bluuebunny/crossref_metadata_2025 huggingface dataset:
The IMS Development Plan in Construction Phase for Nuclear Power Plants
Authors: Chung Kyung Hwa , Han Jae Hee
Published: January 2011

The implementation of advanced information technology for nuclear power plant construction industry in the Republic of Korea is a challenging task to improve the competitive edge for APR1400 across the global nuclear business market. This paper will briefly describe the phases of development of an information management system (hereinafter IMS) in nuclear power construction during the last 40 years in Korea. Additional development task which strengthen the IMS capability will be described from our experience. It will analyze the IMS development and implementation stages for Korean nuclear power plant construction projects such as YGN 3&4, UCN 3&4, YGN 5&6 and UCN 5&6, and the successful application case of the UCN 5&6 Radwaste Building in which 3D CAD technology was implemented for the first time. By reviewing lessons learned, this paper will define the Information Technology advancements resulting the reduction of project costs and construction schedule both by project execution procedures and IT systems including 3D CAD application. The future plan will include integrating project management systems based on data-centric approach and handover strategy for better O&M phase through configuration management technology. In this report, we address the functions to be developed and added in the new IMS.
The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals
Authors: Park K.-W. , Bae J.-H. , Park S.-H.
Published: January 2011

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.
APR1000: Further Improvements of the OPR1000 Design
Authors: Kim Insik , Chi Sung-Goo , Yoo Keun-Bae
Published: January 2011

Korean nuclear industry has demonstrated a success of KSNP (Korean Standard Nuclear Power Plant) program through operations of its six units starting from Ulchin 3&4 since 1998, and four more units are currently under construction with its new name of OPR1000 (Optimized Pressurized Reactor 1000). During the past 20 years of OPR1000 constructions and operations, Korean nuclear industry has put continuous effort to secure its better confidence in safety and performance. As a result, OPR1000 is showing the highest capacity factor and the lowest unplanned trip records among the worldwide operating nuclear power plants. The goal of APR1000 (Advanced Power Reactor 1000) program launched recently is to develop an improved version of OPR1000 design. It is required to adopt selected advanced design features of APR1400 (Advanced Power Reactor 1400) and to fulfill the new and strengthened design requirements including longer plant lifetime, colder head reactor, longer refueling interval, enhanced cyber security, etc. It is also essential to consider regional specific conditions such as electric frequency, sea water temperature and seismic conditions, etc. In this paper, the overall APR1000 development program is introduced, and design requirements as well as advanced design features (ADFs) for APR1000 are discussed comparing with those for OPR1000.
A Study of Elasto-Plastic Response of Single Degree of Freedom System Using Artificial Ground Motions With Given Time-Frequency Characteristics
Authors: Ichihashi Ichiro , Sone Akira , Masuda Arata , Iba Daisuke
Published: January 2011

In this paper, a number of artificial earthquake ground motions compatible with time-frequency characteristics of recorded actual earthquake ground motion as well as the given target response spectrum are generated using wavelet transform. The maximum non-dimensional displacement of elasto-plastic structures excited these artificial earthquake ground motions are calculated numerically. Displacement response, velocity response and cumulative input energy are shown in the case of the ground motion which cause larger displacement response. Under the given design response spectrum, a selection manner of generated artificial earthquake ground motion which causes lager maximum displacement response of elasto-plastic structure are suggested.
Local Structure Response of High-Rise Empty Cylindrical Shell Towers to Wind-Induced Resonance Vibration
Authors: Pan Jianhua , Chen Xuedong , Ai Zhibin , Jiang Huifeng
Published: January 2011

High-rise towers are very sensitive to wind loads because of their high flexibility. Wind-induced resonance vibrations due to Karman vortex street have great impact on these structures, especially for the high empty cylindrical shell towers. In a petrochemical enterprise, a few local cracks were detected in some empty towers and the wind-induced resonance vibrations were considered to be responsible for that. In this paper, by employing software SW6 and ANSYS, the tower cracking cause was analyzed with emphasis on the simulation of local structure response for an empty tower under wind loads. It can be found that the stress level was quite severe at the head-skirt conjunctions which were consistent with those actual cracking positions.
The Function Test for the Reliability Confirmation of High Temperature and High Pressure Vessel for Irradiation
Authors: Park Kook-Nam , Lee Jong-Min , Ahn Sung-Ho , Wu Sang-Ik , Kim Young-Ki
Published: January 2011

The Fuel Test Loop (FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR (Pressurized Water Reactor) and CANDU (CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI (Korea Atomic Energy Research Institute). It consists of In-Pile Section (IPS) and Out-of Pile System (OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA (IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to maintain the functionality of the reactor coolant pressure boundary. The functional test and verification of the IVA were done through pressure drop, vibration, hydraulic and helium leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by a local technique and has finally been tested under high temperature and high pressure. The IVA and piping did not experience leakage, as the piping, flanges, and assemblies have been fully checked. Good data was obtained during the three cycle test which included a pressure test, pressure and temperature cycling, and constant temperature. A new concept of IVA, as easy attaching and separating test fuel at IVA will be designed thereafter.
STYLE: Mock-Up3 Design—FE Simulation of Crack Growth in a Cladded Ferritic Pipe
Authors: Nicak Tomas , Schendzielorz Herbert , Keim Elisabeth , Meier Gottfried , Moinereau Dominique , Le Delliou Patrick
Published: January 2011

This paper describes numerical analyses performed in connection with the design of a large scale mock-up test planned within the European project on structural integrity STYLE. There are three large scale mock-up tests planned in STYLE, each of them dedicated to investigate specific effects. Mock-up3 (cladded ferritic pipe with an outer diameter of 420 mm) is foreseen to investigate transferability of material data, including fracture mechanics properties. Usually, material data are obtained by testing small specimens, which are subsequently used for the assessment of large scale structures (real components). The Mock-Up3 is an original part of a surge line made of low alloy steel 20MnMoNi55 (similar to SA 508 Grade 3, Cl. 1). The test will be performed on a 4 point bending test facility provided by EDF under displacement control at room temperature. The goal of the test is to obtain the stable crack growth of an inner surface flaw until a break through the wall occurs. The range of assessment tools applied within STYLE includes assessment of component failure by fracture mechanics analyses using methods based on fracture mechanics parameters (e.g K1 or J-Integral) as well as methods based on local micromechanical models (e.g. Gurson’s porous plasticity model and its variations). Micromechanical models have some advantages compared to those based on single term fracture parameters, especially if one considers designing a large scale mock-up test. The precise description of the entire damage process, beginning from with initiation on brittle particles, their growth leading to the crack initiation and finally to the macroscopic crack growth, can be seen as the most valuable attribute of these methods. Such methods make it possible to perform a set virtual tests prior to the real one, on which different test conditions can be investigated. A comparison of the common assessment method based on J-Integral with a local approach method, based on the Gurson’s porous plasticity theory, will be presented in this paper. Details on the Gurson model calibration will also be provided. Moreover, influence of boundary conditions on the large scale test will be discussed.
Seismic Design Margin of the Piping and Support System: Part 3—Evaluation of Seismic Margin of the Piping System
Authors: Shirai Eiji , Yamada Takanori , Ikeda Kazutoyo , Yoshii Toshiaki , Kondo Masami , Tai Koichi , Ogo Takuya
Published: January 2011

Seismic safety is one of the major key issues of nuclear power plant safety in Japan. It is demonstrated that nuclear piping possesses large safety margins through the piping and support system test, which consisted of three dimensional piping, supports, U-bolts, and concrete anchorages, using the E-defense vibration table of National Research Institute for Earth Science and Disaster Prevention, Hyogo Earthquake Engineering Research Center, on extremely high seismic excitation level [1,2,3]. In the above test, the non-linear hysteretic behaviors of the support are quite complicated, but the dissipated energies introduce large damping effects on the piping system response. In order to evaluate the inelastic behavior of the support with respect to the whole piping system response, the following simulation methodology for the support re-evaluation is proposed. 1) Non-linear modeling of the support: • Failure mode and failure capacity of each support. • Simplified non-linear modeling of each support. 2) Simulation Analysis of the piping and support system: • Considering the non-linearity both of the supports and elbows in the piping system. 3) Evaluation of seismic margin: • Focused on the failure level for the support system, and the fatigue damage for the strain range of the piping. The limit state analysis of the typical piping system of a nuclear power plant is presented in this paper, and it is demonstrated that these evaluations of the seismic margins would give important insight into the support reinforcement program on the seismic re-evaluation work.
Mathematical Analysis on Two Modelling Techniques for Dynamic Responses of a Structure Subjected to a Ground Acceleration
Authors: Kim Yong-Woo , Jhung Myung Jo
Published: January 2011

We considered two kinds of numerical modeling techniques for the dynamic response of a structure subjected to ground acceleration. One of the techniques is based on the equation of motion relative to ground motion, and the other is based on the equation of absolute motion of the structure and the ground. The analytic background of the former is well established while the latter is not yet. The latter is called large mass technique, which allocates an appropriate large mass to the ground so that it can cause the ground to move according to a given acceleration time history. In this paper, employing a single degree-of-freedom spring-mass system, we analyzed the equations of motion of the two techniques and provided some theorems on the large mass technique. Using simple examples, we compared numerical results of two modeling techniques with analytic solutions. We have shown that the theorems give us a clear insight on the large mass technique through numerical tests.
Incorporation of Materials’ Sensitive Simulation Technique Into the PINTIN-CAM Code for Integrity of Nuclear Piping System
Authors: Datta Debashis , Jang Changheui
Published: January 2011

Previously, the development process of a combined probabilistic fracture mechanics code for nuclear piping system was proposed based on some pre-assigned (by default) crack initiation and crack growth techniques e.g. assumption of preexisting cracks in case of no initiated cracks resulted from either fatigue or stress corrosion crack mechanisms and individual growth of those cracks by fatigue or stress corrosion cracking phenomenon without considering crack coalescence criteria. In one sense this type of approach might produce conservative results. To reduce the undue conservatism, in this version, materials’ sensitive crack initiation and subsequent crack growth of these simulated cracks along with crack coalescence criteria have been given prior importance. The logic behind this revised version of PINTIN-CAM PFM code has been presented and one previously published NUREG benchmark numerical problem has been reinvestigated by the improved features of this code. Based on design condition stress data, a typical PWR RPV outlet nozzle section shows the circumferential and axial through wall crack combined failure probability in the order of about 10−2 and 10−3 respectively. Among individual locations of a RPV ON section, the Inconel part shows relatively highest susceptible failure location followed by SS and LAS locations.
Guidance of Equipment Risk Management With RBI Techniques
Authors: Qu Yu , Hao Xiangmin , Jiang Anhe , Gao Mingchao
Published: January 2011

The West Pacific Petrochemical Company Limited, Dalian is the first enterprise that adopts risk-based inspection (RBI) technology in the oil refinery plants in China, the application of RBI technology plays a positive role in guaranteeing long-cycle operation of the plant. With the continuous deepening of degree of understanding of API 581 standard and continuous enriching of experience with use of RBI software, by tracing the risks, there are three major factors that affect the risk of equipment or component, i.e. risk consequence, failure likelihood and management system, which are taken as the direction of efforts for guiding the implementation of equipment risk management measures. If there are litigation measures in the release consequence (sprinkler system or passive litigation technology) that can reduce continuous release rate and release time, then the release consequence of equipment can be reduced; if inspection techniques are adopted with respect to corrosion failure mechanisms or the proportion of inspection is increased in the failure likelihood, then the failure likelihood can be reduced effectively; and other measures such as analyzing the problems listed in each item of management system assessment and taking measures to improve management level and thus increase scores of management system by comparing the difference between the management of the enterprise itself and those listed in the standard. By implementing above measures, the ability of equipment risk management is improved and thus the intrinsic safety of equipment management is effectively guaranteed.
Guaranteeing Long-Cycle Safe Operation of Ethylene Plants by RBI Technology
Authors: Ai Zhibin , Chen Xuedong , Yang Tiecheng , Hu Jiushao , Cheng Chuanqing
Published: January 2011

Ethylene plant is the critical plant in petrochemical industry, the output of its major product - ethylene directly reflects the development situation of a country’s petrochemical industry, therefore, the long-cycle safe running of the ethylene plant is of important significance. At present, the operation cycle of ethylene plants in developed countries is generally 4∼6 years, whereas that in China is mostly 3 years, which seriously restricts the production capacity, cost and international competitive power of Chinese ethylene production enterprises. With a great number of new or expansion projects of 1,000,000 tons/year ethylene plants put into production successively in China, it is extremely urgent to guarantee long-cycle running of ethylene plants. Starting from 2003, we initially adopted risk-based inspection (RBI) technology in Chinese ethylene plants, and achieved the objective of extending the inspection interval of ethylene plants from the original 3 years to 6 years in combination with technical means like online inspection etc. Up to the present, we have completed RBI of 13 ethylene plants, and the running cycle of most of them has been extended. In this paper the application situation of RBI technology in ethylene plants is presented, by means of RBI, major factors that influence long-cycle safe running of ethylene plants are identified effectively, and feasible risk reduction measures have been developed, which provide reliable assurance for long-cycle safe running of the ethylene plants.
Parametric Study of Interaction Between Multiple Nonaligned Part-Through-Wall Cracks
Authors: Konosu Shinji , Kasahara Kyosuke , Ito Kenji , Kanamaru Shinichiro
Published: January 2011

Multiple nonaligned part-through-wall cracks are common flaws in many structural components. It is, therefore, necessary to be able to readily estimate the interaction of stress intensity factor in order to assess these cracks in failure assessment diagram and fatigue crack growth life prediction. However, because of the complicated nature of the foregoing, adequate equations are not available. In this paper, a parametric study of the interaction between two nonaligned adjacent cracks is undertaken by the body force method and three-dimensional finite element analyses, together with the analyses of the existing results available in the literature. The empirical expressions for the three-dimensional modification factor for nonaligned part-through-wall cracks subjected to uniform tension, which is derived with reference to the two-dimensional interaction factor, are proposed as a function for a wide variety of crack configurations.
Accuracy Comparison of the Electrical Potential Drop Technique and Ultrasonic Testing for the Pipe Wall Thinning in the Thermal Power Plant
Authors: Yoshino Shin , Hamada Seiichi , Kaieda Yoshi
Published: January 2011

The electrical potential drop technique is one of the promising methods for monitoring the pipe wall thinning. In the previous paper [1], it was reported that this technique had a good performance equal to the ultrasonic testing in the preliminary test and that it had a long-term durability under a severe condition on the real pipes in operating thermal power plants. This time one of these pipes was replaced because its thickness approached the threshold. The removed pipe was cut in many pieces and its thickness was measured with calipers at several locations corresponding to where the electrical potential drop technique was applied. This result was compared with the thickness resulted from this technique and the conventional ultrasonic testing that were conducted just before the pipe removal. This comparison led to the conclusion that the accuracy in the electrical potential drop technique was almost equivalent to that in the ultrasonic testing even in the real pipe under a severe condition. In this paper, the accuracy in the electrical potential drop technique is mainly discussed.
Active Control of Non-Linear System by a Fuzzy-Optimal Control Technique
Authors: Mochio Takashi
Published: January 2011

The purpose of this paper is to estimate the real time vibration control of an actively-controlled nonlinear structure due to non-stationary external loads. When the optimal control theory is adopted as a control law against the concerned task, the derivation of time dependent optimal control gains may be required because of a remarkable non-stationarity of response amplitude. In addition, since the system is nonlinear, it takes more time to calculate those time dependent gains. This means that it is difficult to strictly execute the real time active control with optimal control theory as for the non-stationary and nonlinear system. In this paper, therefore, one approximate technique, coupled fuzzy-optimal control, is proposed in order to realize the real time control of non-stationary and nonlinear system. Finally, results by deterministic analysis based on numerical simulations are compared with those by stochastic analysis using statistical equivalent linearization technique.
Proposals of Guidelines for High Temperature Structural Design of Fast Reactor Vessels
Authors: Kasahara Naoto , Satoh Kenichiro , Tsukimori Kazuyuki , Kawasaki Nobuchika
Published: January 2011

Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induce incremental deformation and creep fatigue crack at critical portions around the sodium surface, thermal stratification layer and core support structures. Those phenomena are so complex that design evaluation becomes sometimes too conservative. In order to achieve precise high temperature design for realizing compact reactor vessels of fast reactor plants, such guidelines are proposed as for thermal load modeling, structural analysis and strength evaluation. This paper gives the brief summary of these guidelines. GUIDELINES FOR THERMAL LOAD MODELING: One of main difficulties of thermal load modeling is their inducement mechanism by interaction between thermal hydraulic and structural mechanics. Design evaluation requires envelope load conditions with considering scatter of design parameters. Proposed guidelines enable precise load modeling by grasping sensitivities of thermal stress to design parameters including thermal hydraulic ones. GUIDELINES FOR INELASTIC DESIGN ANALYSIS: Guidelines are proposed to apply inelastic analysis methods for design of reactor vessels. There are so many influence parameters in inelastic analysis that conservative and unique solutions are hardly found. To overcome such difficulties, mechanism and main parameters of inelastic behaviors of reactor vessels were clarified. Guidelines give conservative results within the same mechanism as expected reactor vessels. HIGH TEMPERATURE STRENGTH EVALUATION METHOD: Incremental deformation and creep fatigue strength evaluation methods were proposed. Accumulated strain is limited within no influence of fatigue and creep-fatigue strength. Taking design conditions of reactor vessels into account, creep fatigue evaluation considers strain concentration and an intermediate stress hold effect on creep-fatigue strength. Influences of thermal aging were also confirmed.
Methods for Life Assessment of Hypercompressor Poppet Valves
Authors: Bagagli Riccardo , Campo Nicola , Francini Stefano , Simpson Karl C. , Morgan Christopher
Published: January 2011

Valves are extremely important for the availability of hypercompressors in LDPE plants. A methodology for the life assessment of these valves was developed based on numerical calculations and experimental results. A test rig was used to measure the behavior of different poppet valves in terms of differential pressure and valve lift as a function of the process flow. Water is used as the working fluid instead of ethylene, since the flow is almost incompressible. The Measured data were reduced to non-dimensional flow and drag coefficients as a non-linear function of the poppet position. The valve characteristics and real gas data were used in a one-dimensional dynamic simulator to evaluate the poppet position during a single crank cycle together with pressure losses. The calculated dynamic behavior of the poppet was correlated to the failure modes experienced in the field. A dynamic finite element analysis was used to calculate the mechanical stresses on both the poppet and valve body from impact velocities. The stresses from each impact (both during the opening and closing phases) and the axial poppet displacement were accounted for to estimate the total cumulative damage and assess the life of each valve design.
Numerical Design of Tubesheet Structure Based on Finite Element Analysis
Authors: Yu Xiaodong , Qian Caifu , Yu Hongjie
Published: January 2011

In this paper, finite element analysis models were established to simulate a tubesheet connected with tubes by expanding or welding method. Stresses and deformations in the tubesheet produced by pressure load were calculated with emphasis on the effect of the tubes support. Results showed that the deflections and stresses at the un-perforated zone are greatly affected by the radius of the perforated circle for a partly perforated tubesheet, but they are not sensitive to the radius of the tubesheet when it is fully supported by tubes. If applying numerical design for heat exchangers based on finite element analysis, the tubesheet thickness could be significantly reduced. Compared with the welding connection, expanding connection of the tubes and tubesheet is more helpful to raise the strength and stiffness of the perforated tubesheet.
Relationship Between Singularity Strength Factors and Practical Bonding Strength of Ceramic to Metal Joint
Authors: Tateno Masayoshi , Morikawa Hiroki
Published: January 2011

This study correlates the stress intensity factors (Kii, Kpaii), the order of the stress singularity p-1, and the practical bonding strength of ceramic to metal joints with arbitrary interface geometry. First, in order to describe the stress behavior of TiB2 to Ni joints bonded at 1173K, the stress singularity factors (Kii, Kpaii, p-1) were theoretically derived for wedge angles of φ1:30°<φ1<150°. Secondly, the dependences of the singularity factors on the wedge angle φ1 were compared with experimental results using the same wedge angles on TiB2 to Ni joints with interface convexity or concavity produced by Electric Discharge Machining (EDM). In order to obtain the highest strength joints, the optimum interface shape is determined using the relationship between the singularity factors.
Further Welding Residual Stress and Flaw Tolerance Assessment of Dissimilar Metal Welds With Alloy 52 Inlays
Authors: Rudland D. , Brust F. , Shim D. J. , Zhang T.
Published: January 2011

Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal (DM) welds (specifically Alloy 82/182 welds) in pressurized water reactors (PWRs) can cause a safety concern due to the high crack growth rate and irregular shaped flaws. Since many of these welds reside in primary piping systems that have been approved for Leak-Before-Break (LBB), the domestic commercial nuclear power industry has proposed a number of mitigation strategies for dealing with the issue and assuring LBB is still applicable. Some of these methods include Mechanical Stress Improvement Process (MSIP), Full and Optimized Structural Weld Overlay (FSWOL, OWOL), and Inlay and Onlay cladding. The industry claims that these methods provide either a reduction in the inner diameter residual stress field (MSIP and WOL), and/or apply a non-susceptible corrosion resistant barrier to stop or retard PWSCC crack growth to form a through-wall leak path (WOL, Inlay, Onlay). At last years PVP conference, a companion paper was published that described the initial welding residual stress and flaw evaluation analyses to investigate the effectiveness of inlay welds as a mitigative technique. The results from that effort suggested that the time to leakage with an inlayed weld is highly affected by the depth of the inlay and the crack growth rate within the inlay. In this ongoing effort, further welding residual stress analyses are presented that investigate the effects of the inlay depth and a variety of weld repair options before the standard 3mm deep inlay. In addition, further crack growth analyses, assuming idealized crack shapes, were conducted to investigate the effects of weld residual stress, crack growth rate, global bending stress, and flaw size and orientation. The results of these analyses aid in determining appropriate inspection intervals for dissimilar metal welds with this mitigation technique.
Effects of Delta - Ferrite Content on the Integrity of Reactor Vessel Cladding
Authors: Shin Ho-Sang , Hong Jin-Ki , Chung Koo-Kab , Chung Hae-Dong , Kim Gwang-Yil , Cho Sung-Woo
Published: January 2011

As the design life of new nuclear power plant increases, the austenitic stainless cladding integrity of reactor vessel becomes one of the new concerns. Since 1970’s, there have been some specific recommendations on delta ferrite content of austenitic cladding of reactor vessels and welds. It has been known that the delta ferrite is beneficial for reducing micro-fissure in welds, though the high delta ferrite content increases the probability of embrittlment of welds. In this study, the mechanical and microstructural properties of austenitic weld metals with the limit values of the recommended range (5 ∼ 18 FN) of the delta ferrite control on low alloy steels were characterized by using bending test and scanning electron microscopy. The base metal was ASME Code Sec. II specification SA 508 Gr. 3 Cl. 1 plate and weld materials were EQ308L and EQ309L strips. Four kinds of cladding were deposited with submerged arc welding process on SA508 cl.3 plates. The bending tests were performed through ASME code Sec. IX and the microstructure of fractured surfaces was analyzed by scanning electron microscopy (SEM). In bending tests, there were no fractures except the highest delta ferrite content specimens (28FN). From the SEM observation of fractured surfaces, cracks initiated from the interface between austenite and ferrites phases in the cladding layer and propagated through the continuous interfaces between two phases. For specimens without continuous interfaces of two phases, though the cracks were observed in the interface of phases, the propagation of cracks was not observed. From the test results, continuous interfaces between austenite matrix and ferrite phase provide the path for crack propagation. And the delta ferrite content affects the integrity of cladding of reactor vessel.
Mechanical Fatigue of Wall Surface Caused by Liquid Droplet Impingement
Authors: Satou Manabu
Published: January 2011

Pipe wall thinning caused by water or steam flow was observed associated with oxide layer inside of the pipe. Interaction between oxide formation and corrosion or erosion due to the water or steam flow may be an essential phenomenon of the wall thinning. Thinning rate of the wall therefore depends on the formation of the oxide. In the case of wall thinning caused by liquid droplet impingement (LDI) erosion models, mechanical fatigue of the layers is of interest from estimation of the wall thinning rate. In this paper, from a fundamental point of view, to examine parameters related to adhesion strength of the interface in the model equation of material removal from the wall by multiple droplet impingements, evaluation of adhesion strength between piping material and surface oxide layer was carried out using a laser shock method. Several model oxide layers were prepared at elevated temperatures in oxidizing environments on a carbon steel. Results from the measurements of the adhesive strength of the oxide layer formed on the carbon steel at elevated temperatures, the interface had a comparative strength or less of the yield stress of the carbon steel. It was found that reputation of the loading by laser shots up to 104 times did not affect the adhesive strength so far. A kinetic modeling of the wall thinning caused by the LDI was suggested higher cycle mechanical fatigue.
Authorizing the DOT Specification 6M Packaging for Continued Use at the Savannah River Site
Authors: Watkins Robert W. , Loftin Bradley M. , Hoang Don S.
Published: January 2011

The U.S. Department of Transportation (DOT) Specification 6M packaging was in extensive use for more than 40 years for in-commerce shipments of Type B quantities of fissile and radioactive material (RAM) across the USA, among the Department of Energy (DOE) laboratories, and between facilities in the DOE production complex. In January 2004, the DOT Research and Special Programs Administration (RSPA) Agency issued a final rule in the Federal Register to amend requirements in the Hazardous Materials Regulations (HMR) pertaining to the transportation of radioactive materials. The final rule became effective on October 1, 2004. One of those changes discontinued the use of the DOT specification 6M, along with other DOT specification packagings, on October 1, 2008. A main driver for the change was due to the fact that 6M specification packagings were not supported by a Safety Analysis Report for Packagings (SARP) that was compliant with Title 10 of the Code of Federal Regulations (CFR) Part 71 (10 CFR 71). The regulatory rules for the discontinued use have been edited in Title 49 of the Code of Federal Regulations (CFR) Parts 100 – 185, 2004 Edition and thereafter. Prior to October 1, 2008, the use of the 6M within the boundaries of the Savannah River Site (SRS), called an onsite transfer, was governed by an onsite transportation document that referenced 49 CFR Parts 100 – 185. SRS had to develop an Onsite Safety Assessment (OSA) which was independent of 49 CFR in order to justify the continued use of the DOT Specification 6M for the transfer of radioactive material (RAM) at the SRS after October 1, 2008. This paper will discuss the methodology for and difficulties associated with authorizing the DOT Specification 6M Packaging for continued use at the Savannah River Site.
Packaging and Transportation of Additional Neptunium Oxide
Authors: Watkins Robert W. , Hensel Stephen J. , Jordan Jeffrey M.
Published: January 2011

The Savannah River Site’s HB-Line Facility completed a second neptunium oxide production campaign in which nine (9) additional cans of neptunium oxide were produced and shipped to the Idaho National Laboratory and Oak Ridge National Laboratory in the 9975 shipping container. These additional cans were from a different feed solution than the first fifty (50) cans of neptunium oxide that were previously produced and shipped via a Letter of Amendment to the 9975 Safety Analysis Report for Packaging (SARP) content table. This paper will address the challenges associated with demonstrating the neptunium oxide produced from the additional feed solution was equivalent to the original neptunium oxide and within the content description of the Letter of Amendment.
Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1
Authors: Yin Shengjun , Dickson Terry L. , Williams Paul T. , Bass B. Richard
Published: January 2011

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.
Hydrodynamic Forces, Pressure and Mass Flux in Two-Phase Air-Water Flow Through Transparent Safety Valve Model
Authors: Kourakos Vasilios , Chabane Sai¨d , Rambaud Patrick , Buchlin Jean-Marie
Published: January 2011

Safety devices are of major importance in the nuclear and chemical industry. Examples of such systems are safety valves and rupture disks. The behavior of this type of valve in single-phase flow (gas or liquid) has been described in detail in the literature, while for two-phase flow there is a lack of relative models since the phenomena are much more complex. The design and sizing of this apparatus is an important issue which would prevent its wrong functionality that could cause a hazardous situation. In this paper, a transparent (made in Polymethyl methacrylate) model of a safety valve is studied (1 1/2 in G 3 in); this has allowed full optical access and therefore the identification of the structure of the flow and the observation of the different phenomena occurring. Instead of a spring, used in an actual safety valve, the disk is fixed and its position can vary from completely closed to fully opened position. Thus, the static flow behavior of the valve is examined.
Historical Rail Accident Analyses Identifying Accident Parameters That Could Impact Transportation of Spent Nuclear Fuel
Authors: Mintz Todd S. , Adams George , Necsoiu Marius , Mancillas James , Bajwa Chris , Easton Earl
Published: January 2011

As the regulatory authority for transportation of spent nuclear fuel (SNF) in the United States, the Nuclear Regulatory Commission (NRC) requires that SNF transportation packages be designed to endure a fully engulfing fire with an average temperature of 800 °C (1,475 °F) for 30 minutes, as prescribed in Title 10 of the Code of Federal Regulations (CFR) Part 71. The work described in this paper was performed to support NRC in determining the types of accident parameters that could produce a severe fire with the potential to fully engulf a SNF transportation package. This paper describes the process that was used to characterize the important features of rail accidents that would potentially lead to a spent nuclear fuel transport package being involved in a severe fire. Historical rail accidents involving hazardous material and long duration fires in the United States have been analyzed using data from the Federal Railroad Administration (FRA) and the Pipeline and Hazardous Materials Safety Administration (PHMSA). Parameters that were evaluated from this data include, but were not limited to, class of track where the accident occurred, class of hazardous material that was being transported, and number of railcars involved in the fire. The data analysis revealed that in the past 34 years of rail transport, roughly 1,800 accidents have led to the release of hazardous materials resulting in a frequency of roughly 1 accident per 10 million freight train miles. In the last 12 years, there have only been 20 accidents involving multiple car hazardous material releases that led to a fire. This results in an accident rate of 0.003 accidents per million freight train miles that involved multiple car releases and a fire. In all the accidents analyzed, only one involved a railcar carrying Class 7 (i.e., radioactive) hazardous material (HAZMAT).
Drop Tests Results of Revised Closure Bolt Configuration of the Standard Waste Box, Standard Large Box-2, and Ten Drum Overpack Packagings
Authors: May Cecil G. , McKeel Charles A. , Opperman Erich K.
Published: January 2011

The Transuranic (TRU) Disposition Project at Savannah River Site will require numerous transfers of radioactive materials within the site boundaries for sorting and repackaging. The three DOT Type A shipping packagings planned for this work have numerous bolts for securing the lids to the body of the packagings. In an effort to reduce operator time to open and close the packages during onsite transfers, thus reducing personnel exposure and costs, an evaluation was performed to analyze the effects of reducing the number of bolts required to secure the lid to the packaging body. The evaluation showed the reduction to one-third of the original number of bolts had no affect on the packagings’ capability to sustain vibratory loads, shipping loads, internal pressure loads, and the loads resulting from a 4-ft drop. However, the loads caused by the 4-ft drop are difficult to estimate and the study recommended each of the packages be dropped to show the actual effects on the package closure. Even with reduced bolting, the packagings were still required to meet the 49 CFR 178.350 performance criteria for Type A packaging. This paper will discuss the effects and results of the drop testing of the three packagings.
Simulation of Ductile Fracture for Multiple Flaws
Authors: Kikuchi Masanori , Suga Kazuhiro , Kawasaki Shota
Published: January 2011

Interaction effect of multiple flaws on ductile fracture is studied numerically by using Gurson’s constitutive equation. Based on experiments, 2 parallel cracks problems and 3 parallel cracks problems are simulated. In some problems, crack coalescence does not occur, and it occurs in other cases. In all cases, ductile fracture processes are obtained, and results are compared with experimental ones. Fracture pattern and load-displacement curves agree well with experiments. It is shown that void growth term is dominant for the coalescence of cracks. Slanted crack problem and non-uniform length crack problem are simulated and evaluation method of multiple cracks problem is discussed.
Cracking of Non-PWHT’d Carbon Steel Operating at Conditions Immediately Below the Nelson Curve
Authors: McLaughlin James , Krynicki Joseph , Bruno Thomas
Published: January 2011

Cracking was observed in non-PWHT’d carbon steel piping and vessels operating at conditions immediately below the “Nelson” curve. This curve provides a threshold limit for high temperature hydrogen attack as a function of the equipment operating temperature and hydrogen partial pressure. This curve is based on industry experience with steel equipment operating for many years in high temperature high pressure hydrogen service. Our investigation indicated that cracking occurred in 2 stages. Stage 1 or the initial stage of cracking appeared to occur as a result of intergranular “hydrogen assisted” cracking very similar to high temperature hydrogen attack. It appears that Stage 1 cracking is driven by the combined effects of residual welding stresses and “methane pressure” stresses from the decomposition of carbides. The circumstances of the observed cracking indicate that, unlike high temperature hydrogen attack, it occurred over a relatively short period of time after an operating change to a higher hydrogen partial pressure. Stage 2 cracking results from sulfide scale packing the crack during high temperature operation in a sulfidizing environment and causes the Stage 1 cracks to further propagate through wall. Once the sulfide scale filled crack cools down during a shutdown, the surrounding metal contracts around the scale and causes a high load on the crack tip which promotes further crack propagation. The presence of dissolved hydrogen in the steel further promotes Stage 2 crack propagation during shutdown periods.
Assessing Equivalent Viscous Damping Using Piping System Test Results
Authors: Nie Jinsuo , Morante Richard J. , Hofmayer Charles H. , Ali Syed A.
Published: January 2011

The specification of damping for nuclear piping systems subject to seismic-induced motions has been the subject of many studies and much controversy. Damping estimation based on test data can be influenced by numerous factors, consequently leading to considerable scatter in damping estimates in the literature. At present, nuclear industry recommendations and nuclear regulatory guidance are not consistent on the treatment of damping for analysis of nuclear piping systems. Therefore, there is still a need to develop a more complete and consistent technical basis for specification of appropriate damping values for use in design and analysis. This paper summarizes the results of recent damping studies conducted at Brookhaven National Laboratory.
On the Correct Application of the 100-40-40 Rule for Combining Responses Due to Three Directions of Earthquake Loading
Authors: Nie Jinsuo , Morante Richard J. , Miranda Manuel J. , Braverman Joseph I.
Published: January 2011

The 100-40-40 rule is often used with the response spectrum analysis method to determine the maximum seismic responses from structural responses resulting from the three spatial earthquake components. This rule has been referenced in several recent Design Certification applications of nuclear power plants, and appears to be gaining in popularity. However, this rule is described differently in ASCE 4-98 and Regulatory Guide 1.92, consequently causing confusion on correct implementation of this rule in practice. The square root of the sum of the squares method is another acceptable spatial combination method and was used to justify the adequacy of the 100-40-40 rule during the development of the Regulatory Guide 1.92. The 100-40-40 rule, when applied correctly, is almost always conservative compared to the SRSS method, and is only slightly unconservative in rare cases. The purpose of this paper is to describe in detail the proper application of the 100-40-40 rule, as prescribed in ASCE 4-98 and in Regulatory Guide 1.92, and to clarify the confusion caused by the two different formats of this rule.
Offshore Pipelines Thermo-Mechanical Performance
Authors: Suschitz Luca , Nair Shaju
Published: January 2011

Higher production fluid temperature and pressure requires increased offshore hydrocarbons pipeline performance. Thermal-mechanical loading induces considerable compression in pipelines resulting in potential elastic instability. One family of design solutions restrains the pipeline, impeding its displacement and maintaining the axial compression. A second family of design solutions aims to relieve this compression in a controlled manner thereby allowing the pipeline to either displace sideways or axially. For both options, the global buckling design needs a peculiar set of input parameters, methods of assessment and an acceptability criteria. This paper presents an original development in the methodology to determine the sections of the pipeline route that are susceptible to global buckling. The process has found remarkable benchmark in field applications. Specific attention is given to a macro scale of the entire pipeline routing while addressing anisotropies such as out-of-straightness, crossings, sideways-installation adjustments, sequence of loading and three-dimensional seabed bathymetry.