bluuebunny/crossref_metadata_2025_split huggingface dataset:
Comparison of Numerical Methods for Calculation of Vertical Soil Pressures on Buried Piping due to Truck Loading
Authors: Marohl Michael P. H. 
Published: November 2014

Calculation of pressures in a soil body due to finite loads imposed on the soil surface is a necessary step in the design and analysis of buried commodities. This study compares two commonly-applied numerical methods used to develop the vertical soil pressure profiles applied to buried pipes.
The methods compared in this study differ in theory, basis, assumptions, complexity, and results, and therefore the comparison is meaningful. Provided is a comparison between design vertical forces on different sizes of buried pipes at a range of soil depth, determined using an integration of Boussinesq’s equation [1] and the method specified by AASHTO [2]. The Boussinesq equation is defined as a function of location in varying two dimensional soil planes and the integration is performed over the boundary of the pipe as well as the applied soil surface footprint. The soil surface loading considered in this study includes the AASHTO Design Truck and the AASHTO Design Tandem, positioned as required by the AASHTO LRFD code [2]. Recommendations for application of the results is provided based upon the resulting force magnitude calculated and ease of application of the methods. Consideration of the effects of redistribution of loading due to pavements or other rigid surfaces is outside the scope of this study.
Creep Life Evaluations of ASME B31.1 Allowance for Variation From Normal Operation
Authors: Cohn Marvin J. 
Published: November 2014

The ASME B31.1-2012 Power Piping Code (Code) paras. 102.2.4, 102.3.3, and 104.8.2 provide an allowance regarding operating above the design temperature and internal pressure for short time periods.
This study is an evaluation of the permitted increased life consumption associated with the above Code operating allowances for piping materials operating in the creep range. Three base metal materials are considered in this study, ASTM A335 Grades P11, P22, and P91.
Two case studies were evaluated, A) 15% stress increase for 10% of the operating hours, and B) 20% stress increase for 1% of the operating hours. It was determined that these allowances increased the base metal creep rupture life consumption of Grade P11 material up to 8%, Grade P22 material up to 14%, and Grade P91 material up to 25%. Allowance A results in permitting significantly more creep damage life consumption than Allowance B.
Main steam and hot reheat piping system typical operating temperatures and stresses were evaluated for these variation allowances. This study revealed that Grade P22 base metal creep damage is slightly more sensitive to stress than Grade P11 material creep rupture damage, and Grade P91 base metal creep damage is substantially more sensitive to stress than Grade P22 material creep rupture damage.
Advanced INLAY System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants
Authors: Yamamoto Kazuhide , Kizawa Masahiko , Kawazoe Hiroki , Kobayashi Yuki , Miyoshi Yoshiyuki 
Published: November 2014

Since PWSCC has been observed in Alloy 600 used for butt welds between the low alloy steel RV-nozzles and stainless steel pipes in recent years, MHI has developed the inlay technology for outlet/inlet nozzles as a preventive maintenance and repair method. The inner surface, even if it has PWSCC, is exchanged from Alloy 600 to Alloy 690 which has excellent PWSCC resistance, by applying inlay.
This paper introduces the experiences of the MHI inlay System (RV-INLAY) and describes the merits as a counter measure against PWSCC. MHI has adopted the Cylindrical container for RV-INLAY. The Cylindrical container is installed into RV when the cavity is filled with water, then we can secure an atmospheric working space inside outlet/inlet nozzles. 3 or 4 manipulators are installed on the bottom of the container and carry out precise works like ambient temperature temper bead welding, UT, ECT, dimensional or visual inspection exchanging many kinds of tools (end-effectors) on the tip of the manipulators remotely. Heavy devices for machining, blast-decontamination or nozzle-plug handling are carried down into the container and set remotely using the special handling crane. Manipulators and other heavy devices move remotely to reduce the radiation exposure and simultaneously in 3 or 4 nozzles to shorten the working period. By using the Cylindrical container, we can directly access the inside outlet/inlet nozzles. This is different from the other preventive maintenance methods which need to access from outside of outlet/inlet nozzles. We can apply RV-INLAY from inside the nozzles without considering the difficulty in access due to the tight space and high radiation outside the nozzles. Furthermore when a RV is manufactured in a factory, we apply PT for the groove face before welding and the surface after welding. During RV-INLAY operation, we apply PT in the same manner because we can secure an atmospheric working space inside outlet/inlet nozzles. As a result, we can ensure the quality of welding same as manufacturing process in a shop. In addition, we don’t have to consider the crack propagation after RV-INLAY and we can perform ISI without preparing repair equipment against PWSCC. The Cylindrical container has a seal function same as the seal-plate used in usual outages, and it can separate the water inside the cavity from RCS. While the Cylindrical container is installed into RV, the RCS is drained and RCS piping works can be performed in parallel. In fact, we have shortened the outage duration by implementing RV-INLAY and RCS piping works in parallel. This is one reason why MHI has not selected under water welding but RV-INLAY.
Consistency Between Evolution of Codified Rules and SFR Prototype Development: Astrid and RCC-MRX Example
Authors: Blanc M. , Pétesch C. , Gélineau O. , Lebarbé T. , Blat-Yrieix M. 
Published: November 2014

The main objective of a standard or a code is to capitalize the technical feedback of constructions. It is the case of the RCC-MR code (now RCC-MRx) which has integrated the technical results of projects such as PHENIX, SUPERPHENIX, Jules Horowitz Reactor, ITER vacuum vessel. These rules are now planned to be used for new plants and it appears that they have to be updated to meet the needs of Gen IV projects like the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project or MYRRHA (Multi-purpose hybrid research reactor for high-tech applications) project.
This paper presents the ASTRID project approach to initiate in the earlier stage of the design an evaluation of the rules in order to improve as soon as possible the code. The result of this approach in terms of code modifications and improvement objectives are also presented.
Application Case of RCC-MRX 2012 Code in Significant Creep
Authors: Lejeail Yves , Lamagnère Pierre , Petesch Cécile , Lebarbé Thierry , Matheron Philippe , Martin Antoine 
Published: November 2014

This paper gives an application case of the RCC-MRx mechanical design code for nuclear components in the domain of significant creep. It could be seen as a guide for engineers who have to perform mechanical creep and creep-fatigue analyses with this code. The application case is a spherical shell with an internal radius of 1250 mm and a thickness of 50 mm, which is made in AISI 316L. The structure is assembled by manual arc welding of plates using 19Cr-12Ni-2Mo type rods. The shell is intended to operate at 550°C mean temperature under an internal pressure of 5 MPa due to argon gas (the atmospheric pressure is considered outside). During operating, the internal temperature is equal to 600°C and the external temperature is 500°C. The shell is periodically stopped for servicing: during these periods, the temperature is 50°C and uniform, and the internal pressure is equal to the atmospheric pressure. A cycle is defined by a 500 h dwell time at 550°C followed by a 24 h arrest at 50°C; transient temperatures, thermal shocks during starting and arrest are neglected. The application of the RCC-MRx is shown for different damages, such as:
- Excessive strain, plastic instability and rupture for primary loadings (negligible and significant creep);
- Ratcheting, fatigue, creep-fatigue for all type of loadings.
The effect of welds on expected life (creep, fatigue), and the comparison of calculated lives and allowed operating cycles for 316L and 316LN are described.
Brittle Fracture Probabilistic Assessment of WWER-1000 RPVs
Authors: Orynyak Igor , Zarazovskii Maksym , Batura Anatolii , Borodii Mykhaylo , Danil’chuk Evgen 
Published: November 2014

This paper presents a methodology for brittle fracture probability assessment of WWER-1000 RPVs subjected to PTS. The main attention of the study is focused on the definition of the stochastic input data: fracture toughness, CTB, size and shape of the defects.
Fracture toughness of RPV metal in the initial state is determined, separately for the base metal and the welds, as the mean curve lines of relevant experimental data. Metal embrittlement is taken into account by increasing CTB.
Based on the results of the Ukrainian surveillance specimens program, a CTB database of WWER-1000 RPVs is created. As a result of processing these data, the CTB temperature dependence on the neutron fluence and chemical composition are obtained. The probability of the fracture toughness curve position on the temperature axis is defined by the CTB standard deviation in accordance with normal distribution law.
The distribution laws of depth and shapes of defects are taken according to the NRC data and verified on the base of the statistics of the defects, which were found in the WWER-1000 RPVs of the Zaporozhye NPP.
In fact, probabilistic calculation of the RPV brittle fracture is performed similarly to the deterministic one. Fields of temperature and stresses in different RPV zones in each timepoint of considered emergency scenario are calculated by the high-effective variant of transfer matrix method in axisymmetric elastic formulation. Semielliptical axial cracks of different sizes and proportions are conservatively considered on the inner surface of RPV. SIF along the crack front is defined using the original variant of Weight Function Method. The probability of failure of a particular defect is determined as the probability that CTB of the appropriate metal reaches the maximum allowable CTB value for the scenario.
The presented method is used for the renewal of operating licenses process of Unit 2 RPV of the South Ukrainian, and Units 1 and 2 RPVs of the Zaporozhye NPPs.
Environmental Influences on the Fatigue Assessment of Austenitic and Ferritic Steel Components Including Welds
Authors: Roth Armin , Herbst Matthias , Rudolph Jürgen , Wilhelm Paul , Schuler Xaver , Herter Karl-Heinz , Kammerer Matthias , Weissenberg Thomas 
Published: November 2014

The fatigue assessment of safety relevant components is of importance for ageing management with regard to safety and reliability. For cyclic stress evaluation, different country specific design codes and standards provide fatigue analysis procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as limiting criteria, the influence of different factors like e.g. environment, surface, temperature and data scatter must be taken into consideration in an appropriate way. In this context there is a need of consolidating and increasing the current knowledge.
In the framework of an ongoing three years German cooperative project performed by Materials Testing Institute MPA Stuttgart and AREVA GmbH (Erlangen) it is the aim to both improve the state of the art based on an experimental program on the factors mentioned above including hold-times at transient free static load and on the derivation of a practicable engineering fatigue assessment concept. Emanating from a review of the current state of the art the cooperative project is split up into three major parts:
1) Experimental investigations concerning the influence of loading parameters and environmentally assisted fatigue (EAF) effects (light water reactor environment) on the fatigue strength of ferritic steels including weldments.
2) Experimental investigations concerning the influence of long hold times and the EAF effects on the fatigue strength of austenitic and ferritic steels.
3) The results of the outlined experimental program and published results will constitute the input for the proposal of an engineering fatigue assessment concept. This concept includes the differentiation between numerous factors of influence as an essential feature. In this context the margins between mean data curves and design curves are to be discussed in detail considering the factors of influence in general and EAF in particular.
Based on a comprehensive consolidation of the state of the art and previous investigations in air and in light water reactor environment an experimental program is set up with the following key aspects:
- Strain controlled fatigue tests on welded (microstructure of the weldment excluding microscopic and macroscopic weld notch effects) and unwelded smooth laboratory specimens subjected to constant and variable strain amplitude loading in air and light water reactor environment.
- Strain controlled fatigue tests on notched specimens for the consideration of multi-axiality effects in air and light water reactor environment.
- Strain controlled fatigue tests on smooth round laboratory specimens in air and in light water reactor environment focusing on long (power plant relevant) hold time effects.
Understanding Behaviors of Bolted Flanges With Non-Axisymmetric Temperature Distribution
Authors: Jung Gong H. , Kim Minsu 
Published: November 2014

It is critical to maintain uniformity of gasket stress in control against leakage in piping and heat exchangers. Many innovative bolting tools have been developed, but their benefit could be downgraded if there were sources disrupting initially achieved uniformity of gasket stress.
The characteristics behaviors of gasket stress in flanges with NATD (Non-Axisymmetric Temperature Distribution) were studied using finite element analysis on ANSI standard flanges.
It was found that change of gasket stress was independent of the magnitude of initial bolt stress and size of flange, but strongly dependent upon type of flange class. Empirical curves (curve fitting equation) were proposed to determine change of gasket stress which was function of degree of NATD and flange class (rigidity index). The change of bolt stress with NATD was relatively small compared to the change of gasket stress, which may make it difficult to assess flange integrity by monitoring bolt stress in service.
A Comparison of the Deterministic and Probabilistic Aspects of the ASME III Fatigue Evaluation Procedures
Authors: Watson Chris , Beardsmore David , Wright Keith 
Published: November 2014

Fatigue is a common failure mode in mechanical components that are subject to repeated load cycles. In nuclear plants the cyclic loading is usually associated with variations of pressure and temperature. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provide fatigue evaluation procedures in ASME III Section NB-3220. It is required to calculate a Cumulative Usage Factor (CUF). The Code criterion for acceptance is that the calculated CUF must be less than one. The ASME method is deterministic and it is assumed that uncertainties are accounted for by largely un-quantified inherent safety margins in the assessment.
The conservatism in the ASME III Code fatigue evaluation methods may come from the assessment procedure used and/or the fatigue design curve. In this paper the inputs to an ASME III fatigue assessment that is based upon a finite element determination of stresses are outlined and assessed. Based upon these inputs possible areas of conservatism within the analysis are then discussed and conclusion drawn on the inherent margins.
In addition a probabilistic model that uses Monte Carlo (MC) methods is developed. By assuming the scatter in the fatigue data, used to generate the design fatigue curve, to be the most significant random variable, the MC model was used to calculate the conditional probability of initiation associated with CUFs from a number of plant representative examples.
Implementation in VIBIC of an Improved Time-Domain Simulation Model for Fluidelastic Instability in Tube Arrays
Authors: Sawadogo Teguewinde , Han Yingke , Mureithi Njuki 
Published: November 2014

VIBIC (Vibration of Beams with Intermittent Contacts) is a non-linear dynamics computer code developed and maintained by Atomic Energy of Canada Limited over the past 40 years in collaboration with universities. Its main application is the assessment of possible vibration damage to steam-generator and heat exchanger tubes. This assessment is done by performing simulations of the vibration response of beam-like structures to various flow-induced excitation mechanisms, such as turbulence buffeting, vortex shedding, and fluidelastic excitation. Fluidelastic excitation is potentially the most damaging flow-induced excitation mechanism. The fluidelastic effect has, until now, been incorporated in VIBIC using the frequency-based Connors model.
To properly perform a time-domain simulation of fluidelastic-induced vibration, a time-domain fluidelastic force model is needed. In the present work, a time-domain formulation of the fluidelastic forces based on the quasi-steady model is implemented in VIBIC. The time delay inherent to the quasi-steady model is taken into account by using a delayed displacement in the expression of the fluidelastic forces. The resulting modal equations are delay differential equations that are solved using a continuous extension of the Runge-Kutta method. Both linear and nonlinear fluid force models are incorporated. The fluidelastic instability results predicted by the models are compared to known theoretical and experimental results for validation. The predictions of the models are in good agreement with those results. The results given by the nonlinear quasi-steady fluidelastic forces are found to be more realistic than those of the linear quasi-steady fluidelastic forces. A realistic simulation of the post-instability behaviour is made possible through the use of the nonlinear fluidelastic forces.
Prediction of Margins in the TBS Seismic Considerations Analysis for Circumferential Surface-Cracked Piping Under Beyond Design Basis Seismic Loading
Authors: Uddin Mohammed , Brust Frederick , Wilkowski Gery , Zhang Tao , Betervide A. A. , Mazzantini O. , Fernandez Ricardo A. 
Published: November 2014

The Transition Break Size analysis is given in Draft Regulatory Guide 1216 for assessment of ECCS requirements. There are additional analyses required for seismic considerations for beyond design basis seismic loading. The beyond design basis loading was for seismic event with a probability of occurrence of 1e−6 per year, whereas safe-shutdown earthquake loading is typically closer to a probability of 1e−4 events per year. The peak-ground accelerations for US plants are typically in the 0.1 to 0.2 g’s range for SSE loading, while the 1e−6 seismic loading may be about 3 times higher (depending on site specific seismic hazard curves).
A simplified method was created for the TBS seismic consideration analysis, which is given in Appendix A and B in the Draft Reg Guide. The technical basis is in NUREG-1903. The TBS analysis approach utilizes a simple method for scaling the seismic stresses from the SSE to 1e−6 loading, allowing for a bilinear stress correction (linear to yield of the material and an ultimate point consistent with uncracked seismic tested pipe results). A best-estimate fracture analysis is then conducted using the ASME Section XI Service Level D flaw size, but a best-estimate fracture analysis uses more realistic material properties and more accurate fracture analyses than the ASME Code.
In this paper, the TBS flaw size was calculated by the Draft Regulatory Guide approach for the Atucha II nuclear plant in Argentina that is about to start up. Additionally a full 3D FE model of the plant including the whole NSSS, containment building, and supports between the building and the NSSS components was developed. Circumferential surface cracks were put in the nonlinear time-history FE model at the highest stressed locations in the primary pipe loop to determine the depth of the flaw that would fail at the 1e−6 seismic excitation to the plant building. This was done to assess the margins in the flaw size for the TBS analysis, and also characterize the magnitude of the LOCA.
Where the TBS simplified model showed that a surface flaw of 0.709 and 0.669 of the thickness for the RPV/hot-leg and pump/cold-leg (respectively) and 270-degrees around the circumference could be tolerated, the full 3D FE analysis showed that even a surface crack of 90-percent of the thickness and 270-degrees around the circumference would not reach crack initiation for the material used in this plant with its seismic hazard curve. The information developed here may also be useful for assessing the piping integrity of a plant once it has exceeded the Service Level D limits of the ASME Code.
Probabilistic Seismic Hazard Assessment in Countries With Low Seismicity
Authors: Demjancukova Katerina , Prochazkova Dana 
Published: November 2014

The region of the Czech Republic is mostly composed of the Bohemian Massif which is considered as a geological unit with low seismic activity. Nevertheless, all critical objects as the nuclear power plants, big dams etc. are built as aseismic structures. The nuclear installations have to satisfy the IAEA safety standards and requirements. One of important phenomena that have to be involved in the PSHA process is the diffuse seismicity.
In 2010 International Atomic Energy Agency issued a specific safety guide SSG-9 Seismic Hazards in Site Evaluation for Nuclear Installations. The key chapters are focused on general recommendations, necessary information and investigations (database), construction of a regional seismotectonic model, evaluation of the ground motion hazard, probabilistic seismic hazards analysis (PSHA), deterministic seismic hazards analysis, potential for fault displacement at the site, design basis ground motion, fault displacement and other hazards, evaluation of seismic hazards for nuclear installations other than NPPs.
In the paper a numerical example of seismic hazard assessment will be presented with emphasis on problems and particularities related to PSHA in countries with low seismic activity.