bluuebunny/crossref_metadata_2025_split huggingface dataset:
Beyond Design Basis Seismic Analysis for Atucha II Nuclear Plant
Authors: Zhang Tao , Brust Frederick W. , Wilkowski Gery , Xu Heqin , Betervide Alfredo A. , Mazzantini Oscar 
Published: October 2013

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. Nuclear power plants must be designed to maintain their integrity and performance of safety functions for a bounding set of normal operational events as well as abnormal events that might occur during the lifetime of the plant. Seismic fracture mechanics evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to an event with a probability of 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size for this case is 94-percent of the circumference. These analyses are performed by placing cracked-pipe-elements into a complete model of the primary cooling system including the reactor pressure vessel, pumps, and steam generators as summarized in the paper.
This paper summarizes these results and further shows how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent (about 120°) around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 10−6 seismic event accelerations, or 240 times higher than the SSE accelerations. This suggests there is a huge safety margin for beyond design basis seismic events and Atucha II plant rupture is pragmatically impossible. These surprising results are discussed and could be potentially applicable to other nuclear power plants as well.
Finite Element Analysis of Elastomers Using ANSYS
Authors: Kunnil Manoj , Yamarthi David , Kompally Santhosh K. 
Published: October 2013

Elastomeric materials have a capability to withstand large deformations and still be able to fully recover their original dimensions. Natural and synthetic elastomers and their derivatives can reach strains as high as 500–1000%. Engineering materials, such as crystalline metals are classified as linear elastic solids, whereas elastomeric materials are considered as nonlinear elastic solids. Elastomers present a very complicated mechanical behavior that exceed the linear elastic theory and contain large deformations, plastic and viscoelastic properties. Finite element (FE) is a powerful tool to analyze such elastomers.
Design of elastomeric systems in an industrial scenario generally requires (i) reliability and (ii) a minimum cycle time. This paper starts with a review of the hyperelastic theory, followed by a detailed discussion on the process involved in the material characterization of hyperelastic material like DuPont™ Viton® fluoroelastomer and polyacrylic elastomer in industrial application point of view. The paper also discusses guidelines to be followed in the various stages of material characterization such as testing, sampling and finite element simulation. Numerical stability issues associated with elastomeric modeling in finite element context and a set of guidelines to be followed in finite element analysis of elastomers are illustrated through a DuPont™ Viton® fluoroelastomer and polyacrylic elastomer pad vibration isolation systems. The above technique has been applied for designing vibration isolation systems for generators.
Recent Advances in Test vs Finite Element Correlations Through Design for Six Sigma Tools
Authors: Kompally Santhosh K. , Ramanath Vinay , Jeevanandan Karthikeyan , Kunnil Manoj 
Published: October 2013

In general, thermal generators have a combination of composites and metals in different assemblies. It is important to note that the material properties and interface stiffnesses change during the assembly process. Added to this change, the complex geometry and assembly procedures result in huge variation in material characterizations. These variabilities triggered a requirement of a unique process for material characterization at both component and assembly levels.
This paper covers the details of a 6-stage DFSS methodology, which involves filling the above-stated gaps by performing mechanical tests at component and sub-assembly levels, followed by series of finite element correlations at various stages of design cycle. This paper emphasizes a DFSS-based probabilistic approach, developed with a built-in validation for evaluating finite element variables to match with assembly tests. This paper also discusses the success of this DFSS-based process in bench marking with two test cases.
A Comparative Study of Direct Soil Model and Soil Spring Model in Common Basemat Analysis of APR1400 Nuclear Power Plant
Authors: Ahn Seong Moon , Lee Dong Seung , Park Jae Wan , Basu Bhaskar 
Published: October 2013

It is common practice to analyze the basemat using soil springs to represent the supporting soil media. A uniformly loaded mat foundation analyzed for a uniform spring constant will indicate uniform deflection. However, in reality, the behavior of the basemat is influenced by the flexibility of the supporting soil media. The settlement profiles of the basemat demonstrate smaller displacements at the edge and larger displacements toward the center. The spring constants used in the analysis should also consider stiffness effect of superstructure and the effect due to loading pattern.
For APR1400 Nuclear Power Plant, the mat foundation of the Reactor Containment Building and the Auxiliary Building is a common reinforced concrete mat covering an area approximately 12,210 square meter (110 m × 111 m) with a thickness of 3.05 m in the Auxiliary Building area and various thicknesses from 3.35 m to 10.06 m in the Reactor Containment Building area.
In this study, the soil layers are directly modeled into the structural analysis model using finite/infinite solid elements to represent the soil properties and then the analysis results are compared with the analysis using spring constants by iterative method, and uniform spring constant by conventional method. The results of analysis using the Direct Soil Model are similar to the results calculated by the Iterative Soil Spring Model, but have wide variance with the results calculated by the Uniform Soil Spring Model.
Improved Regulatory Oversight and Immediate Retrofits for Operating Pressurized Heavy Water Reactors for Beyond Design Basis Accidents
Authors: Nijhawan Sunil 
Published: October 2013

The most painful lessons that engineers will learn from reviews of the Fukushima disaster relates to the unacknowledged failure of regulators, designers and utilities in better retrofitting existing reactors in a timely manner to better withstand and mitigate known severe accident related challenges to reactor and containment integrity. Commercial nuclear power reactors have operated for over 50 years, and the first severe accident progression studies and understanding of related phenomena began to mature over 30 years ago. However, some very basic accident prevention, mitigation and management measures have not been required by the regulators and hence not initiated by the utilities.
In public reviews that inevitably followed, severe accidents in power reactors at Three Mile Island, Chernobyl and Fukushima have been often presented as site specific aberrations in design, operations, safety culture and acts of God beyond mortal imagination. If a sincere soul searching, regulatory overhaul and actual, effective, timely and far reaching measures are not taken for operating reactors, the world is bound to witness recurring severe core damage scenarios followed by series of studies that will predictably conclude that such severe accidents can happen only in other jurisdictions and in other designs and that ‘our’ reactors are ‘safe’ and only minor enhancements are required to placate the anxious public or inquiring regulators.
The paper examines the role played by regulators and other Canadian stakeholders in failure to ensure that the so-called residual risk from operating nuclear reactors is minimized in a timely manner. It presents specific examples of known deficiencies in PHWR designs that may degrade a sustained loss of power from a recoverable outcome into significantly more severe consequences. It discusses ideas for potential design retrofits and reiterates the need for more open, concerted and cooperative efforts internationally in accident progression and consequence analyses and supporting experiments.
Taking station blackout as example of an instigator of a severe accident, the paper discusses certain elements of the current PHWR designs that accelerate the onset and progression of core damage and present substantially degraded opportunities for mitigation and control. It demonstrates how in many cases, a sustained loss of power in a PHWR may cause a containment bypass with early and unacceptable off-site consequences. Many issues discussed in this paper are common to other reactor designs and regulatory jurisdictions.
Validation Results of TRACG05 Against BWR and ABWR Start-Up Test Data
Authors: Ochi Hitoshi , Takii Taichi , Kudo Yoshiro , Kagiyama Rui , Yamada Takeshi , Isono Kimiko , Nishijima Mitsuko , Kaneko Junichi 
Published: October 2013

Studies have been made on application of TRACG05, which is the latest version of TRACG code, to Best Estimate Plus Uncertainty (BEPU) approach for Anticipated Operational Occurrences (AOO) in BWR and ABWR. This paper presents validation results of TRACG05 against BWR and ABWR start-up test data.
TRACG05 analyses of BWR and ABWR start-up tests, such as generator load rejection (pressurization event) and recirculation pump trip (core flow decrease event), were performed using standard plant nodalization and nominal inputs.
In addition to the nominal cases, sensitivity studies and statistical analyses were conducted. The sensitivity studies include the sensitivities to TRACG05 model uncertainties, which have relatively high importance being tabulated in a Phenomena Identification and Ranking Table (PIRT), and nodalizaton. The sensitivity analyses include the random variation of TRACG05 model uncertainties, which have High ranking in PIRT, and initial core power. These uncertainties were preliminarily evaluated from comparisons of calculated results to separate effects test data, integral test data and component test data.
TRACG05 calculated results of the nominal cases agree to the start-up test data with reasonable accuracy. For statistical analyses, the standard deviations of the key parameters related to the safety parameters (critical power ratio, etc) for AOO were calculated at each time step. The test data basically fall inside the 2 sigma bands. It indicates that the differences between calculated results and test data are within the uncertainties of TRACG05 model and initial core power that are taken into account for safety analyses in BEPU process.
It is concluded that the TRACG05 code using the standard nodalization, inputs and model uncertainties is validated against the full-scale plant data and is capable of predicting the AOO transients.
Validation of TRACG05 for Application to BWR AOO
Authors: Matsui Asuka , Tamitani Masashi , Kudo Yoshiro , Takano Sho , Iwamoto Tatsuya , Nishijima Mitsuko , Kaneko Junichi , Ochi Hitoshi , Takii Taichi , Soneda Hideo 
Published: October 2013

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method.
In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues.
Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.
Semi-Analytical Study of Aerosol Washout by Rain
Authors: Lemaitre Pascal , Nuboer Amandine , Querel Arnaud , Depuydt Guillaume , Porcheron Emmanuel 
Published: October 2013

The accidents of Chernobyl and Fukushima have shown the necessity to better understand all the mechanisms implied in the scavenging of aerosol particles released to the atmosphere during a nuclear accident.
Among all the phenomena involved in the deposition of aerosol particles, we focus here on the aerosol particles scavenging by the raindrops below the clouds, also called washout (as opposed to the rainout, which concerns scavenging inside the clouds). The strategy of IRSN to enhance the knowledge and the modelling of any mechanism involved in the washout of aerosol particles by rain spans from environmental studies, to analytical ones.
The semi-analytical approach chosen here is halfway between these two modes of reasoning. A companion paper is also submitted to the conference to present the microphysical approach chosen at IRSN.
In order to perform this study, aerosol particles were dispersed in the TOSQAN chamber, which is a large cylindrical enclosure (4.8 m height with 1.5 m internal diameter). The aerosol particles once dispersed, synthetic rains of different kinds (from stratiform to convective rains) can be activated. Finally, the instantaneous spectral scavenging coefficients are determined from the spectral decrease of aerosol particles concentration in the chamber as a function of time.
In order to be able to produce synthetic rains representative of any tropospheric events, a special generator has been designed; it is based on a vibro-rotative disk. This generator is able to produce monodispersed rains at the top of the TOSQAN chamber with rainfall rates from 7 to 15 mm/h and drops diameters from 0.5 to 2.5 mm injected at velocities close to their terminal one.
During these tests, the spectral aerosol concentration is measured in line with the help of a Welas granulometer. This instrument is based on white light scattering.
The results of these experiments highlight the influence of “meteorological” conditions inside the chamber on the washout of the chamber atmosphere, especially when the relative humidity is reaching saturation.
Study on Control Method of Natural Circulation by Injection of Helium Gas
Authors: Nomura Masashi , Takeda Tetsuaki , Funatani Shumpei , Shimura Takuya 
Published: October 2013

This study is to investigate a control method of natural circulation of air by injection of helium gas. A depressurization accident is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomena during the depressurization accident of the VHTR, therefore, it is important to develop the method for prevention of air ingress during the accident.
The experiment has been carried out regarding natural circulation using a circular tube consisting of the loop type or the reverse U-shaped type. The vertical channel consists of the one side heated tube and the other side cooled tube. The experimental results were obtained as follows. When the temperature difference between the vertical tubes was kept at 52K, the velocity of natural circulation flow became about 12cm/s. During this steady state, a small amount of helium injected to the channel. Then, the flow velocity of natural circulation suddenly decreased. The volume of injected helium is about 3% of the total volume of the channel. The velocity became around zero. After 1500 seconds elapsed, the natural circulation suddenly produced again. The experimental results show that the natural circulation flow of air can be controlled by the method of helium gas injection. This paper also discusses an overview of the method for the prevention of air ingress during the primary pipes rupture accident.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration — (5) Measurement of Bubble Deformation Near Wall Under Structure Vibration
Authors: Mizuno Kousuke , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki 
Published: October 2013

In a nuclear power plant, one of the important issues is evaluation of the safety of reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs. However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there are little knowledge how coolant in a core response when large earthquake acceleration is added.
Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction or the power of core is fluctuated with the oscillation by the experiments and numerical analysis. However detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow structure. In order to investigate it, we visualize changing of bubbly flow structure in pipeline on which sine wave is applied.
Bubbly flow is produced with injecting gas into liquid flow through a horizontally circular pipe. In order to vibrate the test section, the oscillating table is used. The frequency of vibration added from the table is from 1.0 Hz to 10 Hz and acceleration is from 0.4 G to 1 G (1 G = 9.8 m/s2). The test section and a high speed video camera are fixed on the table. Thus the relative velocity between the camera and the test section is ignored.
In the visualization experiment, the PIV measurement is conducted. Then the motion of bubbles, for example the shape, the positions and the velocity are measured with observation.
In addition, by varying added oscillation amplitude, frequency and flow rate of the fluids, the correlation between these parameters and bubble motion was evaluated.
It was clarified that the behavior of liquid phase and bubble through horizontal circular pipes was affected by an oscillation. When structure vibration affects the flow, two main mechanisms are supposed. One is the addition of body force of the oscillation acceleration to liquid phase and bubble, and the other is the velocity oscillation of the test section and the effect of the boundary layer of the pipe wall. It was also found that when the added oscillation frequency and amplitude was changed, the degree of the fluctuation of liquid phase and bubble motions were changed.
Sensitivity Analysis of CFD Code FLUENT-12 for Supercritical Water Heat Transfer in Vertical Bare Tubes
Authors: Farah Amjad , Haines Patrick , Harvel Glenn , Pioro Igor 
Published: October 2013

The ability to use FLUENT 12 or other CFD software to accurately model supercritical water flow through various geometries in diabatic conditions is integral to research involving coal-fired power plants as well as Supercritical Water-cooled Reactors (SCWR). The cost and risk associated with constructing supercritical water test loops are far too great to use in a university setting. Previous work has shown that FLUENT 12, specifically realizable k-ε model, can reasonably predict the bulk and wall temperature distributions of externally heated vertical bare tubes for cases with relatively low heat and mass fluxes. However, sizeable errors were observed for other cases, often those which involved large heat fluxes that produce deteriorated heat transfer (DHT) regimes, which involves a reduction in the rate of heat transfer between the wall and the fluid, and therefore a rise in wall temperature. These errors were believed to be caused by FLUENT’s over-estimation of changes in thermal physical properties of the fluid which occur around the pseudocritical point.
The goal of this research is to gain a more complete understanding of how FLUENT 12 models supercritical water cases and where errors can be expected to occur. One control case is selected where expected changes in bulk and wall temperatures occur and they match empirical correlations’ predictions, and the operating parameters are varied individually to gauge their effect on FLUENT’s solution. The model used is the realizable k-ε, and the parameters altered are inlet pressure, mass flux, heat flux, and inlet temperature.
As a result of this work, under supercritical conditions, the only finite limitations found in the k-ε model are in the mass flux which is predicted correctly above 300 kg/m2s, and in the temperature and related physical properties of water where they exceed the limitation of the fluid properties in the fluid database.