bluuebunny/crossref_metadata_2025 huggingface dataset:
Cryogenic Fatigue Strength Assessment for MARK-III Insulation System of LNG Carriers
Authors: Kim Myung Hyun , Kil Yoon Pyo , Lee Jae Myung , Chun Min Sung , Suh Yong Suk , Kim Wha Soo , Noh Byung Jae , Yoon Jang Ho , Kim Min Soo , Urm Hang Sub
Published: April 2011

The objective of this study is to investigate the typical failure mode and to obtain the stress range versus number of cycles to failure (S-N) data of MARK-III type liquefied natural gas (LNG) insulation system under the fatigue loading at actual cryogenic environment. A systematic experimental research is carried out for the assessment of the fatigue strength of MARK-III insulation system at cryogenic temperature. Three different types of test specimens are tested for the evaluation of fatigue performance of MARK-III insulation system. Test specimens are determined considering the fatigue vulnerable locations such as mastic area, slit area, and top bridge pad area inside the actual LNG cargo tanks. All test specimens are fabricated as close as possible to the actual yard practice. A series of fatigue test results is represented as S-N curves. Cyclic fatigue loadings were carefully considered similar to the actual sloshing loads. The effect of sloshing impacts is considered by selecting the stress ratio (R=−10). The load levels have been determined based on the ultimate strength of reinforced polyurethane foam as 12.2 bars. Different cryogenic temperatures are employed according to the test locations in consideration of temperature gradient within the insulation system. All test results including fatigue life, as well as failure locations of MARK-III insulation system at cryogenic temperatures, are reported and compared with those at room temperature. Consistent S-N curves of MARK-III insulation system at both room and cryogenic temperatures are obtained and compared. The slopes of S-N curves from both fatigue test results are observed to be almost identical, and the fatigue strengths are found to exhibit similar trend. The results from this research can be used for the fatigue assessment of the LNGC insulation system, as well as a design guideline of LNG CCS at cryogenic temperature.
Distribution of Wave Height Maxima in Storm Sea States
Authors: Cherneva Z. , Soares C. Guedes , Petrova P.
Published: April 2011

The effect of the coefficient of kurtosis, as a measure of third order nonlinearity, on the distribution of wave height maxima has been investigated. Measurements of the surface elevation during a storm at the North Alwyn platform in the North Sea have been used. The mean number of waves in the series is around 100. The maximum wave statistics have been compared with nonlinear theoretical distributions. It was found that the empirical probability densities of the maximum wave heights describe qualitatively the shift of the distribution modes toward higher values. The tendency for the peak of distribution to diminish with an increase in the coefficient of kurtosis up to 0.6 is also clearly seen. However, the empirical peak remains higher than the theoretically predicted one. The exceedance probability of the maximum wave heights was also estimated from the data and was compared with the theory. For the highest coefficients of kurtosis, estimated at nearly 0.6, the theoretical distribution approximates very well the empirical data. For lower coefficients of kurtosis, the theory tends to overestimate the exceedance probability of the maximum wave heights.
Validation of a New Method for Finding the Rotational Axes of the Knee Using Both Marker-Based Roentgen Stereophotogrammetric Analysis and 3D Video-Based Motion Analysis for Kinematic Measurements
Authors: Roland Michelle , Hull M. L. , Howell S. M.
Published: April 2011

In a previous paper, we reported the virtual axis finder, which is a new method for finding the rotational axes of the knee. The virtual axis finder was validated through simulations that were subject to limitations. Hence, the objective of the present study was to perform a mechanical validation with two measurement modalities: 3D video-based motion analysis and marker-based roentgen stereophotogrammetric analysis (RSA). A two rotational axis mechanism was developed, which simulated internal-external (or longitudinal) and flexion-extension (FE) rotations. The actual axes of rotation were known with respect to motion analysis and RSA markers within ±0.0006 deg and ±0.036 mm and ±0.0001 deg and ±0.016 mm, respectively. The orientation and position root mean squared errors for identifying the longitudinal rotation (LR) and FE axes with video-based motion analysis (0.26 deg, 0.28 m, 0.36 deg, and 0.25 mm, respectively) were smaller than with RSA (1.04 deg, 0.84 mm, 0.82 deg, and 0.32 mm, respectively). The random error or precision in the orientation and position was significantly better (p=0.01 and p=0.02, respectively) in identifying the LR axis with video-based motion analysis (0.23 deg and 0.24 mm) than with RSA (0.95 deg and 0.76 mm). There was no significant difference in the bias errors between measurement modalities. In comparing the mechanical validations to virtual validations, the virtual validations produced comparable errors to those of the mechanical validation. The only significant difference between the errors of the mechanical and virtual validations was the precision in the position of the LR axis while simulating video-based motion analysis (0.24 mm and 0.78 mm, p=0.019). These results indicate that video-based motion analysis with the equipment used in this study is the superior measurement modality for use with the virtual axis finder but both measurement modalities produce satisfactory results. The lack of significant differences between validation techniques suggests that the virtual sensitivity analysis previously performed was appropriately modeled. Thus, the virtual axis finder can be applied with a thorough understanding of its errors in a variety of test conditions.
Numerical Investigation of 2D Optimal Profile of Backward-Bent Duct Type Wave Energy Converter
Authors: Suzuki Masami , Kuboki Toshiari , Nagata Shuichi , Setoguchi Toshiaki
Published: April 2011

A floating type backward-bent duct buoy (BBDB) is a wave energy conversion device with an oscillating water column (OWC) at the front side. The device captures the wave energy using the heaving, the pitching, the surging motion of BBDB, and the heaving motion of OWC. Investigations are carried out to find more reasonable devices than the traditional OWC type floating device. An eigenfunction expansion method is introduced for analyzing the BBDB with OWC. It is confirmed that these solutions give good agreement with several experimental results in this paper. It is shown in a design method how to make BBDB match the turbine characteristics. This feature is being able to select the optimum profile of the turbine and the BBDB individually from each characteristic before comprehensive evaluation of the BBDB and the turbine in the design. After grasping the element characteristics, which are appropriate for the wave energy conversion system, the synthetic design method is built. The BBDB size and the turbine diameter are determined by considering the cost corresponding to the smallest size under the same output. In this way, we can obtain the optimal profiles considering the construction cost including the turbines.
A New Model for Estimating Vibrations Generated in the Defective Rolling Element Bearings
Authors: Behzad Mehdi , Bastami Abbas Rohani , Mba David
Published: April 2011

Prediction of the vibration response due to defects on rolling element bearings requires having an accurate representative vibration model. In this paper, a new model for vibration generation in the rolling element bearings has been introduced. The proposed model assumes a stochastic source of vibration excitation, which is produced as a result of metallic contact between bearing elements during rolling. This model explains high frequency vibration in the acceleration spectrum clearly. When a defect grows in the bearing, the roughness of the contacting surfaces increases locally and stochastic excitation becomes stronger in the defective area. The increased vibration level at the defective area is a good indicator of bearing faults. A numerical simulation of the proposed model was validated with experimental results.
Dynamic Performance of the LeBlanc Balancer for Automatic Washing Machines
Authors: Urbiola-Soto Leonardo , Lopez-Parra Marcelo
Published: April 2011

The paper describes a high-speed camera and a particle image velocimetry (PIV) technique used on a transparent liquid balancing device for washing machines. Experimental results indicate that the baffle-liquid interaction renders fluid modes of vibration of circumferential and axial types. This complex swirl flow is comprised of two inertial waves; one of such waves is synchronous with the rigid body motion, while the other is a fluid backward traveling wave, thus enhancing the system damping capability. This damping phenomenon was revealed by the fluid flow visualization and PIV technique employed.
A Coupled Homotopy-Variational Method and Variational Formulation Applied to Nonlinear Oscillators With and Without Discontinuities
Authors: Akbarzade M. , Langari J. , Ganji D. D.
Published: April 2011

In this paper, two novel and different methods are applied to nonlinear oscillators. It has been found that the coupled method of homotopy perturbation method and variational formulation and amplitude-frequency formulation work very well for the whole range of initial amplitudes. The analytical approximate frequency and the corresponding periodic solution are valid for small as well as large amplitudes of oscillation. Contrary to the conventional methods, only one iteration leads to high accuracy of the solutions. Some examples are given to illustrate the accuracy and effectiveness of these methods.
Application of Algorithms for Percentiles of von Mises Stress From Combined Random Vibration and Static Loadings
Authors: Tibbits Patrick A.
Published: April 2011

Gaussian time-varying loading induces Gaussian components of the stress tensor in a linear structure, where the loading is assumed stationary. For any stress component, finite element spectrum analysis obtains the standard deviation, and any percentile can be calculated as a multiple of the standard deviation. However, a yield criterion requires a percentile of von Mises stress. The distribution of von Mises stress arising from random vibration loading stymies closed-form characterization, but several algorithms estimate its percentiles. One algorithm treats combined random vibration and static loadings. This paper improves computational efficiency for special plane stress cases, e.g., combining finite element spectrum and static analyses of piping models. All the algorithms are applied to a simple test model. Results match Monte Carlo simulation. Computational efficiencies are evaluated and compared.
Dynamic Stability of an Isotropic Metal Foam Cylindrical Shell Subjected to External Pressure and Axial Compression
Authors: Belica Tomasz , Malinowski Marek , Magnucki Krzysztof
Published: April 2011

This paper presents a nonlinear approach with regard to the dynamic stability of an isotropic metal foam circular cylindrical shell subjected to combined loads. The mechanical properties of metal foam vary in the thickness direction. Combinations of external pressure and axial load are taken into account. A nonlinear hypothesis of deformation of a plane cross section is formulated. The system of partial differential equations of motion for a shell is derived on the basis of Hamilton’s principle. The system of equations is analytically solved by Galerkin’s method. Numerical investigations of dynamic stability for the family of cylindrical shells with regard to analytical solution are carried out. Moreover, finite element model analysis is presented, and the results of the numerical calculations are shown in figures.
Dynamic Modeling and Stability Analysis of Flat Belt Drives Using an Elastic/Perfectly Plastic Friction Law
Authors: Kim Dooroo , Leamy Michael J. , Ferri Aldo A.
Published: April 2011

This paper presents an analysis of a nonlinear (piecewise linear) dynamical model governing steady operation of a flat belt drive using a physically motivated elastic/perfectly plastic (EPP) friction law. The EPP law models frictional contact as an elastic spring in series with an ideal Coulomb damper. As such, the friction magnitude depends on the stretch of the elastic belt and is integral to the solution approach. Application of the extended Hamilton’s principle, accounting for nonconservative work due to friction and mass transport at the boundaries, yields a set of piecewise linear equations of motion and accompanying boundary conditions. Equilibrium solutions to the gyroscopic boundary value problem are determined in closed form together with an expression for the minimum value of the EPP spring constant needed to transmit a given torque. Unlike equilibrium solutions obtained from a strict Coulomb law, these solutions omit adhesion zones. This finding may be important for interpreting belt drive test-stand results and the experimentally determined friction coefficients obtained from them. A local stability analysis demonstrates that the nonlinear equilibrium solutions found are stable to local perturbations. The steady dynamical operation of the drive is also studied using an in-house corotational finite element code. Comparisons of the finite-element solutions with those obtained analytically show excellent agreement.
Stiffness Analysis of a 2DOF Planar Tensegrity Mechanism
Authors: Arsenault Marc
Published: April 2011

This paper presents the stiffness analysis of a planar 2DOF tensegrity mechanism. A stiffness model is first derived based on an existing formulation. Several stiffness indices having physical meaning are then extracted from the stiffness matrix for performance evaluation purposes. Stiffness mappings based on these stiffness indices are then plotted over the mechanism’s workspace and observations are made. It is shown, for instance, that in the case of the planar 2DOF tensegrity mechanism, the effect of the prestress on the stiffness is generally not significant when the stiffnesses of the cables and struts are assumed to be linear.
A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core
Authors: Dong Zhe , Huang Xiaojin , Zhang Liangju
Published: April 2011

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.
A Predictor-Corrector Scheme for the Microscopic Depletion Solver of the COCAGNE Core Code
Authors: Hoareau Fabrice , Schwartz Nadine , Couyras David
Published: April 2011

EDF R&D is developing a new calculation scheme based on the transport-Simplified Pn (SPn) approach. The lattice code used is the deterministic code APOLLO2, developed at CEA with the support of EDF and AREVA-NP. The core code is the code COCAGNE, developed at EDF R&D. The latter can take advantage of a microscopic depletion solver expected to improve the treatment of spectral history effects. However, the direct use of the microscopic depletion solver within COCAGNE is computationally very intensive because very small evolution steps (typically 100 MWd/t) are needed to reach a good accuracy, which is not always compatible with industrial applications. In order to reduce the calculation time associated with the use of the microscopic depletion solver, a predictor-corrector scheme has been implemented within COCAGNE. It enables the use of larger evolution steps, up to 1000 MWd/t. Two kinds of tests were performed to validate this predictor-corrector scheme. Firstly, direct comparisons with APOLLO2 results were made on COCAGNE depletion calculations of fuel assemblies. These calculations involved depleting the fuel from 0 GWd/t to 60 GWd/t while keeping fixed thermal-hydraulical conditions and boron concentration, so as to be consistent with the calculation conditions of APOLLO2. It is shown that isotopic concentrations and the multiplication factor (keff) obtained by COCAGNE are consistent with APOLLO2 results. This test was also used to perform a numerical analysis of the convergence order of the new scheme. Secondly, results of the predictor-corrector scheme were compared with those of the existing calculation procedure of COCAGNE on a realistic core calculation case. The impact on global indicators such as boron concentration and cycle length was also considered. This second test indicates that, for industrial applications, the predictor-corrector procedure also gives fairly accurate results while significantly reducing the calculation time related to microscopic depletion.
The Analysis of Effect on the Piping Due to the Displacement of EPR Nuclear Island Buildings
Authors: Wang Junwei , Zhang Zhouhong , Wu Gaofeng
Published: April 2011

The displacements of nuclear island buildings occur because of earthquake, thermal expansion, creep shrinkage etc. Because there are penetrations and some supports that connect to the wall of reactor building, the displacement loads must be considered when the piping and the supports are verified, the displacement loads have great effect on the piping and supports stress analysis. Studying the effect and looking for that how to reduce this effect are helpful to find a way to reduce the piping stress and the loads on the supports, and it is also helpful to make the plant safe. In this paper some piping calculation units of EPR project are showed as examples, the calculation units are done with a new software ANASYS PIPE, and finite element method is used. there are penetrations and supports that connects to the wall of reactor building in the calculation units, the building displacements due to earthquake, thermal expansion and creep shrinkage are considered, and the effect of displacement loads and the combinations of all kinds of loads under different category conditions on the piping and supports are studied, the result shows that the stress of piping and the loads on the supports caused by building displacement loads are prominent, but the piping and the supports can meet the requirement by optimizing the support concept.
The Effect of Fuel Thermal Conductivity Degradation With Burnup on PWR Licensing Limits
Authors: Qi Yusen , Henningson Paul , Strumpell John , Shann Shih-Hsiung
Published: April 2011

Most fuel modeling computer codes of a decade ago did not explicitly account for fuel thermal conductivity degradation with burnup. The overestimation of thermal conductivity is compensated by other code adjustments. Following the approval of the legacy fuel thermal codes, the quantity, quality and comprehensiveness of fuel measurement data have increased dramatically. Recently, the US NRC has informally requested the nuclear industry review the thermal performance codes with the consideration of thermal conductivity degradation. This paper compares the legacy thermal conductivity models with the most recent available models that are published publicly and may have been incorporated into present generation fuel performance codes. Also included in this paper is one method which AREVA NP Inc. developed to evaluate the thermal conductivity degradation effect. The effect of thermal conductivity degradation on the PWR licensing limits, such as centerline fuel melt and strain, is demonstrated.
Floor Response Spectra of Loviisa NPP Reactor Buildings
Authors: Ka¨hko¨nen Jukka , Varpasuo Pentti
Published: April 2011

For the updated seismic probabilistic risk assessment of the Loviisa nuclear power plant in Finland, knowledge of reactor building floor response spectra were needed. For this purpose a large finite element (FE) model of the reactor building was constructed. The model included all the major primary circuit components. The FE -model was parameterized and the Latin hypercube method was utilized to construct sixty samples of reactor building. Parameters used were material properties for concrete and steel, modal damping and the ground motion. The ground motions were derived from the seismic studies conducted for the Loviisa site. Following results were achieved in this analysis: The probabilistic spectra were determined for different elevations in the reactor building as well as for the representative locations in the main mechanical components that were explicitly modeled.
Reactor Experiments Course of the Swiss Nuclear Engineering Master
Authors: Chawla R. , Girardin G. , Joneja O. P.
Published: April 2011

Since September 2008, a new Master’s programme in Nuclear Engineering (NE) is being offered jointly by the two Swiss Federal Institutes of Technology, viz. EPFL at Lausanne and ETHZ at Zurich. The present paper discusses the development and running of one of the key compulsory courses, viz. Reactor Experiments. This is centred mainly around the utilization of EPFL’s teaching reactor, CROCUS. The course is of seven weeks’ duration, with 10 contact hours per week, and is held during the second half of the first semester. Following an introductory session, the class is divided into six groups. With a total of 12 experiments to be carried out by each group, there are usually 2 half-day experiments to be conducted by a given group per week. Effectively, six experiments have to be running simultaneously at any given time. There are two basic types of experiments which constitute the course contents, viz. those related to radiation measurements and neutronics phenomena, and those conducted using the reactor itself. Details are provided concerning the practical realisation of the various experiments. For example, certain difficulties encountered in the context of achieving adequate counting statistics are discussed, and the corresponding solutions presented. The course has been conducted twice to date, and the feedback from the students has been very positive.
Steam Generator and Reactor Coolant Pump Coupled Flow Field Analysis
Authors: Xia Shuan , Weng Minghui , Qiu Jian , Pan Xinxin
Published: April 2011

In the design of some passive PWRs, Reactor Coolant Pump (RCP) is welded directly to the Steam Generator (SG) channel head. This design cancels the support of RCP and simplifies the layout of Reactor coolant system. What’s more, this design also do good to the mitigation of SBLOCA. But this design makes the flow field in the SG channel head and RCP inlet complex and there may be vortex in this flow field for which reason the SG outlet resistance will increase and affect the long-term steady operation. For this issue, some company made tests on it. But the cost of test is high and the applicability of the test result is limited. If the parameters or components size changed a little, the test result will be no longer applicable. To solve this problem, this article considers using 3-D CFD flow field analysis software to analyse the SG and RCP coupled flow field. Through steps of 3D model establishing – meshing in Gambit – analyzing in Fluent, this paper obtains the flow filed condition of SG-RCP coupled part during normal operation and so as to support plant design.
Neutron-Gamma Discrimination Using Zero-Crossing Method for BC454 Boron Loaded Plastic Scintillator
Authors: Wang Dong , He Bin , Zhang Quanhu
Published: April 2011

Boron loaded plastic scintillator could detect both fast neutrons (thanks to hydrogen) and slow neutrons (thanks to 10B). The large cross sections of both reactions lead to high detection efficiency of incident neutrons. However, gamma rays must be rejected first as the scintillator is also sensitive to them. In the present research zero crossing method was used to test neutron-gamma discrimination performance of BC454 boron loaded plastic scintillator. Three contrast experiments were carried out and different thermalization degrees lead to different time spectra in the MCA. Further analysis proved that three Gaussian curves could be used to fit the spectra; they corresponded to gamma rays, fast neutrons and slow neutrons respectively. The slow neutron curve could be clearly separated from the gamma curve. Discrimination performance for fast neutrons became poor, but their peaks could also be separated.
Numerical Investigation of Chemical and Volume Control System Behavior in Normal Operation and Incident Situation
Authors: Wang Zhigang , Wang Xiaojiang , Li Jun , Li Lijuan , Liu Jiang
Published: April 2011

The Chemical and Volume Control System (CVCS) in nuclear pressurized power plant are in charge of coolant pump seal injection, primary loop volume control and plant safety or security. The characteristics of the system in normal operation and incident situation will leave a remarkable impact on the power plant safe operation and security. The numerical verification of CVCS behavior in normal operation shows that the model established in this paper can predict the thermal hydraulic characteristics of the system accurately. CVCS and SIS are numerical investigated with the Computational Software, Flowmaster V7.5, and the models are validated with design data. If SGTR takes place, the flow limits of letdown, charging and coolant pump seal injection will significantly influence the reactor safety. In this accident situation, the safety injection mode of the Reactor Safety Injection System will be switched into the charging mode of the Chemical and Volume Control System, the charging and seal injection maximum flowrates variations with Reactor Coolant System pressure are predicted, and different charging modes are simulated and compared in detail. The comparison results show that one low head safety injection pump of SIS will increase the charging and coolant pump seal injection flow significantly, but two low head safety injection pumps of SIS will not increase these two flows more greatly than the one low head safety injection pump mode.
Development of Real-Time Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor
Authors: Li Lei , Zhang Zhijian
Published: April 2011

A multi-channel model thermal-hydraulic analysis code in real-time for plate type fuel reactor is developed in this paper. In this code, every fuel assembly in reactor is divided into a subchannel. A series of reasonable mathematical and physical model are set up based on the structure and operational characteristics of plate type fuel core. As for the choice of flow friction and heat transfer models, all possible flow regimes which include the laminar flow, transient flow and turbulent flow, and heat transfer regimes which include single liquid phase heat transfer, sub-cooled boiling, saturation boiling, film boiling and single vapor phase heat transfer, are considered. The correlations and constitutive equations used in the code are fit for the rectangular channel. Look-up table method is used to calculate the properties of water and steam. The code has been loaded on the real-time simulation supporting system SimExec. The reactivity insertion accident and loss of flow accident, which has been defined in the IAEA 10MW MTR benchmark program, were calculated by the code in this paper for validation. Furthermore, the steady state of CARR (China Advanced Research Reactor) is analyzed by this code. The detailed flow distribution in each fuel assembly is obtained. The temperature of coolant, quality, void fraction, DNBR in each subchannel is calculated. The results show that the recently developed code can be used for real time thermal hydraulic analysis of plate type fuel reactor.
Development and Validation of an LS-DYNA Model for Simulating Vertical Drop Test of a Nuclear Fuel Shipping Package
Authors: Zhao W. , Hempy B. , Liu J. , Stilwell W. , Rochow R.
Published: April 2011

To meet shipping package safety requirements for transporting fresh nuclear fuel assemblies, structural performance of the shipping package under hypothetical accident conditions must be evaluated and demonstrated to have adequate protection to the fuel assembly it transports. To efficiently evaluate design changes in the shipping package, a simplified finite element model for the shipping package and fuel assembly has been developed using LS-DYNA. The paper describes the development and validation of the finite element model, along with a few design analysis examples to illustrate its usefulness.
Spread Spectrum Time Domain Reflectrometry Applied to Electrical Systems in Nuclear Power Plants
Authors: Sandquist Gary M. , Sandquist Carl J.
Published: April 2011

A recently developed technique “Spread Spectrum Time Domain Reflectometry” (SSTDR), and supporting test devices will be adapted and tested to monitor and diagnose nuclear plant electrical systems. Current time domain reflectometry methods cannot detect or locate small faults after arc fault events, because their impedance discontinuity is too small and transient to create a measurable reflection. However, on-line, unobtrusive SSTDR can detect and locate arc and other electrical faults when the (∼msec) short circuit returns a strong reflected signal. These observations have led to development of SSTDR. If SSTDR can be successfully adapted to present and future nuclear plant electrical systems, it will be possible to monitor, on-line, the integrity of the electrical system continuously and with only minor equipment modification and no consequential safety issues. An integrated circuit (IC) is under development at the University of Utah for applications in the aircraft industry that will be adapted and used for this proposed development.
On-Line Sequestration of Fission Product Gases in Operating Nuclear Reactors
Authors: Sandquist Gary M. , Kunze Jay F.
Published: April 2011

A research program is outlined to theoretically and experimentally establish the effectiveness and acceptability of on-line sequestering of fission product gases (FPGs) as they are generated in a fissile fuel pin using adsorbents and/or molecular sieves selectively distributed within the lower spring-plenum region of the LWR fuel. If successful this innovation can be implemented in existing and next generation LWRs and may: • reduce accident source term (releasable FPGs); • enhance reactor safety and acceptance; • extend fuel life suppressing clad swelling and fuel failure; • reduce thermal neutron poisons in high flux core regions; • improve operational and economical performance; • require only minor changes to fuel pin assembly and use; • use in LWRs with no license or tech-specifications impact.
Applicability Analysis of Generic Reliability Data for PSA on a Specific Reactor
Authors: Ning Jia
Published: April 2011

Probabilistic safety assessment (PSA) on a specific reactor are often implemented without considering the applicability of generic reliability data, and doubt about such assessments is aroused because of the lack of plant-specific reliability data. The applicability of generic reliability data is analyzed in present paper, in order to remove the doubt in a way. Several sets of reliability data composing from different sources are researched. The following analysis evaluate a fault tree for a typical example of a reactor, using several sets of reliability data and show the differences in the results. Additionally, a comparison is made with a procedure of analysis using reliability data ranges. The results show that the probabilistic safety analysis on a specific reactor using reliability data which come from different sources is feasible. The differences are slight for most components, only a few key components should be separated in the first place and concentrated more attention on them. The superiority of plant-specific data should be advocated. In the mean time, the lack of data should not be a barrier for PSA on a specific reactor. And the analysis facing a lack of data is advised to be encouraged as an approach to improve the safety of specific reactors.
DNB Prediction Method Using Local Void Distribution in Rod-Bundle
Authors: Sasakawa Tatsuya , Ikeno Tsutomu , Kataoka Isao
Published: April 2011

A computational multi-fluid dynamics (CMFD) code that predicted the void distribution in sub-channel was coupled with sub-channel analysis code to predict departure from nucleate boiling (DNB). The main assumption was that the void fraction near heated wall was the dominant parameter in DNB. A sub-channel analysis code was used to calculate three dimensional distribution of sub-channel averaged values of mass flux, void fraction, density and quality. These were used as a boundary condition in the CMFD code to predict local void fraction in a subchannel. A bubble diffusion equation was used assuming the wall peak void distribution caused by turbulence. The present method was applied to the analysis of DNB tests. The coupled codes showed a reasonable profile of void fraction in a rod bundle and reproduced DNB heat flux at low void fraction. To investigate this analysis result, the local condition were compared with a DNB flow regime map. This suggested an approach to improve the predictability: the critical void fraction should be modified at low void fraction condition; the bubble diffusion model should be modified to handle the flow regime transition from the isolated nucleation type to churn turbulent flow type.
Thermal Hydraulic Analysis of Severe Accident in PFBR
Authors: Velusamy K. , Chellapandi P. , Raviprasan G. R. , Selvaraj P. , Chetal S. C.
Published: April 2011

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.
Conceptual Structure Design of High Temperature Isolation Valve for High Temperature Gas Cooled Reactor
Authors: Takada Shoji , Abe Kenji , Inagaki Yoshiyuki
Published: April 2011

The high temperature isolation valve (HTIV) is a key component to assure the safety of a high temperature gas cooled reactor (HTGR) connected with a hydrogen production system, that is, protection of radioactive material release from the reactor to the hydrogen production system and combustible gas ingress to the reactor at the accident of fracture of an intermediate heat exchanger and the chemical reactor. The HTIV used in the helium condition over 900 °C, however, has not been made for practical use yet. The conceptual structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed around the seat and a liner because high temperature helium gas over 900 °C flows inside the valve. Inner diameter of the top of seat was set 445 mm based on fabrication experiences of valve makers. A draft overall structure was proposed based on the diameter of seat. The numerical analysis was carried out to estimate temperature distribution and stress of metallic components by using a three-dimensional finite element method code. Numerical results showed that the temperature of the seat was simply decreased from the top around 900 °C to the root, and the thermal stress locally increased at the root of the seat which was connected with the valve box. The stress was lowered below the allowable limit 120 MPa by decreasing thickness of the connecting part and increasing the temperature of valve box to around 350 °C. The stress also increased at the top of the seat. Creep analysis was also carried out to estimate a creep-fatigue damage based on the temperature history of the normal operation and the depressurization accident.
Thermal Hydraulics Behavior Around a Single Heated Rod With Sodium-Water Reaction (1)
Authors: Kudoh Hideyuki , Nishizaki Masanori , Sugiyama Ken-ichiro , Ohshima Hiroyuki
Published: April 2011

The LMFBRs (Liquid Metal Cooled Fast Breading Reactors) is a key technology for the future electricity demand. Therefore, the establishment of safety of LMFBRs is deeply desired. To evaluate the safety of LMFBR steam generators, empirical studies have been performed for the ruptures of heat transfer tube caused by the overheat due to sodium-water reaction in case of a practical scale conditions. But few systematical experiments have been performed for the clarification of phenomena from the viewpoint of thermal hydraulics and physical chemistry. The absence of such studies is derived from the fact that sodium is chemically active and is not feasible for visualized experiments. To evaluate the safety against the secondary failure of heat transfer tubes by an analytical code, which has been tried in JAEA, it is required to understand the heat transfer phenomena around the tube. In this study, we investigated experimentally the thermal hydraulics behavior around a single heated rod with sodium-water reaction as an essential study for the clarification of raptures phenomena of heat transfer tube. The experimental apparatus was consisted of a sodium pool tank, an electrically heated test rod, a gas jet nozzle of argon-water mixture. We set horizontally in the sodium pool the test rod that is heated with the constant heat flux, and to which a gas jet of argon-water mixture was supplied for the sodium-water reaction from its lower side. The temperature of sodium pool was kept to be 420K so that a product of sodium-water reaction, NaOH (melting point is 591K), may exist as a solid phase. Gas jet velocity was set to be 17.3m/s, and the amount of water vapor in the gas mixture was 3% in mass. Just after the introduction of gas mixture, the temperature of sodium pool increased by the heat of chemical reaction. At the same time, the heat transfer to the rod surface decreased rapidly. By the observation after the experiment, it was confirmed that coarse reaction products deposited thickly at the upper side of the rod and finely granular products adhered to the lower ones. Thus, at low-sodium temperature conditions, the products of sodium-water reaction on the rod surface cause the decrease of heat transfer rate between the rod and sodium pool, which depends on the local distribution of deposits. The present authors therefore obtained experimentally the phenomena important for the development of an analytical code of JAEA.
A Review on Specific Features of Small and Medium Sized Nuclear Power Plants
Authors: Khan Salah Ud-din , Peng Minjun , Zubair Muhammad , Wang Shaowu
Published: April 2011

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.
Study on Integrated Method for Constructing Proactive Trouble Prevention Knowledge Base
Authors: Yoshikawa Hidekazu , Yang Ming
Published: April 2011

Integrated method for constructing proactive trouble prevention knowledge base has been studied before starting the practical system development towards various application areas of knowledge base system for failure modes and effect analysis (FMEA), fault tree analysis (FTA), and so forth. In this paper, basic ideas of configuring the structured knowledge based systemization to be used for such purpose are summarized with the following order: (i) Effective re-use of various trouble information, (ii) Method of structuralizing trouble knowledge, (iii) Knowledge on trouble prediction and proactive prevention, (iv) Description of energy, mass and information flow by multilevel flow model, and (v) Method of knowledge base systemization.
CFD Analysis of PWR Core Top Region: Top Fuel Assembly and Top Nozzle Regions
Authors: Wu Chung-Yun , Kao Min-Tsung , Chieng Ching-Chang , Yuan Kun , Xu Yiban , Dzodzo Milorad B. , Conner Michael E. , Beltz Steven A. , Ray Sumit
Published: April 2011

This paper aims to study the pressure distribution and flow patterns in the top fuel region of the AP1000™ reactor using CFD. This study is being performed as part of a CFD evaluation of the flow in the top fuel and upper plenum regions of a PWR reactor vessel. The flow patterns, including cross flows in the top fuel region, are inter-related with the flow distribution and pressure forces in the reactor vessel upper plenum region. Before detailed computations of the flow in the whole top fuel and upper plenum region are performed, conducting local computations for segments of the domain can provide information about physical aspects of the flow as well as mesh sensitivities. The domain of interest in this paper is the top fuel region including the upper part of the fuel assembly (top grid, fuel rods, top nozzle), upper core plate, and core component hold-down device. The commercial CFD computer code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. The complicated geometry of the top fuel region needs to be simplified so that the mesh size for the CFD model of the whole upper plenum and top fuel region does not exceed current software and hardware capabilities. In this study, several different trimmed meshes have been generated to study the effects of the geometries of the hold-down device and the lateral flows. Mesh sensitivity studies have been conducted for each individual part, i.e., the top grid, top nozzle, upper core plate, and hold-down device, in order to determine the proper geometrical simplifications. Pressure drop measurement data are compared with the predicted CFD results and act as a guideline for the mesh selection. These studies support the applicability of the geometrically simplified models and chosen mesh size for the CFD model of the full upper plenum and top fuel regions.
CFD Analysis of PWR Reactor Vessel Upper Plenum Sections: Flow Simulation in Control Rods Guide Tubes
Authors: Kao Min-Tsung , Wu Chung-Yun , Chieng Ching-Chang , Xu Yiban , Yuan Kun , Dzodzo Milorad B. , Conner Michael , Beltz Steven , Ray Sumit , Bissett Teresa
Published: April 2011

The AP1000™ PWR reactor vessel upper plenum contains numerous control rod guide tubes and support columns. Below the upper plenum are the upper core plate and the top core region of the fuel assemblies. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions are performed, conducting local simulations for smaller sections of the domain can provide crucial and detailed physical aspects of the flow. These sub-domain models can also be used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. The study discussed in this paper focuses on the sections of the domain related to the control rod guide tubes. The commercial CFD code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. Two sub-domains are modeled and analyzed: (1) a 1/4 section of one control rod guide tube by itself and (2) a representative unit cell containing two sections of adjacent control rod guide tubes and one 1/4 section of a neighboring support column. For the 1/4 guide tube model (sub-domain 1), trimmed meshes of up to 16 million cells are generated to compute the flow and pressure fields in both complete and simplified (without chamfers and narrow gaps) models. Comparisons of the results lead to the conclusion that the simplified geometry model might be used when developing larger domain models in the future. The representative unit cell (sub-domain 2) is assumed to be positioned in the center of the upper plenum where the global lateral flow effects are minimal. At this position, the lateral flows are generated mainly by the flow as it exits the guide tubes. After flow enters the unit cell from the bottom, there are three potential locations for flow to leave the unit cell: (1) lower locations near the support column and the upper core plate, (2) side windows in the lower portion of the guide tubes, and (3) upper locations near the guide plates positioned inside the guide tubes. Both trimmed and polyhedral meshes are generated as part of the mesh sensitivity studies. Predicted flow rates at each of the outflow locations in conjunction with results from the mesh sensitivity studies provide guidance on (1) what geometry to preserve or remove, (2) what geometry can be simplified to reduce the required mesh, and (3) an estimate of the total mesh required to model the control rod guide tube and support column geometry in the much larger simulation of the entire upper plenum and top fuel domain.