bluuebunny/crossref_metadata_2025_split huggingface dataset:
3D Solution of the Flow Field in a Boiling Water Reactor Annulus due to a Recirculation Outlet Line Break Using Heat Transfer Analogy
Authors: Ananth Raju , Fong Minji 
Published: December 2016

During a Loss of Coolant Accident (LOCA) in a Boiling Water Reactor (BWR), subcooled water flows past jet pump assemblies located in the annular region between the Reactor Pressure Vessel (RPV) and the shroud as it moves toward the break location and is subsequently discharged from the RPV. Flow loads caused by such an event are required design basis loads that must be considered for BWR internal components.
In previous works [1, 2], the three dimensional (3-D) flow field problem was simplified to be a 2-D problem by assuming the radial velocity variations to be negligible. The 2-D problem was solved using complex function techniques by assuming a potential flow. Further, the velocity field had to be suitably scaled up to account for the presence of components such as the jet pumps in the annulus. In order to solve the problem in the realistic environment of a populated annulus, this paper illustrates a methodology where Finite Element Analysis (FEA) is used to perform a 3-D potential fluid flow calculation utilizing the analogy that exists between steady state heat transfer and potential flow problems.
For an ideal fluid, the potential flow and irrotational flow assumptions will result in the Laplace equation as the governing equation for the velocity field. This is the same equation that governs the steady state heat transfer in any domain of interest where the temperature field is determined by solving the Laplace equation and applying the appropriate boundary conditions.
Once the analogy between steady state heat transfer problems and potential flow problems governed by Laplace equation can be established, any commercially available finite-element code may be employed to solve such fluid flow problems involving complicated regions of interest by employing elements meant to solve heat transfer problems. For illustration purposes, a LOCA flow problem will be solved using Finite Element Model (FEM) thermal elements and compared against 2-D flow field results.
Risk Management of Peening Application to Nuclear Power Plant Components
Authors: Kobayashi Hiroyuki , Nakama Shohei , Fujino Takushi 
Published: December 2016

The improvement of residual stress in component surface by shot peening was applied mainly to the core shroud for the purpose of preventing stress corrosion cracking (SCC) in BWR Nuclear Power Plants (NPP’s) in the late 1990’s. Subsequently, advanced peening techniques using pressure shock wave impact, such as Water Jet Peening (WJP) and Laser Peening (LP), were developed. Subsequently, PWR NPP’s also adopted these techniques to prevent Primary Water Stress Corrosion Cracking (PWSCC) in Japan.
These peening techniques were implemented with performance demonstration tests conducted in advance by JAPEIC (Japan Power Engineering and Inspection Corporation). Operators also planned on-site applications, including risk management. This risk refers to the risk that components integrity or outage schedule may be affected by peening. In fact, Tsuruga-2 faced a difficult situation, despite the preparation for the risk that cracks might be detected during pre-peening inspection. It was beyond anything that could have been expected.
This paper presents the experiences of peening technique applications at Tokai-2 (2 loop, 1,100 MWe BWR) and Tsuruga-2 (4 loop, 1,160 MWe PWR), analyzing common risk factors present in general application cases based on specific experiences. As a result, high-risk possibilities are indicated depending on risk management, and the necessity is shown for controls that take into consideration materials, design, operation time and other factors. Additionally, the relationship between the risk management and Codes & Standards is also referenced.
Probabilistic PTS Analysis
Authors: Bhimanadam V. R. , Blom F. J. 
Published: December 2016

This paper describes the effect of variability of fracture toughness of nuclear pressure vessels during a PTS event. The model used in this paper is based on the NESC-1 experiment. To determine the behavior of the surface breaking defect NRG performed three dimensional finite element calculations and subsequently extended these calculations to the probabilistic calculations. Three-dimensional finite-element model of the cladded cylinder was generated using ANSYS with semi-elliptical surface crack having a crack depth of 75 mm and a crack length of 205 mm. The cylinder specimen was subjected to thermal-shock and centrifugal loading conditions and analyzed with a themo-elastic-plastic material model and subsequently determined the fracture mechanics parameters (J and K) along the elliptical crack front as a function of time and temperature. The determined stress intensity factor K has been compared with the available cleavage fracture toughness (KJC) data with 50% fracture probability which has been obtained from the Master Curve according to BS7910. The comparison has been performed for the locations in the base metal as well as the locations in the heat affected zone.
Deterministic analysis has been extended to probabilistic analysis to calculate the failure probability for the crack initiation at the locations in the base metal as well as the locations in the heat affected zone along the crack front by considering probabilistic distributions from Master Curve and FAVOR. Master Curve analysis through the ASME code case N-629 has been applied to the material. Results obtained from these two methods have been compared and also the results are used to compare the inherent safety factors in the deterministic analysis using RTNDT and Master Curve.
R&D of Structural Alloy for Molten Salt Reactor in China
Authors: Li Zhijun , Leng Bin , Yuan GuangZhou , Wu Guanyuan 
Published: December 2016

Structural materials (Alloy N) of molten salt reactor are used to fabricate the reactor vessel, loops, heat exchangers, control rod sleeves, which should have long-term stable properties and enough strength and ductility in multiple extreme environments e.g. high temperature, high pressure, corrosion and neutron irradiation.
The choice and evaluation of materials for different reactor components depend on the specific working conditions of the components. For thorium molten salts reactor-solid fueled one (TMSR-SF1), the reactor vessel is filled with argon gas and the outside parts are exposed to the air. It contains a primary loop of FLiBe coolant and a secondary loop of FLiNaK coolant. The outlet temperature is 700 °C, and the core operating pressure is less than 0.5MPa.The calculated total neutron fluence after 20 years’ service is 2.01×1017n/cm2 for the reactor vessel. Based on these working conditions, the different properties of structural materials are considered and evaluated.
The most important issues of TMSR-SF1 materials include the supply of materials (their preparation, processing, welding, for example) and their qualification in ASME Code. We review alloy N for TMSR-SF1 including alloy physical properties, creep and rupture life time, thermal stability, corrosion, irradiation performance, as well as reliability evaluation of the weldments and so on.
Sealing Properties of Joints Between Metal Surfaces Holding Wavy Defects
Authors: Bourniquel Julien , Lasseux Didier , Rit Jean-François 
Published: December 2016

Internal sealing of many power plant valves is performed by a metal to metal contact, compatible with severe pressure and temperature conditions of liquid water flow. Contact surfaces are treated by hard alloy coating and lapping in order to enhance sealing properties. However, wavy deformation may occur on the surface because of stress induced during manufacturing and all through the device life, for example as a consequence of thermal shocks.
We aim to quantitatively correlate the leakage which results from wavy defects through experimental measurements and predictions obtained from modeling. In order to leave as few uncontrolled parameters as possible, the sealing joint is isolated from the valve device as a separate system consisting of a flat and rigid plug and an annular seat holding a wavy defect. The leak sealing properties of the metal to metal contact are expressed by a transmissivity value. This physical quantity has a length cube dimension and relates flow rate to pressure difference by a linear relation. It is intrinsic to the interstitial geometry.
The experimental section of this paper relies on two key steps. First, a sample seat, holding a controlled wavy defect, is manufactured. This is achieved by elastically deforming a cylindrical metal part having a tube shape in a lathe’s mandrel. The end-face is machined, lapped and finally the part is released, leaving a wavy defect on the end-face as a consequence of elastic relaxation. Secondly, the transmissivity of the sample seat, pressed against a flat, rigid plug, is determined from leak rate and fluid pressure measurements carried out on specific experimental devices. The procedure is repeated for several clamping loads yielding the transmissivity to apparent contact pressure dependence.
In the modeling section, leakage is calculated following three successive main stages. In the whole procedure, the flow action on the structure is neglected. First a geometrical model of the seat holding the manufactured wavy defect is built using a linear modal decomposition on a basis of vibrational eigen modes allowing to represent any kind of defect. Secondly, the remaining open space -i.e. the opening between the plug and seat surfaces-resulting from the clamping force normal to the mean plane of the contact, is computed. A finite element model is used to solve the contact problem on the non-trivial form of the wavy defect on the whole 3D geometry. Finally, the transmissivity is computed by solving the Reynolds equation for incompressible flow within the percolating opening between the plug and the seat. The final result of the modeling section is a transmissivity function of the apparent contact pressure, characterizing the leakage of the given wavy defect on the seat.
Predictions of the transmissivity reveal to be in good agreement with that obtained experimentally at low clamping forces and overestimate the leakage for larger ones.
Recommended Configuration of Boiler Drum Water Level Control and Protection System
Authors: Liu Zhelin , Harris John Michael , Liu Jichuan , Yu Shengli , Zhou Mingxin 
Published: December 2016

For a high pressure boiler above 1,400 psi (9.7 MPa), the boiler drum level Control and Protection systems serve two distinct, but complementary functions. The Control system needs to control drum level. Protection is done for safety. If the drum level reaches a critical level, the Protection system shuts down the boiler to avoid damage.
One must meet ASME Boiler Code requirements, get the job done correctly, and minimize the number of instruments required. By ASME Boiler Code, one must have at least one Boiler Water Gauge Glass. How many Control instruments are needed? If a single boiler drum water level instrument is used and this instrument suffers a failure, there is a problem. Even if two instruments are used and one fails, one does not know which to trust. Logic dictates that one requires a minimum of three Control instruments. Before discussing Protection, it is good to discuss the characteristics of the instruments available:
Study on Evaluation Procedure for Calculating the Stress Intensity Factor of Flaws Beneath RPV Cladding During Pressurised Thermal Shock Events by FE Analysis
Authors: Sakamoto Hiroyuki , Hirota Takatoshi , Ogawa Naoki 
Published: December 2016

Elastic-plastic finite element (FE) analysis is performed to determine the plastic behavior of the reactor pressure vessel (RPV) inner surface caused by rapid cooling during pressurized thermal shock (PTS) events. However, as the J-integral is not path-independent for elastic-plastic material in the unloading process, it is necessary to apply a suitable correction method using elastic material. In addition, it is also necessary to consider the effect of the welding residual stress appropriately.
Therefore, we investigated the stress intensity factor derived from FE analysis based on a model consisting of elastic-plastic cladding and linear elastic low-alloy steel with subsequent plastic zone correction, since the stress level of low-alloy steel remains within the elastic region except the crack front during a PTS event. Furthermore, we examined whether the stress mapping method is applicable for reflecting the effect of welding residual stress in FE analysis, even though the plastic strain generated during welding is ignored.
Failure Analysis of Spiral Water Wall Tube Cracking in Ultra-Supercritical Power Plant Boiler
Authors: Si Jun , Yan Zhenrong , Chen Jianjun 
Published: December 2016

The efficiency of conventional boiler/steam turbine fossil power plants has strong relationship to the steam temperature and pressure. At present, steam temperatures of the most efficient fossil power plants are now in the 600°C range. Higher-strength materials are needed for upper water wall tube of boilers with steam pressure above 24 MPa. A high-strength 2.5%Cr steel recently approved by ASME code as T23 is the preferred candidate material for this application. Due to its superior properties, T23 steel is typically not post-weld heat treated. However, after several years running there are a lot of incident reports for T23 tubes especially the breakage of weldment in the ultra-supercritical power plant. This is cause for concern for T23 tubes weldment used under high temperature environments. Previous studies showed that the residual stress may play an important role to the performance of spiral water wall tube.
In this paper, the distribution of residual stress in T23 tube weldment has been investigated in detail. Inner wall cracks were found at the butt-jointed seam region of spiral water wall tubes by radiographic testing after one year’s operation. Failure analysis of the spiral water wall tube cracking was conducted by chemical composition analysis, mechanical testing and finite element analysis in this paper. It was found that localized residual stress after the weld process caused concentrated stress, which is the primary reason for failure. Our studies illustrate the necessity of post weld heat treatment for the T23 tubes used under high temperature.
Improvement of the LBB.ENG2 Circumferential Through-Wall Crack J-Estimation Scheme
Authors: Olson Richard , Kalyanam Sureshkumar , Park Jeong Soon , Brust Frederick W. 
Published: December 2016

The LBB.ENG2[1] circumferential through-wall crack (TWC) J-estimation scheme forms the basis for the Extremely Low Probability of Rupture (xLPR)[2] probabilistic pipe fracture analysis for TWC elastic-plastic fracture mechanics (EPFM) stability assessment. The LBB.ENG2 methodology uses a reduced thickness pipe wall analogy to approximate the behavior of actual cracked pipe and sets the thickness of the reduced section by making the usual cracked pipe limit load assumption. Sometime during the original LBB.ENG2 development process, it was discovered that LBB.ENG2 was not as good as desired at predicting the maximum moment carrying capacity of pipe fracture experiments with longer cracks. Accordingly, the effective thickness equation was modified to be 1.0 at crack angles less than π/4, 4/π at angles greater than π/3, and linear between these values using a so-called ψ function.
When LBB.ENG2 was coded for the TWC stability module for xLPR, TWC_fail, the behavior described above was implemented. Quite unexpectedly, with the new coding, exploration of TWC_fail’s bounds uncovered two discontinuities in the complete moment-pressure-critical crack size failure surface. Subsequently, it was found that these discontinuities were caused by the discontinuity in the derivative of the ψ function.
This paper documents the approach used to smooth the TWC_fail moment-pressure-critical crack size surface by making a ψ function fit that minimizes the difference between J from LBB.ENG2 and J from finite element analyses results. The results of the finite element analyses and fitting methodology are described and the basic equations for the solution are presented. Following this, the new ψ function is applied to cases to evaluate the efficacy of the approach.
Effects of Hot Water and Holds on Fatigue of Stainless Steel
Authors: Solin Jussi , Alhainen Jouni , Karabaki Ertugrul , Mayinger Wolfgang 
Published: December 2016

Direct strain controlled LCF data for solid specimens is still very rare. In PVP2013-97500 and PVP2014-28465 we reported results for niobium stabilized X6CrNiNb1810mod steel (type 347) fatigued in 325°C and 200°C PWR water according to VGB water chemistry specification. New data in this paper further confirms the conclusions: we are unable to repeat as high Fen factors or short lives as predicted according to NUREG/CR-6909.
The slowest strain rate used 4·10−6 in 325°C water would predict Fen > 12, i.e. laboratory specimen data below the current ASME design curve, but our results are superior for this steel generally used in German NPP’s. However, the difference is not necessarily grade specific. Use of 100% relevant fabricated material batch and standard LCF methodology are regarded to play an important role.
Notable hardening can be measured, when long duration holds in elevated temperatures are introduced between blocks of cyclic strains at lower temperatures. This is the case for thermal gradient loaded primary circuit components, e.g. the PWR pressurizer spray lines or surge line, which connects the pressurizer to primary coolant line. In PVP2011-57942 we reported improved endurances in fatigue tests aiming to roughly simulate steady state operation between fatigue transients in such NPP components. New test types have been introduced to generalize the results.
Mechanisms of time and temperature dependent relaxation of fatigue damage and/or improvement of material fatigue performance during holds are not yet fully revealed, but the rate controlling thermal activation energy is below shown to be near that for vacancy and interstitial atom diffusion. This allows us to draft a thermodynamic prediction model.
Improved accuracy of fatigue assessment helps in focusing optimally scheduled nondestructive testing to the most relevant locations and maintaining high level of reliability without excessive cost and radiation doses for inspection personnel.
This paper provides previously unpublished experimental results and proposes methods to improve transferability of laboratory test data to fatigue assessment of NPP components. The effects of material, water environment, temperature and service loading patterns are discussed.
Oore-Burns Function of Form Application in Numerical Treatment of Mode I Flat Crack Problem in Infinite Body
Authors: Orynyak Igor , Batura Anatolii , Oryniak Andrii , Lokhman Igor 
Published: December 2016

The general approach of numerical treatment of integro-differential equation of the flat crack problem is considered. It consists in presenting the crack surface loading as the set of the polynomial functions of two Cartesian coordinates while the corresponding crack surface displacements are chosen as the similar polynomials multiplied by the function of form (FoF) which reflects the required singularity of their behavior. To find the relations matrixes between these two sets a new effective numerical procedure for the integration over the area of arbitrary shape crack is developed. In based on the classical hyper-singular method, i.e. Laplace operator is initially analytically applied to the integral part of equation and the resulting hyper singular equation is subsequently considered.
The presented approach can be implemented with any variant of FoF, but Oore-Burns FoF, which was earlier suggested in their famous 3D weight function method, is supposed to be the most accurate and universal. It takes into account all points of crack contour, which provides perfect physical conditionality of the solution, but such FoF is relatively heavy in implementation and of low computational speed. The special procedure is developed for the approximation of the crack contour of arbitrary shape by the circular and straight segments. It allows to easily obtain analytical expression for Oore-Burns FoF, which greatly increases the calculation speed and accuracy. The accuracy of the considered method is confirmed by the examples of the circular, elliptic, semicircular and square cracks at different polynomial laws of loading.
The developed methods are used in the implemented procedure for crack growth simulation. It allows to model growth of crack of arbitrary shape at arbitrary polynomial loading, at that all contour points are taken into account and can expand with their own speeds each. Procedure has high accuracy and don’t need complex and high-cost re-meshing process between the iterations unlike FEM or other numerical methods. At that usage of Oore-Burns FoF provides high flexibility of the presented approach: unlike similar theoretical methods, where FoF calculation procedure is rigidly connected with the crack shape, which complicates the adequate crack growth modeling, the used FoF automatically takes into account all points of crack contour, even if its shape became complex during the growth. Presented crack growth procedure can be effectively used to test accuracy and correctness of correspondent numerical methods, including the newest XFEM approach.
Core Seismic Experiment of a Full-Scale Single Model for a Fast Reactor
Authors: Iwasaki Akihisa , Sawa Naoki , Matsubara Shinichiro , Kitamura Seiji , Okamura Shigeki 
Published: December 2016

A fast reactor core consists of several hundred core elements, which are hexagonal flexible beams embedded at the lower support plate in a hexagonal arrangement, separated by small gaps, and immersed in a fluid. Core elements have no support for vertical fixing in order to avoid the influence of thermal expansion and swelling. These days, in Japan, larger earthquake vibrations are postulated in seismic evaluations. So, it is necessary to consider vertical displacements (rising) and horizontal displacements of the core elements simultaneously because vertical seismic vibrations are larger than the acceleration of gravity. The 3D vibration behavior is affected by the fluid force of the ambient coolant and contact with the surrounding core elements.
In this study, single-model vibration tests using a full-scale test model were conducted, and the basic characteristics of 3D vibration behavior of the core element were examined. In addition, structures restricting vertical displacements (dashpot structure) were devised, and their effectiveness was verified. As a result of the tests, the effects of the ambient condition (in air, in static water, and in flowing water), gap between the pads, vibration directions, vibration waves, and dashpot structures on the vibration behavior of the core element were examined. As regards the ambient condition, the vertical displacements were larger in flowing water that simulates the coolant flow than in air and in static water, because of upward fluid force in flowing water. As regards the gap between the pads, the larger the gaps was, the stronger the interferences due to horizontal displacements, and the smaller the vertical displacements were. The dashpot structure was verified to be suitable for reducing vertical displacements.