal
Authors: Marie S. , Drubay B. , Le Delliou P. , Chapuliot S. , Deschanels H. , Barthelet B. , Schwab J.
Published: January 2011

RSE-M and RCC-MR codes provide flaw assessment methodologies and related tools for Nuclear Power Plant. For these two codes, AREVA, CEA and EDF developed a large set of compendia for the calculation of the parameter J for various components (plates, pipes, elbows, …) and various defect geometries. The last step of these developments deals with the weld joints: since 2004, a methodology have been developed to calculate the J parameter for a defect located in a weld, less conservative than usual methods. This methodology is based on the definition of an equivalent material that leads to the same J value (with same loading conditions and defect geometry) than the bi-material component. The stress-strain curve of this equivalent material is deduced from a combination of the tensile curves of the base metal and of the weld metal. The weigth coefficients applied are specifically defined for the J calculation and generalized to deal with any weld joint geometry.
Flaw Testing of Fiber Reinforced Composite Pressure Vessels
Authors: Makinson John , Newhouse Norman L.
Published: January 2011

The ASME BPV Project Team on Hydrogen Tanks, in conjunction with other ASME Codes and Standards groups, is developing Code Cases and revisions to the Boiler & Pressure Vessel Code, including such to address the design of composite pressure vessels. The Project Team had an interest in further understanding the effect of cuts to the surface of composite tanks, and how the burst pressure would be affected during the lifetime of the pressure vessel. A test program was initiated to provide data on initial burst pressure, and burst pressure after pressure cycling, of composite cylinders with cuts of different depth. This test program was conducted by Lincoln Composites under contract to ASME Standards Technology LLC, and was funded by NREL. These results were considered during the development and approval of the ASME Code Cases and Code Rules. Thirteen pressure vessels with a design pressure of 24.8 MPa (3600 psi), approximately 0.406 meter (16.0 inches) in diameter and 1.02 meters (40.2 inches) long, were tested to investigate the effects of cuts to the structural laminate of a composite overwrapped pressure vessel with respect to cycling and burst pressure. Two flaws, one longitudinal and one circumferential, were machined into the structural composite. The flaws were 57 mm long by 1 mm wide (2.25 inch × 0.04 inch) and varied in depth from 10% to 40% of the structural composite thickness of 11.4 mm (0.45 inch). These pressure vessels were cycled to design pressure 0, 10,000 and 20,000 times then burst. The resulting burst pressures were evaluated against the performance of a pressure vessel without flaws or cycling. The burst pressures were affected by depth of cut, but the pressure cycling did not have a significant effect on the burst pressure.
Fatigue Behavior of Circumferentially Notched Round Bars of Austenitic Stainless Steel Under Torsional and Axial Loads
Authors: Itatani Masao , Tanaka Keisuke , Ohkawa Isao , Yamada Takehisa , Saito Toshiyuki
Published: January 2011

Fatigue tests of smooth and notched round bars of austenitic stainless steels SUS316NG and SUS316L were conducted under cyclic tension and cyclic torsion with and without static tension. Fatigue strength under fully reversed (R=−1) cyclic tension once increased with increasing stress concentration factor up to Kt=1.5, but it decreased from Kt=1.5 to 2.5. Fatigue life increased with increasing stress concentration under pure cyclic torsion, while it decreased with increasing stress concentration under cyclic torsion with static tension. From the measurement of fatigue crack initiation and propagation lives using electric potential drop method, it was found that the crack initiation life decreased with increasing stress concentration and the crack propagation life increased with increasing stress concentration under pure cyclic torsion. Under cyclic torsion with static tension, the crack initiation life also decreased with increasing stress concentration but the crack propagation life decreased or not changed with increasing stress concentration then the total fatigue life of sharper notched specimen decreased. It was also found that the fatigue life of smooth specimen under cyclic torsion with static tension was longer than that under pure cyclic torsion. This behavior could be explained based on the cyclic strain hardening under non-proportional loading and the difference in crack path with and without static tension.
Investigation of an Internal Arcing Event in an On Load Tap Changer Using Fluid Structure Interaction
Authors: Landis Ben , Brady Ryan , Petrovan-Boiarciuc Margareta , Pe´rigaud Guillaume
Published: January 2011

Electrical arcing can occur within the oil-filled tanks of transformers and On Load Tap Changers (OLTCs) when the oil dielectric properties are reduced. The electrical arc then, within milliseconds, vaporizes a portion of the oil, creating a highly pressurized gas bubble. The rapid pressure rise inside the generated bubble creates a dynamic pressure peak which then propagates and interacts with the tank structure. The wave reflections on the walls generate secondary pressure waves that raise the static pressure inside the tank. This static pressure increase leads to tank rupture since tanks are not designed as pressure vessels and therefore cannot withstand such levels of static pressure. This results in explosions, fire, expensive damages and possible environmental pollution. Despite all these risks, conventional protections are not effective to avoid explosion and contrarily to usual pressure vessels, no specific standard has been set to protect sealed tanks subjected to large dynamic overpressures. In order to study transformer and OLTC explosions and their prevention, experiments were performed on large scale transformers. Nevertheless, live tests require specific laboratory equipment and they are thus expensive and can be dangerous. In order to limit the costs, to reduce risks and to get a deep insight on the physical phenomena, numerical simulation tools are necessary. The development of such tools was presented at previous ASME Conferences. The present paper’s goal is to illustrate the use of this tool by an investigation of an arcing event in an OLTC. This arcing was simulated in the OLTC in both the unprotected case and also when the OLTC has been protected using a fast depressurization method. The stress and strain results are then used to determine how effectively the depressurization relieved pressures in the tank.
Investigation on the Failures Experienced in Deaerators and Solution Recommendation
Authors: Lim Aun Ming , Yuen Simon
Published: January 2011

The internals in the deaerators of a refinery plant were reported to have experienced a series of failures since their installation in 1985. These failures included development of cracks in the floor plates, damage of supports and breakage of fillet welds. Two possible root causes were initially identified; thermal stresses due to transient conditions and flow induced vibration. The former cause was classified as unlikely since the deaerators were always operating on steady-state conditions. No cyclic operating conditions were imposed on these deaerators. Vibrations however posed as the most likely root cause for the series of failures. The refinery plant inspectors reported that vibrations on the deaerators, although have not been measured, could be physically felt. These vibrations appear to be continuous and increase linearly with load. A finite element analysis was performed to determine the natural frequency of the deaerators. Mode shapes predicted from this calculation show that vibrations could have caused the failures of the internals. Furthermore, the lowest natural frequency of the deaerators appeared to fall within the actual vibration frequency on site (∼20 Hz). Although not confirmed, it is highly suspected that the vibration was excited by the flow (low pressure steam). Several repair options were explored to overcome this problem. These options were concentrated in increasing the stiffness of the steam inlet pipe and the deaerator floor. Finite element assessments demonstrated that the current flexible deaerator floor was the reason for the low natural frequency. An option of introducing reinforcement strips to the bottom side of the floor was identified as the best option to increase the natural frequency of the deaerator and this is expected to overcome the vibration problem. Only one vessel was assessed but the results apply to the other vessels since they are similar in design.
Sealing Performance Evaluation of Box-Shaped Flange Connections Under Thermal Conduction Conditions and Internal Pressure
Authors: Tenma Kentaro , Kurosawa Ryou , Sawa Toshiyuki
Published: January 2011

Bolted connections with gaskets such as flexible box-shaped flange connections have been used in mechanical structures. The connections are usually used under internal pressure as well as other loadings such as thermal, impact loadings and so on. In designing the bolted flexible box-shaped flange connections with gaskets, it is important to evaluate the sealing performance of the connections under internal pressure and thermal conduction conditions. In this paper, the contact gasket stress distributions and changes in bolt preload in the bolted flexible box-shaped flange connection with compressed sheet gaskets subjected to internal pressure and thermal conduction condition are analyzed using the finite element method (FEM). The leakage tests were conducted using an actual box-shaped flange connection with a compressed sheet gasket. Using the contact gasket stress distributions under internal pressure at an elevated temperature (Helium gas) obtained from the FEM calculations and an amount of the gas leakage measured in the experiment, the sealing performances are evaluated experimentally and numerically. The estimated amount of gas leakage is in a fairly good agreement with the measured results. In addition, the effect of the thermal conduction condition on the sealing performance is examined. Furthermore, the effects of the bolt pitch and the linear thermal expansion coefficients of the flanges and the gasket are examined at elevated temperature (20,80,100°C). Discussion is made on the effects of some factors on the sealing performance.
Core Design Experience to Adopt Long Term (18-Month) Cycle Operation Strategy From Initial Core
Authors: Jung Dong-Chul , Seong Ki Bong
Published: January 2011

SKN(Shin Kori Nuclear Power Plant) Unit 3, the first unit of APR1400 reactor in Republic of Korea, which is under construction, is scheduled to realize 18-month cycle operation strategy from its initial core. SKN Unit 3 is a pressurized water reactor with a thermal output of 4000 MWth having 3-batch 241 fuel assemblies. Its target cycle length will be realized as long term(about 455 EFPD) directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after 3rd ∼ 4th cycle. SKN Unit 3 has also been designed by meeting the following requirements as well as the above cycle length requirement; 1) peaking factor (Fxy) ≤ 1.60, 2) Negative MTC(Moderator Temperature Coefficient), 3) thermal margin ≥ 10%, 4) sufficient shutdown margin. 5) Xe stability, 6) CEA(Control Element Assembly) Re-arrangement, etc,. Higher enriched assemblies used to accomplish long term cycle at the initial core make higher radial and axial peaking factor than those of OPR1000. APR1400 introduces axial cutback region in a gadolinia fuel rod(6″ in each of top and bottom) and axial blanket region in a normal fuel rod(6″ in each of top and bottom) with low enriched uranium to enhance the neutron economy. It is also expected for these optimized cutback and blanket length to provide an advantage to reduce or avoid xenon oscillation phenomenon in power ascension test(PAT). In order to enhance core operability (RPCS requirement, Xe oscillation control and performance related design requirements, etc,.) as well as meeting shutdown margin requirement, CEA re-arrangement were also made. In this paper, APR1400 core design experience including all other design endeavors will be introduced.
Data Mining of General Corrosion of Most Commonly Used Iron-Based Alloys Using Kohonen Mapping
Authors: Kirsch Kathryn L. , Urquidi-Macdonald Mirna
Published: January 2011

This study was developed to provide an innovative methodology for designing steel alloys. Information on the general corrosion of a wide number of steel alloys in various electrolytes and environments was obtained from the National Institute of Standards and Technology (NIST). Parameters such as pH and conductivity used in each experiment (alloy in contact with an environment), along with alloy composition (from UNS number), and corrosion rates, were all collected in a single data row, or vector. To cluster by similarities, a web-based, publicly available Kohonen mapping software was used to perform the clustering analysis; Kohonen maps work by clustering together similar vectors and separating those vectors that differ. A vector was formed for each experiment for which corrosion rates were recorded; 1521 experiments were performed and each of those vectors was used to train the Kohonen Map. Once the Kohonen map is trained, each one of the cells forming the two-dimension Kohonen map will form clusters of vectors. Vectors containing similar information will be clustered together while dissimilar vectors will be clustered separately on the Kohonen map. The cells of the Kohonen map will adopt a “prototype” vector to be the representative of that cell; the prototype vector adopts the average values of all stored vectors in that cell. After the Kohonen map is trained, new vectors containing fabricated metal alloy composition (steels) and environment information can be input into the map. These new vectors, even though they do not contain corrosion rates, can be classified by the Kohonen map and entered into a cluster on the map. This methodology can be use to explore “if-then” scenarios of a new alloy in a different environment as well as obtain an expected corrosion rate of that particular alloy in that particular environment. Preliminary results of the trained Kohonen map are shown and discussed. The map results are used to explore the effects of the experiment environments and alloy composition on the general corrosion of the stainless steels.
The Effect of OBE Elimination on Class 1 Piping Systems
Authors: Nam Il Kwun , Hur Seok Hwan , Oh Dong Bum
Published: January 2011

The seismic analysis for the OPR1000 (previously entitled Korea Standard Nuclear Power Plant, now Optimized Power Reactor 1000 MW-Class) was carried out based on the SSE to assure the integrity of the reactor coolant pressure boundary and to secure a safe shutdown of reactor as well as the OBE to ensure the functionality of all the necessary systems for the continued operation of the reactor without undue risk to the health and safety of the public. In fact, the OBE with some supplementary functions, however, tends to control the design of nuclear power plants instead of the SSE due to its conservative analysis methodology. As a result of analysis work by means of more sophisticated techniques and continuously raised questions by the nuclear industry about the adequacy of using the OBE with at least one-half the maximum vibratory ground acceleration of the SSE as a design input, the OBE elimination design is exceptionally allowed through the alternative design rule by KINS (Korea Institute of Nuclear Safety: Independent nuclear regulatory expert organization), as long as the OBE is established as less than or equal to one-third of the SSE. Evaluated in this paper was the effect of the application of the OBE elimination-related alternative design rule on the KEPIC (Korea Electric Power Industry Code)/ASME Code Class 1 piping systems of an OPR1000-based nuclear power plant. The SSE (e.g. 0.3g) at least three-folds larger than the OBE (e.g. 0.1g) ground acceleration could decrease the flexibility of piping systems by adding more restraints to mitigate the increased SSE-induced seismic loads. Therefore, the flexibility of four piping subsystems, whose acquisition is one of the main purposes of the OBE elimination, was evaluated for an APR1400-based nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class), which has practically adopted the application of the OBE elimination for the first time in Korea.
Effect of Neutron Flux at High Fluence on Microstructural and Hardness Changes of RPV Steels
Authors: Dohi Kenji , Nishida Kenji , Nomoto Akiyoshi , Soneda Naoki , Matsuzawa Hiroshi , Tomimatsu Minoru
Published: January 2011

The effect of the neutron flux at high fluence on the microstructural and hardness changes of reactor pressure vessel (RPV) steels was investigated in succession to the previous study [1]. An accelerated test reactor irradiation of copper containing RPV materials, previously irradiated in commercial reactors, was carried out at the lowest possible neutron fluxes in order to obtain neutron fluences up to approximately 1×1020 n/cm2 (E>1MeV). State-of-the-art experimental techniques such as three-dimensional atom probe were applied to carry out advanced quantitative characterization of defect features in the materials. Results for the same materials irradiated in both high and low flux conditions are compared. For neutron fluences above 6×1019 n/cm2 (E>1MeV) the difference in the neutron fluence dependence of the increase in hardness is not seen for any neutron flux condition. The number densities and the diameters of solute atom clusters for the low flux irradiation materials tend to be lower and larger, respectively, than that for the high flux irradiation materials, while the volume fraction of solute atom clusters increases linearly with increasing neutron fluence, and the effect of neutron flux is not significant. The component elements and the chemical composition of the solute atom clusters formed by irradiation for the same material do not change regardless of the neutron fluence and flux. The square root of the volume fraction of the solute atom clusters provides a good correlation with the increase in hardness.
Comparison of PWSCC Growth Analysis With Field Data of Japanese PWR Plants
Authors: Asada Seiji , Suzuki Harutaka , Hirano Shinro , Sato Shoji , Matsumoto Kenji , Dozaki Kouji
Published: January 2011

The Subgroup on Fit-for-Service in the Japan Society of Mechanical Engineers (JSME) committee on Power Generation Facility Codes is developing fracture evaluation methodology on Alloy 600. A PWSCC (Primary Water Stress Corrosion Cracking) propagation diagram has been proposed by the Japan Nuclear Energy Safety Organization (JNES) NiSCC (SCC Growth on Ni-based Alloys) Project. The diagram has not only mean curve but also 2σ curve. PWSCC growth analyses on dissimilar metal welds of hot-leg nozzles have been performed by using the mean curve of PWSCC crack growth rate and the stress distribution by elastic-plastic analysis that was developed by the JNES IAF (Integrity Assessment of Flawed Components with Structural Discontinuity) Project. The PWSCC growth curves analyzed were compared with the field data that were measured for PWSCCs found in the dissimilar welds of nozzles in several Japanese PWR plants. The results showed the PWSCC growth curves using the mean crack growth rate curve could conservatively evaluate the field data.
Recent Activity of KEPIC Application for Construction and Operation of Nuclear Power Plants in Korea
Authors: Byun Hoonseok , Kim Jonghae , Son Youngho , Park Banuk
Published: January 2011

Korea Electric Power Industry Code (KEPIC), a set of integrated standards applicable to the design, construction and operation of electric power facilities including nuclear power plants, has been developed on the basis of referring to the prevailing U.S. codes and standards which had been applied to the electric power facilities in Korea. KEPIC has been applied to the construction of Ulchin Nuclear Units 5&6 since 1997 as per the endorsement of Ministry of Education, Science and Technology (MEST), and is being completely applied to the construction of Shin-Kori and Shin-Wolsung Nuclear Power Plants (NPPs) in Korea. Although the technical requirements of KEPIC related to design and inspections for pressure vessels have been developed on the basis of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Korea Electric Association (KEA) that is KEPIC maintenance organization is trying to find and improve unreasonable requirements based on the acquired experiences for the continuous construction and operation of NPPs. In the administrative requirements, KEPIC, unlike ASME, adopted not only mechanical field but also electrical field related to NPPs. Furthermore, the qualifying system for services such as Nondestructive Examination, Heat Treatment, Design and Equipment Qualification Organization was adopted to improve safety and reliability for NPPs. In the technical requirements, KEA has been performing betterment study such as pressures for system hydrostatic test, temperatures for Post-Weld Heat Treatment (PWHT) of P-No. 4 materials, counterbore for Pre-Service Inspection (PSI) of elbows in piping systems. The construction method of Steel Plate Concrete (SC) structure studied by Korea Hydro & Nuclear Power Co., Ltd. (KHNP), has been adopted to KEPIC, recently. In the case of inservice NPPs, KEA is trying to establish the equivalent material tables for ASME unassigned materials to solve problems of material during the Repair & Replacement step. KEA is performing development study for Standard Welding Procedure Specification (SWPS) which can be used for production weld without any specific procedure qualification test performed by manufacturers in order to improve reliability and reduce costs. The SWPS will be adopted in KEPIC-MQW Appendix. KEA hopes to collaborate and grow with other Standard Development Organizations (SDOs) through the above efforts.
Experimental Study on Critical Derivative of Cracking Magnetic Field Intensity in High Frequency Fatigue Test
Authors: Zhang Yi-liang , Gou Rui-bin , Li Ji-min , Shen Gong-tian , Wang Jing
Published: January 2011

This paper focuses on engineering application of metal magnetic memory (MMM) method. High-cycle fatigue test were applied to a large number of defect-free samples, and MMM signals were captured before fatigue fracturing. A concept of critical magnetic intensity (Hp) derivative (dHp/dx) of cracking was proposed. To study the characteristics of Hp and its derivative (dHp/dx) during cracking, fatigue test were applied to a total number of 50 pieces of Q235, 16MnR and base metal and welded specimens. In different fatigue periods, MMM signals were tested and microscopic metallographic were observed, so as to quantitatively analyze Hp and dHp/dx from both macro and micro angles. The results show that various materials differ in their critical dHp/dx limit, which could be the fatigue crack initiation time. It thus proves that MMM has special efficiency in the diagnosis of premature defects and stress concentration of metal structures.
Modularization Technology Development and Application for NPP in Korea
Authors: Lee Young Il , Lee Ung Kwon , Kim Tae Il
Published: January 2011

The construction period in a Nuclear Power Plant project is one of the most influential factors in determining the competitiveness of the project. In Korea, various methods have been applied to shorten the construction time of NPPs over the past 30 years. Recently, modularization technologies have been developed to shorten the duration dramatically. SC(Steel Plate Concrete) technology has been newly developed to modularize the building structure. SC structure is expected to reduce the construction period drastically by eliminating the formwork which takes lots of time during the construction of the conventional reinforced concrete structure. With the application of SC structure, the bulk commodities such as pipe, cable tray, duct and so forth can be modularized together with the SC structure. This type of module is referred to as ‘Combined Module’. Combined modules are selected on the basis of the density of bulk commodities inside the SC room modules. Another type of module, Equipment Module, also has been considered to facilitate the modularization of NPPs. A mock-up test is planned to evaluate the technical effectiveness of modularization before applying these modules to upcoming construction projects. With the continuous research for modularization and incremental application to upcoming projects, KHNP plans to revolutionize the relevant technologies and achieve a 36 month construction period from first concrete pouring to fuel loading in the future.
Effect of Ultrasonic Impact Treatment on Fatigue Life in Butt Welded Joints of Austenitic Stainless Steel
Authors: Togasaki Yu , Honda Takashi , Sasaki Tetsuya , Yamaguchi Atsushi , Tsuji Hirokazu
Published: January 2011

Ultrasonic impact treatment (UIT), which is a type of peening method, is usually used as a post-weld treatment for mild steel in order to improve the fatigue strength of its welded joints. As there is insufficient fatigue data available on welded joints of austenitic stainless steel treated by UIT, the authors decide to conduct fatigue tests on butt welded joints of austenitic stainless steel treated by UIT. The results were compared with the fatigue lives of as-weld joints to investigate the effect of UIT on the fatigue lives of welded joints of austenitic stainless steel. The fatigue lives of butt welded joints treated by UIT were more than 1.5 times longer than those of as-weld joints. The following were considered as possible reasons for this improvement in fatigue life: change in residual stress near the weld toes, relaxation of stress concentration at the weld toes, and refinement of grains under the weld toes. The residual stress measured near the weld toes by using the X-ray diffraction method was transformed from tension to compression by the application of UIT. The stress concentration factors at the welded toes were reduced about 10% by the application of UIT.
Development of APESS Software for Power Plant Simulation
Authors: Lee Kihyun , Jeong Kwanghun , Jeon Woowon , Bae Younghoon , Lee Dongsu
Published: January 2011

The description of Advanced Plant Engineering and Simulation System (APESS), its practical and potential uses, are discussed in this paper. The APESS is simulation software used in power plant operations developed by Doosan. Several other commercial simulation tools such as APROS, ProTRAX, JADE, etc. are presently used in the industry. These simulation tools are used for examining plant operation characteristics including start up and shutdown, control system design and analysis, plant behavior in emergency situations, evaluation of operating procedures, etc., making it possible to save commissioning time. For this, we developed the APESS based on our experiences as one of the major power plant suppliers in the world. The APESS treats the thermo-hydraulic problems with the well-known three fundamental conservation equation, i.e., the mass, energy and momentum conservation in principle. The process is considered as network composed of nodes and branches, so that a process component is modeled by the flow networks. The mass and energy exchange between systems and boundaries is assumed to take place only at nodes. On the other hand, a branch plays the role of a resistance element. The mass, momentum and energy conservation equations are formulated at each node, and then a set of resultant linear matrix algebraic is established. Some key aspects of the APESS are dealt in this paper. Up to now, we have been performing several dynamic behavior analyses of fossil-fueled power plant projects and modifying the APESS to improve accuracy of predicted dynamic behaviors. Thus making the APESS is reliable to estimate the dynamic behavior of the real fossil-fueled power plant and capable to predict the operation processes. In the future, we will apply the APESS to analyze dynamic behavior of nuclear power plants.
Seismic Design Margin of the Piping and Support System: Part 2—Proposal for Non-Linear Modeling of the Support
Authors: Shirai Eiji , Yamada Takanori , Ikeda Kazutoyo , Yoshii Toshiaki , Kondo Masami , Shimizu Hiroshi , Tai Koichi
Published: January 2011

On the evaluation of seismic design margin of the piping and support system in the nuclear power plant, the elasto-plastic behavior of support has played the important role in the seismic response analysis. It is desirable to establish the simple and effective non-linear modeling of the support in order to conduct the seismic response analysis considering each elasto-plastic characteristics of actual numerous supports in the piping system. In this study, the simplified modeling methodology related to the elasto-plastic behavior of the support was investigated by classifying the several non-linear types by estimating the limit load capacity of support, in which the results of non-linear type classification and accompanied failure mode were compared to the static loading tests of the support. At the results, methodology to identify the non-linear type of the support with normal bi-linear model was proposed, and it was confirmed that the non-linearity of the support was classified into the five different types.
Seismic Analysis Study of the HP Turbine Steam Supply Piping for APR1400
Authors: Lee Joon Ho , Kim In Yeung , Bae Jae Hwan
Published: January 2011

The high pressure (HP) turbine steam supply piping for APR1400 (Korea’s advanced power reactor 1400 MW-class) is routed from the Containment Building to the Turbine building via the Main Steam Isolation Valve Room in the Auxiliary building and supported by multiple structures. As such, the seismic analysis utilizing the Independent Support Motion (ISM) is preferable to the Enveloped Response Spectrum (ERS) method which is overly conservative but more widely used in seismic evaluation of piping systems. The ISM method is a mathematically rigorous technique that utilizes the principle of modal analysis to calculate the piping responses from each structure or support group and combines them to obtain the final results. In spite of the more realistic and less conservative approach of the ISM method for the piping systems subjected to multiple support excitations, the application of ISM method on a regular basis has been limited in nuclear power plant piping design because guidelines from the nuclear industry and the regulator are conservative and unclear. In this study, several related studies are performed to evaluate the adequacy of utilizing the ISM method for seismic analysis of the HP turbine steam supply piping for APR1400. Topics studied include support grouping effects, combination effects of responses between support groups, damping effects between PVRC and 4% damping per Regulatory Guide 1.61, and closely spaced modes effects. The results show that the support grouping by each floor of buildings, SRSS combination of support groups, 4% damping with ±15% spectrum peak broadening and absolute double sum (ADS) of modal combination are technically sound and preferable.
The Application of Optimized Commissioning Schedule
Authors: Lee Yong Ro
Published: January 2011

The targeted 24.5 months of commissioning time for the ongoing SKN unit #1 project in Korea is a tough challenge in view of saving at least 7 months from the previous 32 months for UCN unit #3 project that has been a prior version of OPR1000 and also the reference plant of this unit. Such commissioning time-saving has been made possible by introducing renovated technology, optimization of commissioning technique, and adoption of lessons-learned from the previous projects. Our experiences are discussed herein with respect to the scheduling of commissioning operation of previous NPP projects on a milestone basis, including trouble areas that could cause schedule delay and our approaches adopted for solving the key issues. In addition, this paper will introduce a summarized 24 month commissioning schedule which has been optimized to the most efficient level by reflecting the prior project experience, and applying the new techniques and methods.
Investigation on Evaluation Method Based on J Integral for Retardation of Crack Growth Due to Excessive Loading Beyond Small Scale Yielding Condition
Authors: Yamaguchi Yoshihito , Katsuyama Jinya , Onizawa Kunio , Sugino Hideharu , Li Yinsheng
Published: January 2011

Niigata-ken Chuetsu-Oki earthquake occurred in July 2007, whose magnitude was beyond the assumed one provided in seismic design. Therefore it becomes an important issue to evaluate the effect of excessive loading, in particular, for the components with existing crack. Fatigue crack growth rate is usually expressed by Paris’s law using the range of stress intensity factor (ΔK). However, applicability of the model to loading conditions beyond the small scale yielding remains as an issue since ΔK is inappropriate in such a high loading level. In this study, the fatigue crack growth behaviors after applying the excessive loads were investigated using austenitic stainless steel and carbon steel. Instead of ΔK, J-integral value for crack growth evaluation due to cyclic loading has been applied based on the experimental data to treat the excessive loading beyond small scale yielding. The finite element method (FEM) analyses were conducted to evaluate the stress distribution and plastic zone size for the excessive loading condition. The modified Wheeler model using J-integral range, ΔJ, has been proposed for the prediction of retardation effect on crack growth after excessive loading. It was indicated that retardation effect by excessive loading beyond small-scale yielding could be quantitatively evaluated using the J-Wheeler model.
Study on Ultrasonic Beam Behavior in Nickel-Based Austenitic Weldments
Authors: Ye Jing , Kim Hak-Joon , Song Sung-Jin , Kang Sung-Sik , Kim Kyungcho , Chung Hae-Dong , Song Myung Ho
Published: January 2011

Nickel-based austenitic weldments are widely used in steam generator, however it’s difficult for inspection because of its inhomogeneous and anisotropic elastic behaviors. The anisotropic and heterogeneous characteristics of austenitic welds together with a coarse-grained structure (elongated and oriented grains) lead to the following phenomena: scattering, attenuation, skewing, splitting, and divergence of the ultrasonic beam. To improve the ultrasonic inspection, beam behavior in weld need to be understood theoretically and experimentally. For that purpose, water immersed C-scans have been performed with respect to different parts (base material, austenite weld) of the weldments using one 5MHz transducer as transmitter at one side of the specimen and one 5MHz pinducer on the other side as receiver. Beam skewing and splitting is observed, 2D FFT are performed to obtain scan image in frequency domain and echodynamic curve is extracted from the C-scan data. The experiments results are also compared with the simulation results using multi-Gaussian beam model.
Effects of Overload on Fatigue Strength of SUS316 Having a Crack-Like Surface Defect
Authors: Oosedo Hiroko , Takahashi Koji , Ando Kotoji
Published: January 2011

The effects of overload on the fatigue strength and threshold stress intensity factor range (ΔKth) in SUS316 were studied. Tensile overload was applied to compact tension (CT) specimens with a large crack and fatigue tests were carried out to determine the ΔKth. Tensile or compressive overload was applied to bending fatigue test specimens with a small crack-like surface defect and fatigue tests were carried out to determine the fatigue limit and ΔKth. It was found that the ΔKth increased by tensile overloading. The increasing rate of ΔKth in the CT specimen is larger than that in the bending fatigue test specimen. Thus, the crack size effects on the improvement of ΔKth after overloading were observed. The results are discussed from the viewpoint of fracture mechanics. The size of compressive residual stress is the key factor of the increasing rate.
Formation Mechanism and Micro-Structure of Eigen-Line in Electrofusion Joints of Polyethylene Pipes
Authors: Shi Jianfeng , Zheng Jinyang , Guo Weican
Published: January 2011

Cold welding is the most dangerous defect in the electro-fusion (EF) joint of polyethylene (PE) pipes. A proprietary method was developed to detect the degree of cold welding by using an Eigen-line which was discovered in our previous study. To understand when, where and how the Eigen-line occurs, the forming mechanism was investigated in this paper. Three factors, i.e., micro-air-bubbles, difference of acoustic impendence of PE in the melted and un-melted region and small-crowded crystals may cause the appearance of the Eigen-line. It was found that the number of Eigen-lines was the same as that of welding times when re-welded with decreasing welding time, and only one Eigen-line could be observed when re-welded with increasing welding time. The result showed that small-crowded crystals may be the dominating factor. This was then verified by the temperature analysis and differential scanning calorimeter (DSC) tests, and the forming process of the small-crowded crystals was discussed in detail.
Development of Non-Destructive Test and Safety Assessment of Electrofusion Joints for Connecting Polyethylene Pipes
Authors: Zheng Jinyang , Shi Jianfeng , Guo Weican
Published: January 2011

Polyethylene (PE) pipes are widely used in transporting natural gas since 1970s. However, there is still no method for safety assessment of PE pipes. This paper introduces our recent works in developing ultrasonic NDT equipment and research on safety assessment of electrofusion (EF) joints of PE pipes. According to geometrical characteristic and ultrasonic response, the defects in EF joints can be divided into four categories, i.e. lack of fusion, voids, wire dislocation, and cold welding. Cold welding, which is the most common and dangerous defects, was successfully detected by the developed ultrasonic equipment and assessed with a patented cold welding Eigen-line methods.
Development of Advanced Fatigue Evaluation Methodology for Monitoring Major Components in Nuclear Power Plant
Authors: Kim W. , Kwon Jongjooh , Kang Hong Tae , Koo Gyeong-Hoi , Kim Tae-Ryong
Published: January 2011

In an attempt to develop fatigue monitoring system, two improved fatigue evaluation schemes have been proposed to monitor fatigue degradation in major components and piping of the pressurized water reactor. Proposed methods are both aimed to obtain realistic fatigue usage factors for given plant transients. Developed schemes utilize plant operating signals such as coolant temperature, pressure and flow rate. Finite element method and an improved Green’s function approach were used to calculate stresses and fatigue usage. Case studies were performed to validate effectiveness of each proposed scheme. It has been confirmed that proposed schemes can effectively reduce excessive conservatism in estimating fatigue usage and improve accuracy in stress calculation.
NSSS Design Features of Advanced Power Reactor Plus (APR+)
Authors: Choi Hae Yun , Lee Kwang Won , Seo Jong Tae
Published: January 2011

A feasibility study for developing Advanced Power Reactor Plus (APR+), an improved nuclear power unit to succeed to the Korea’s current Advanced Power Reactor 1400 MWe (APR1400), has been carried out for 2 years from August 2007 to July 2009. The major goals of this study are to identify top-tier requirements for the candidate APR+, to develop a preliminary design concept, and to evaluate conceptual design in terms of safety, economics, and performance characteristics. From this study, it is presumed that the APR+ can be developed as a two loop evolutionary pressurized water reactor with a number of advanced design features to enhance safety and economics based on the APR1400 technology. For economic enhancement, the APR+ core power has increased up to 4,290 MWth which corresponds to a 1500MWe class nuclear power plant. Several new construction technologies are introduced so as to shorten construction period. For safety enhancement, several advanced design features have been proposed in APR+ design such as an improved direct vessel injection (DVI+), an advanced fluidic device (FD+), a passive auxiliary feedwater system (PAFS), and a mechanical and electrical four train safety concept based on N+2 design philosophy. As accident mitigation features are improved, the in-vessel retention through external reactor vessel cooling (IVR-ERVC) system will be incorporated. The standard design of APR+ is in progress and expected to acquire the standard design approval from the Korean nuclear regulatory body by the end of 2012.
Seismic Responses of a Building Isolated With Multiple Friction Pendulum System Subjected to Multi-Directional Excitations
Authors: Tsai C. S. , Lin Yung-Chang , Su H. C.
Published: January 2011

In order to prevent a building from earthquake damage, a base isolation system called the multiple friction pendulum system (MFPS) which has numerous concave sliding interfaces is proposed to isolate a building from its foundation. Mathematical formulations have been derived to simulate the characteristic of the MFPS isolation system subjected to multi-directional excitations. By virtue of the derived mathematical formulations, the phenomena of the sliding motions of the MFPS isolator with several concave sliding interfaces under multi-directional earthquakes can be clearly understood. Also, numerical analyses of a building isolated with the MFPS isolator with several sliding interfaces have been conducted in this study to evaluate the efficiency of the proposed system in seismic mitigation. It has been proved through numerical analyses that structural responses have been reduced significantly and that the proposed system is a good tool to insure the safety of structures during earthquakes.
Numerical Analysis of the Heat Transfer in a Pressurizer Water Reactor Core
Authors: Jo Soo W. , Lee Yong K. , Jo Jong C.
Published: January 2011

For the initiation of emergency operating procedures and severe accident management of most commercial nuclear reactors worldwide, monitoring of the core temperature is required. Currently, it is not practicable to directly measure the temperature of fuel cladding surface temperature due to some technical limitations. Thus, measurement of the coolant temperature by using thermocouples at the core exit locations is widely used. However, the core exit temperature (CET) may not represent the core temperature properly because the measurement locations are somewhat distant from the heat generating part of fuel rod assembly. In this regard, it is important to assess the difference between the fuel cladding temperature and the CET. The objective of this study is to get the general insight and understanding of the boiling-induced multiphase flow inside fuel rod bundle during an abnormal operation mode following a loss-of coolant accident by comparing the calculation results of the CET deviation from the fuel cladding (or in-core) temperature for the two different cases of the present analysis model subjected to a coolant flowrate of either 100% or 50% of the nominal value. To do this, three-dimensional multi-phase computational fluid dynamics (CFD) calculations of a simplified pressurizer water reactor (PWR) core model were performed for both reactor operating modes. As a result, it was found that the calculated CETs are much lower than the maximum fuel rod cladding temperatures during both operating modes. Consequently, it is considered that the temperature deviation should be taken into account carefully to use the measured CETs for the initiation of emergency operating procedures and severe accident management of commercial nuclear reactors.
Nonlinear Analyses of a Building Isolated With Multiple Trench Friction Pendulum System Under Multi-Directional Excitations
Authors: Tsai C. S. , Lin Yung-Chang , Su H. C. , Tseng Ya-Fang
Published: January 2011

The multiple trench friction pendulum system (MTFPS) which has numerous sliding interfaces in each of two perpendicular directions can provide different natural periods, displacement capacities and damping effects at various levels of displacements in the individual direction. In order to characterize the mechanical behavior of the MTFPS isolator under multi-directional earthquakes, a plasticity model named as the multiple yield and bounding surfaces model is proposed in this study. Investigations for buildings isolated with TFPS isolators and subjected to multi-directional excitations have been carried out to assess the performance of the MTFPS isolation bearing in seismic mitigation through numerical analyses. Results from numerical analyses illustrate that the MTFPS isolation bearing can isolate most earthquake induced energy and provide good protection for structures from earthquake damage.
A Numerical Study on Mock-Up Downsizing to Predict Residual Stresses at CRDM Welds
Authors: Kim Ik-Joong , Jeong Jae-Uk , Choi Jae-Boong , Kim Young-Jin , Kim Sung-Woo , Kim Hong-Pyo , Chang Yoon-Suk
Published: January 2011

Since several defects caused by PWSCC (Primary Water Stress Corrosion Cracking) have been observed at the piping system of nuclear power plants, during the last two decades, lots of analytical and experimental researches have been performed to find out the relationship between the residual stress and PWSCC initiations. The present research deals with prediction of the residual stress at welding parts of CRDM (Control Rod Drive Mechanism). Especially, numerical investigations are carried out to support relevant experimental set-up because it is not easy to prepare large-sized mock-ups comparable to real geometry. First, preliminary examination was performed to establish an optimized welding process simulation procedure to accurately predict weld residual stresses. Then, detailed parametric FE analyses were carried out to examine effects of varying geometries and experimental conditions. Key findings were obtained from the FE analyses, which were used for finalizing the configuration of mock-up, are fully discussed in this manuscript.