bluuebunny/crossref_metadata_2025 huggingface dataset:
BWR Signals Interdependence Analysis for Transient Events
Authors: Castillo-Dura´n Rogelio , Ortiz-Villafuerte Javier , Calleros-Micheland Gabriel
Published: September 2008

A methodology for the study of the relationship between different reactor signals is presented. The application of the methodology leads to identify dominant frequencies and the contribution from other different signals to the dominant frequencies. The methodology firstly uses a linear model to estimate the response characteristics of the system and the spectra of the noise source. The estimate of the process linear predictor is obtained by the ordinary least squares method. Then, the model performs a multivariate autoregressive analysis, and a relative power contribution index is computed to determine the inter-relationship between the different reactor noise signals. The power contribution index is an indication of how the fluctuation of one variable depends on other variables, at each frequency. In this work we analyze reactor signals acquired during two operational events occurred at Laguna Verde Nuclear Power Plant (LVNPP), the only NPP of MEXICO. These events are: an event where a 12% amplitude power peak occurred and other event where power oscillations occurred in 1995. For the first event, signals from the average power range monitor, driven recirculation flows, position of recirculation flow control valves, and position and recirculation flow controllers, were used for the analysis. For the second transient, signals from the average power range monitor, total flow through the core, pressure drop in the core, jet pump flow and driven recirculation flow driven in both loops are analyzed. The results of the application of the methodology showed the relation between the variables during the transient events. Specifically, for the first transient, a frequency peak of 1.7 Hz was detected on the power spectrum from an APRM signal. In this case the major contribution to this frequency came from the driven recirculation flow at loop B. In the second event, the prominent power oscillation frequency (0.54 Hz) was tracked within the data from different signals. The major contribution to this 0.54 Hz peak came from both jet pump arrays (A and B) flow signals. Each array has 10 jet pumps.
Use of a Best Estimate Power Monitoring Tool to Maximize Power Plant Generation
Authors: Dziuba Lindsey L.
Published: September 2008

The Best Estimate Power Monitor (BEPM) is a tool that was developed to maximize nuclear power plant generation, while ensuring regulatory compliance in the face of venturi fouling, industry ultra-sonic flowmeter issues and other technical challenges. The BEPM uses ASME approved “best estimate” methodology described in PTC 19.1-1985, “Measurement Uncertainty”, Section 3.8, “Weighting Method.” The BEPM method utilizes many different and independent indicators of core thermal power and independently computes the core thermal power (CTP) from each parameter. The uncertainty of each measurement is used to weight the results of the best estimate computation of CTP such that those with lower uncertainties are weighted more heavily in the computed result. The independence of these measurements is used to minimize the uncertainty of the aggregate result, and the overall uncerainty can be much lower than the uncertainties of any of the individual measured parameters. Examples of the Balance of Plant parameters used in the BEPM are turbine first stage pressure, venturi feedwater flow, condensate flow, main steam flow, high pressure turbine exhaust pressure, low pressure turbine inlet pressure, the two highest pressure feedwater heater extraction pressures, and final feedwater temperature. The BEPM typically makes use of installed plant instrumentation that provide data to the plant computer. Therefore, little or no plant modification is required. In order to compute core thermal power from the independent indicators, a set of baseline data is used for comparison. These baseline conditions are taken from a day when confidence in the value of core thermal power is high (i.e., immediately post outage when venturi fouling is not an issue or from a formal tracer test). This provides the reference point on which to base the core thermal power calculations for each of the independent parameters. The BEPM is effective only at the upper end of the power range, where the independent parameters generally vary in a highly predictable way with changes in core thermal power. This paper will present a detailed description of the BEPM methodology, examples of the BEPM output, and examples of field application. Industry applications of the BEPM include monitoring venturi fouling, verification of an ultrasonic flow meter when used as in input to the secondary calorimetric, and monitoring the performance of other plant equipment that can affect core thermal power. When used routinely as part of a thermal performance monitoring program, the BEPM can be extremely effective in generation maximization, identification of equipment degradation or failure, and identification of potential overpower conditions.
Transient Accident Analysis of a Supercritical Carbon Dioxide Brayton Cycle Energy Converter Coupled to an Autonomous Lead-Cooled Fast Reactor
Authors: Moisseytsev Anton , Sienicki James J.
Published: September 2008

The Supercritical Carbon Dioxide (S-CO2) Brayton Cycle is a promising advanced alternative to the Rankine saturated steam cycle and recuperated gas Brayton cycle for the energy converters of specific reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO2 Brayton cycle coupled to an autonomous, natural circulation Lead-Cooled Fast Reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios.
On the Influence of Heating Surface Structure on Bubble Detachment in Sub-Cooled Nucleate Boiling Flows
Authors: Wu Wen , Chen Peipei , Jones Barclay G. , Newell Ty A.
Published: September 2008

This research examines the influence of heating surface structure on bubble detachment, which includes bubble departure and bubble lift-off, under sub-cooled nucleate boiling condition, in order to obtain better understanding to the bubble dynamics on horizontal flat heat exchangers. Refrigerant R-134a is chosen as a simulant fluid due to its merits of having smaller surface tension, reduced latent heat, and lower boiling temperature than water. Experiments were run with varying experimental parameters e.g. pressure, inlet sub-cooled level, and flow rate, etc. High speed digital images at frame rates up to 4000 frames/s were obtained, showing characteristics of bubble movement. Bubble radius and center coordinates were calculated via Canny’s algorithm for edge detection and Fitzgibbon’s algorithm for ellipse fitting. Results were compared against the model proposed by Klausner et al. for prediction of bubble detachment sizes. Good overall agreement was shown, with several minor modifications and suggestions made to the assumptions of the model.
Application of CFD for Modeling Flows in Feed-Water Pipelines
Authors: Dzodzo Milorad B. , Liu Bin , Cioncolini Andrea , Spiegelman Stanley R.
Published: September 2008

A numerical investigation was performed to study the flows in several PWR power plant feed-water pipeline configurations. Computational Fluid Dynamic (CFD) models were developed with commercial CFD software. The models comprised sections of the pipeline networks including geometrical configurations like; elbows, expansions and contractions, valves, T and Y branches. The purpose of the study was to obtain insight into the flow redistribution, flow patterns and axial velocity profiles resulting from the interaction between various pipeline branches and position of the geometrical configurations. The results, streamlines, axial velocity contours, lateral velocity vectors and velocity profiles are presented and discussed in relation to the shape of the pipeline network. The CFD models results indicate that in some cases axial velocity profiles and flow patterns can differ from the developed turbulent flows in the straight pipes. The presented CFD models are one step towards developing advanced tools that can be used to evaluate the effects of the pipeline geometry on the flow structure and axial velocity distribution.
Numerical and Experimental Investigation of Stratified Gas-Liquid Two-Phase Flow in Horizontal Circular Pipes
Authors: Faccini J. L. H. , De Sampaio P. A. B. , Su J.
Published: September 2008

This paper reports numerical and experimental investigation of stratified gas-liquid two-phase flow in horizontal circular pipes. The Reynolds averaged Navier Stokes equations (RANS) with the κ-ω model for a fully developed stratified gas-liquid two-phase flow are solved by using the finite element method. A smooth and horizontal interface surface is assumed without considering the interfacial waves. The continuity of the shear stress across the interface is enforced with the continuity of the velocity being automatically satisfied by the variational formulation. For each given interface position and longitudinal pressure gradient, an inner iteration loop runs to solve the nonlinear equations. The Newton-Raphson scheme is used to solve the transcendental equations by an outer iteration to determine the interface position and pressure gradient for a given pair of volumetric flow rates. The interface position in a 51.2 mm ID circular pipe was measured experimentally by the ultrasonic pulse-echo technique. The numerical results were also compared with experimental results in a 21 mm ID circular pipe reported by Masala [1]. The good agreement between the numerical and experimental results indicates that the κ-ω model can be applied for the numerical simulation of stratified gas-liquid two-phase flow.
Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup
Authors: Marciulescu Cristian , Sung Yixing , Beard Charles L.
Published: September 2008

The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse’s version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an uprated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition.
Downcomer Boiling During LBLOCA with Direct Vessel Injection of APR1400
Authors: Lee Eu Hwak , No Hee Cheon , Lee Dong Won , Lee Chul Hwa
Published: September 2008

An experimental study has been performed to investigate thermal-hydraulic phenomena in the downcomer during LBLOCA with Direct Vessel Injection (DVI), which is a new Safety Injection System (SIS) of the Advanced Power Reactor 1400 MW (APR1400). In order to understand the downcomer boiling phenomena experimentally, the KAIST Downcomer Boiling experiment has been performed with the same as the thickness of the reactor vessel of APR1400 (21 cm) and a height of 150 cm to observe downcomer boiling phenomena and to measure local parameters such as local phase velocities, local void fraction, collapsed water level and heat flux from the heated wall. From the KAIST Downcomer Boiling experiment, we visually observed strong liquid recirculation and vapor jetting near the heated wall due to the axial migration of voids only in the thin layer (about 4 cm) of the heated wall but little bubble migration out of the bubble region. Local phase velocities and local void fraction were measured to estimate the drift-flux parameters in the downcomer channel. Heat flux from the heated side was back-calculated to find the CHF and to estimate nucleate boiling heat transfer coefficients.
Excitation Models for Buffeting of Cylinder Bundles in Parallel Flow
Authors: Pomiˆrleanu Radu
Published: September 2008

In this paper, the effect of several models representing the state of the art in wall-pressure fluctuations statistics were applied to the geometry of square rod bundles with P/D = 1.33, and used together with the random vibration theory to yield the spectral characteristics of rod vibration. Comparison of model predictions with experimental data representing rod acceleration allowed the selection of Efimtsov model as better suited for the rod bundles flow geometry and low frequency range. Further assimilating the mixing-vanes induced swirling flow as a disturbance to the parallel flow, two corrections were proposed based on comparison of measured vibration characteristics of vaned and vaneless bundles.
Estimation of the Alpha Factor Parameters for the Emergency Diesel Generators of Ulchin Unit 3
Authors: Kang Dae Il , Han Sang Hoon
Published: September 2008

Up to the present, the generic values of the Common cause failure (CCF) event parameters have been used in most PSA projects for the Korean NPPs. However, the CCF analysis should be performed with plant specific information to meet Category II of the ASME PRA Standard. Therefore, we estimated the Alpha factor parameters of the emergency diesel generator (EDG) for Ulchin Unit 3 by using the International Common-Cause Failure data Exchange (ICDE) database. The ICDE database provides the member countries with only the information needed for an estimation of the CCF parameters. The Ulchin Unit A3, pressurized water reactor, has two onsite EDGs and one alternate AC (AAC) diesel generator. The onsite EDGs of Unit 3&4 and the AAC are manufactured by the same company, but they are designed differently. The estimation procedure of the Alpha factor used in this study follows the approach of the NUREG/CR-5485. Since we did not find any qualitative difference between the target systems (two EDGs of Ulchin Unit 3) and the original systems (ICDE database), the applicability factor of each CCF event in the ICDE database was assumed to be 1. For the case of three EDGs including the AAC, five CCF events for the EDGs in the ICDE database were identified to be screened out. However, the detailed information for the independent events in the ICDE database is not presented. Thus, we assumed that the applicability factors for the CCF events to be screened out were, to be conservative, 0.5 and those of the other CCF events were 1. The study results show that the estimated Alpha parameters by using the ICDE database are lower than the generic values of the NUREG/CR-5497. The EDG system unavailability of the 1 out of 3 success criterion except for the supporting systems was calculated as 2.76E−3. Compared with the system unavailability estimated by using the data of NUREG/CR-5497, it is decreased by 31.2%.
Krylov Techniques for 3D Problems in Transport Theory
Authors: Panta Pazos Rube´n
Published: September 2008

When solving integro-differential equations by means of numerical methods one has to deal with large systems of linear equations, such as happens in transport theory [10]. Many iterative techniques are now used in Transport Theory in order to solve problems of 2D and 3D dimensions. In this paper, we choose two problems to solve the following transport equation, μ∂ψ∂x(x,μ)+h(x,μ)ψ(x,μ)==∫−11k(x,μ,μ')ψ(x,μ')dμ'+q(x,μ)(1) where x: represents the spatial variable, μ: the cosine of the angle, ψ: the angular flux, h(x, μ): is the collision frequency, k(x, μ, μ'): the scattering kernel, q(x, μ): the source. The aim of this work is the straightforward application of the Krylov spaces technique [2] to the governing equation or to its discretizations derived of the discrete ordinates method (choosing a finite number of directions and then approximating the integral term by means of a proper sum). The equation (1) can be written in functional form as Aψ=q(2) with ψ in the Hilbert space L2 ([0,a] × [-1,1])., and q is the source function. The operator A˜ derived from a discrete ordinates scheme that approximates the operator Aψ=μ∂ψ∂x(x,μ)++h(x,μ)ψ(x,μ)−∫−11k(x,μ,μ')ψ(x,μ')dμ'(3) generates the following subspace K'''(A˜,q)=Span{q,Aq˜,A˜2q,…,A˜m−1q}(4)i.e. the subspace generated by the iterations of order 0, 1, 2, ... , m−1 of the source function q. Two methods are specially outstanding, the Lanczos method to solve the problem given by equation (2) with certain boundary conditions, and the conjugate gradient method to solve the same problem with identical boundary conditions. We discuss and accelerate the basic iterative method [8]. An important conclusion is the generation of these methods to solve linear systems in Hilbert spaces, if verify the convergence conditions, which are outlined in this work. The first problem is a cubic domain with two regions, one with a source near the vertex at the origin and the shield region. In this case, the Cartesian planes (specifically 0<x<L, 0<y<L, 0<z<L) are reflexive boundaries and the rest faces of the cube are vacuum boundaries. The second problem has a geometry with three regions, one of the source, other is the void region and the third one is a shield domain.
Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident
Authors: Rao P. M. , Kasinathan N. , Kannan S. E.
Published: September 2008

The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE–II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m3 RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa.
Shape Optimization Using an Adjoint Variable Method in ITBL Grid Environment
Authors: Shinohara K. , Okuda H. , Ito S. , Nakajima N. , Ida M.
Published: September 2008

To decrease the fluid drag force on the surface of a specified object subjected to an unsteady flow, under a constant volume condition, the adjoint variable method is formulated by using FEM. Based on the Lagrange multiplier method (a conditional variational principle), this method consists of the state equation, the adjoint equation and the sensitivity equation. To solve the equations effectively using the steepest descent method, a parallel algorithm that finds the Armijo’s line-search step size is constructed. The shape optimization code for solving a large scale 3D problem using a parallel algorithm was implemented on ITBL [1,2] using the HPC-MW library [3]. Results show that, by using shape optimization, the fluid drag force on the object can be reduced by about 17.5%.
Molecular Dynamics Analyses on Microscopic Contact Angle: Effect of Wall Atom Configuration
Authors: Ito Takahiro , Hirata Yosuke , Kukita Yutaka
Published: September 2008

Boiling or condensing phenomena of liquid on the solid surface is greatly affected by the wetting condition of the liquid to the solid. Although the contact angle is one of the most important parameter to represent the wetting condition, the behavior of the contact angle is not understood well, especially in the dynamic condition. In this study we made molecular dynamics simulations to investigate the microscopic contact angle behavior under several conditions on the numerical density of the wall atoms. In the analyses, when the number density of the wall is lower, the changing rate of the dynamics contact angles for the variation of ΔV was higher than those for the case where the wall density is higher. This is mainly due to the crystallization of the fluid near the wall and subsequent decrease in the slip between the fluid and the wall. The analyses also show that the static contact angle decreases with increase in the number density of the wall. This was mainly induced by the increase in the number density of the wall itself.
Seismic Response of Reactor Vessel Internals in the IRIS Reactor
Authors: Aquaro D. , Carelli M. D. , Forasassi G. , Lo Frano R. , Zaccari N.
Published: September 2008

The intent of this paper is the presentation and discussion of a methodology for the evaluation and analysis of seismic loads effects on a nuclear power plant. To help in focussing the presented methodology, a preliminary simplified analysis of an integral, medium size next generation PWR reactor structure (IRIS project, an integral configuration PWR under study by an international group) was considered as an application example also for models/codes evaluation. The performed preliminary seismic analysis, even though by no means complete, is intended to evaluate the method of calculating the effects of dynamic loads propagation to the reactor internals for structural design as well as geometrical and functional optimisation purposes. To this goal, finite element method and separated (sub) structures approaches were employed for studying the overall dynamic behaviour of the nuclear reactor vessel. The analysis was set up by means of numerical models, implemented on the MARC FEM code, on the basis of Design Response Spectra as indicated on the relevant rules for Nuclear Power Plants (NRC 1.60) design. The seismic analysis is indented to evaluate the dynamic loads propagated from the ground through the Containment System and Vessel to the Steam Generator’s tubes.
The Benefits of Recycling for the Future of Nuclear Energy
Authors: Hanson Alan
Published: September 2008

Robust increases in energy demand, improvements in the performance of existing nuclear power plants, renewed interest in assuring domestic energy supply and concern about climate change have recently provided powerful arguments for renewing and further expanding the use of nuclear energy in the United States.
Boiling Heat Transfer Under Oscillatory Flow Condition
Authors: Ohtake Hiroyasu , Koizumi Yasuo
Published: September 2008

Onset of nucleate boiling (ONB), point of net vapor generation (NVG) and critical heat flux (CHF) on subcooled flow boiling under oscillatory flow, focusing on liquid velocity, amplitude and frequency of oscillations were investigated experimentally and analytically. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2, 4 and 6 Hz, respectively; amplitude of oscillatory flow was 25 and 50 % in a ratio of main flow rate, respectively. Temperature at ONB and critical heat flux for oscillatory flow were lower than those for steady flow. The decreasing of liquid velocity by oscillatory caused the ONB and the CHF to decrease. The effects of liquid oscillatory flow on the incipient boiling temperature were examined through a classical stability theory of preexisting nuclei. Heat flux at NVG reduced by superimposed oscillatory flow; the increase of liquid velocity by the oscillatory flow caused the NVG to decrease. The CHF under the oscillatory flow condition was examined analytically by using a liquid sub-layer model and a lumped-heat-capacity model in the heater.
Asset Management: A Sustainable Strategy
Authors: Shoemaker Paul
Published: September 2008

New plants are on the horizon! Plans for no less than thirteen (13) Combined Operating Licenses (COL’s) have been announced in 2005 with corresponding submittals to begin by 2007. Several utilities have announced intentions of breaking ground in 2010, meaning that heavy component orders will have to be placed as early as 2007! Industry analysts’ project that 100 new plants will be required in the next 25 years. With an average lead time of 10 years, the bulk of the new plant orders will have to be placed by not later than 2015. This equates to almost 8–9 new plant orders a year!!
LAPUR5.2 BWR Stability Analysis in Chinshan Nuclear Power Plant
Authors: Hsieh Chang-Lung , Lin Hao-Tzu , Chiang Show-Chuyan , Shih Chunkuan , Wang Jong-Rong , Weng Tung-Li
Published: September 2008

Boiling water reactors have the unique coupling mechanisms between neutronic and two-phase flow thermal-hydraulic behaviors and may induce instability by unstable power/flow oscillations. At each core reload design, it is important to employ decay ratio for the purpose of analyzing system stability and determining its operating boundary. Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such analysis. Comparisons made with vendor’s STAIF results indicated good agreements in decay ratios for Chinshan Nuclear Power Plant Unit 2 Cycle 21 reload design. This research focuses on the parametric sensitivity effect on the variation of decay ratio for different power/flow operating points. Based on the result of sensitivity studies, we presented fractional changes of decay ratios by varying certain important parameters under different power/flow points. It is concluded that density reactivity coefficient, gap conductance and recirculation loop gain on high operating power/flow points have larger fractional change of decay ratio.
Insights from a Full Scope Level 2 PSA Model for a PWR Plant With Spherical Shell Containment
Authors: Klu¨gel Jens U. , Xing Michael , Rao Shobha B. , Deremer Ken
Published: September 2008

A full scope Level 2 PSA was performed in 2005 for a three-loop PWR plant with spherical shell containment. New regulatory definition of large early release, and new reporting requirement related to large late release drive the technical approach in this study. New plant features, new accident management guidelines, and new thermal hydraulics and source terms calculations were incorporated in the linked Level 1 and Level 2 model. The plant features observed to be important for accident mitigation include: 1) primary depressurization; 2) accident induced hot leg failure; 3) hydrogen burn mitigation; and 4) containment filtered venting.
Reprocessing of Spent Nuclear Fuel: A Policy Analysis
Authors: Lagus Todd P.
Published: September 2008

Nuclear spent fuel reprocessing has lately reemerged as a subject of debate in the energy policy world. Since a 1977 Presidential Directive which deferred reprocessing of spent nuclear fuel (SNF), the United States has utilized a “once through” or “open cycle” system of nuclear fuel processing, which leaves most of the energy content in uranium unused. Current reprocessing technology increases the cost of nuclear electricity while only offering limited storage benefits. Advanced technologies have the potential to increase proliferation resistance, the need for more geologic repositories, and allow the United States to regain an international presence in reprocessing technology. The United States should not immediately engage in spent fuel reprocessing, but should begin aggressive research and development for new reprocessing technologies.
Sodium Exposure Tests on Limestone Concrete Used as Sacrificial Protection Layer in FBR
Authors: Parida F. C. , Das S. K. , Sharma A. K. , Rao P. M. , Ramesh S. S. , Somayajulu P. A. , Malarvizhi B. , Kasinathan N.
Published: September 2008

Hot sodium coming in contact with structural concrete in case of sodium leak in FBR system cause damage as a result of thermo-chemical attack by burning sodium. In addition, release of free and bound water from concrete leads to generation of hydrogen gas, which is explosive in nature. Hence limestone concrete, as sacrificial layer on the structural concrete in FBR, needs to be qualified. Four concrete blocks of dimension 600mm × 600mm × 300mm with 300mm × 300mm × 150mm cavity were cast and subjected to controlled sodium exposure tests. They have composition of ordinary portland cement, water, fine and coarse aggregate of limestone in the ratio of 1 : 0.58 : 2.547 : 3.817. These blocks were subjected to preliminary inspection by ultrasonic pulse velocity technique and rebound hammer tests. Each block was exposed for 30 minutes to about 12 kg of liquid sodium (∼ 120 mm liquid column) at 550° C in open air, after which sodium was sucked back from the cavity of the concrete block into a sodium tank. On-line temperature monitoring was carried out at strategic locations of sodium pool and concrete block. After removing sodium from the cavity and cleaning the surfaces, rebound hammer testing was carried out on each concrete block at the same locations where data were taken earlier at pre-exposed stage. The statistical analysis of rebound hammer data revealed that one of the concrete block alone has undergone damage to the extent of 16%. The loss of mass occurred for all the four blocks varied from 0.6 to 2.4% due to release of water during the test duration. Chemical analysis of sodium in concrete samples collected from cavity floor of each block helped in generation of depth profiles of sodium monoxide concentration for each block. From this it is concluded that a bulk penetration of sodium up to 30 mm depth has taken place. However it was also observed that at few local spots, sodium penetrated into concrete up to 50 mm. Cylindrical core samples of 50 mm × 150 mm long were obtained from the exposed cavity and tested for compressive strength and longitudinal ultrasonic pulse velocity (UPV). These are compared with core samples obtained from concrete cubes used as standard reference. The average reduction in UPV and compressive strength were 7% and 29% respectively indicating marginal degradation in mechanical properties of sodium-exposed concrete.
The Structural Analysis of the Loviisa Plant Steel Containments for Service Life Extension
Authors: Varpasuo Pentti
Published: September 2008

This paper reports the activities performed in order to prepare the application documents to extend the service life of the Loviisa nuclear power plant by 20 years. Five key components were selected for detailed analysis process concerning the ageing effects, namely: 1) reactor vessel; 2) primary coolant pump; 3) steam generator; 4) pressurizer; 5) containment. This report concentrates on the containment building. The containment consists of the following main parts: 1) hemispherical dome; 2) cylindrical shell with stiffening and anchoring rings; 3) pre-stressed anchoring bolts; 4) penetrations and openings in dome and cylinder; 5) reinforced concrete base slab and reactor pit; 6) steel lining of the base slab and reactor pit; 7) vertical penetrations and openings in the base slab and reactor pit; 8) columns and walls supporting the base slab. The analysis was based on the results of monitoring the aging of the containment parts during the 30 years operating life of the plant and also on the 3D-structural analysis of the containment taking into account the degradation effects. As a result of the analysis the weakest links of the containment were identified and the remaining lifetime was assessed.
Aging and Plant Life Management with the Software Tool COMSY
Authors: Nopper Helmut , Ro¨ßner Roland , Zander Andre´
Published: September 2008

Within the scope of PLEX, a systematic and efficient ageing and plant life management system is becoming more and more important to ensure a safe and economical power plant operation in spite of continuous plant ageing. For the methodical implementation of PLIM & PLEX strategies, AREVA NP has developed the software tool COMSY. This knowledge-based program integrates degradation analysis tools with an inspection data management system. COMSY provides the capability to establish a program guided technical documentation by utilizing a virtual plant model which includes information regarding thermal hydraulic operation, water chemical conditions and materials applied for mechanical components. It provides the option to perform a plant-wide screening for identifying system areas, which are sensitive for degradation mechanisms typically experienced in nuclear power plants (FAC, corrosion fatigue, IGSCC, Pitting, etc.). If a system area is identified as being susceptible to degradation, a detailed analysis function enables the condition-oriented service life evaluation of vessels and piping systems in order to localize and conservatively quantify the effect of degradation. Based on these forecasts with COMSY, specific strategies can be developed to mitigate the effect of degradation and inspection activities can be focused on degradation sensitive areas. In addition, a risk-informed assessment tool serves to optimize inspection activities in respect to degradation potential and the associated damage consequence. After an in-service inspection is performed for a distinct location, the inspection data is to be evaluated according to generally accepted procedures. For this purpose an integrated inspection data management system module provides standardized, interactively operated evaluation functions. The key inspection results are transmitted as feedback in respect to the as-is condition of the component. Subsequently, all further life evaluations of the associated component are calibrated against the inspection results. The compiled condition-oriented knowledge provides the basis for a continuous optimization resulting in tailored inspection and maintenance programs geared to the specific plant. The systematic closed loop process ensures the generation of up-to-date plant documentation relating to the technical as-is status of the plant, as all the data involved in the process are compiled in a “living” documentation structure. The implementation of COMSY in various nuclear power plants has confirmed that systematic plant life management makes good economic sense, as cost reductions can be achieved while increasing the plants availability.
A Comprehensive Evaluation Method of Human-Machine Interface for Indicator-Meters Based on Gray Theory
Authors: Yan Sheng-Yuan , Zhang Zhi-Jian , Peng Min-Jun , Yang Ming
Published: September 2008

Meters, especially indicator-meters are the most widely used visual displaying unit of all displaying devices in nuclear power plant. The quality of human-machine interface (HMI) of Meters directly affects the efficiency of human-machine interaction and the performance of the Human-Machine system. Therefore, it is significant to study an appropriate design evaluation technology for indicator-meters. In this study, the authors propose a “qualitative-to-quantitative” comprehensive evaluation method based on the gray relation analysis which could be used for estimating whether the designed HMI system could accord with the desires of users by analyzing the relation degree of factors of a HMI system using the similarities between the evaluation criterion curve and the design scheme curve. Based on the proposed method, a subjective evaluation system of indicator-meter has been developed and an experiment was carried out for evaluating the performance of the HMI system. This paper firstly gives a brief introduction of gray relation analysis and then presents the key evaluation indexes of indicator-meters. Finally, the application of proposed method is illustrated by an example of indicator-meter evaluation. The experiment results show that the proposed method is suitable to the comprehensive design evaluation of the HMI system characterized by multi-factor interactions, unknown and uncertain information included.
Application of Near Wall Model to Large Eddy Simulation Based on Boundary Layer Approximation
Authors: Takata Takashi , Yamaguchi Akira , Tanaka Masaaki , Ohshima Hiroyuki
Published: September 2008

Turbulent statistics near a structural surface, such as a magnitude of temperature fluctuation and its frequency characteristic, play an important role in damage progression due to thermal stress. A Large Eddy Simulation (LES) has an advantage to obtain the turbulent statistics especially in terms of the frequency characteristic. However, it still needs a great number of computational cells near a wall. In the present paper, a two-layer approach based on boundary layer approximation is extended to an energy equation so that a low computational cost is achieved even in a large-scale LES analysis to obtain the near wall turbulent statistics. The numerical examinations are carried out based on a plane channel flow with constant heat generation. The friction Reynolds numbers (Reτ) of 395 and 10,000 are investigated, while the Prandtl number (Pr) is set to 0.71 in each analysis. It is demonstrated that the present method is cost-effective for a large-scale LES analysis.
SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts
Authors: Duffey R. B. , Pioro I. L. , Gabaraev B. A. , Kuznetsov Yu. N.
Published: September 2008

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33–35% to about 40–45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 °C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems.
Axial Development of Gas-Liquid Two-Phase Flow in Mini-Channels
Authors: Uematsu Junichi , Hirose Yoshinori , Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi
Published: September 2008

Accurate prediction of the interfacial area concentration is essential to successful development of the interfacial transfer terms in the two-fluid model. Mechanistic modeling of the interfacial area concentration entirely relies on accurate local flow measurements over extensive flow conditions and channel geometries. From this point of view, accurate measurements of flow parameters such as void fraction, interfacial area concentration, gas velocity, bubble Sauter mean diameter, and bubble number density were performed by the image processing method at five axial locations in vertical upward bubbly flows using 1.02 and 0.55 mm-diameter pipes. The frictional pressure loss was also measured by a differential pressure cell. In the experiment, the superficial liquid velocity and the void fraction ranged from 0.475 m/s to 4.89 m/s and from 0.980% to 28.6%, respectively. The obtained data give near complete information on the time-averaged local hydrodynamic parameters of two-phase flow. These data can be used for the development of reliable constitutive relations which reflect the true transfer mechanisms in two-phase flow. As the first step to understand the flow characteristics in mini-channels, the applicability of the existing drift-flux model, interfacial area correlation, and frictional pressure correlation was examined by the data obtained in the mini-channels.
Effect of Gravity on Axial Development of Vertical Bubbly Flow
Authors: Abe Kazuya , Hirose Yoshinori , Hazuku Tatsuya , Takamasa Tomoji , Hibiki Takashi
Published: September 2008

In relation to the development of the interfacial area transport equation, axial developments of void fraction profile, bubble number density, interfacial area concentration and Sauter mean diameter of adiabatic nitrogen-water bubbly flows in a 9 mm-diameter pipe were measured by using a Stereo Image-processing Method under normal- and micro-gravity environment. The flow measurements were performed at four axial locations (axial distance from the inlet normalized by the pipe diameter = 5, 20, 40 and 60) under various flow conditions of superficial gas velocity (0.00823–0.0303 m/s) and superficial liquid velocity (0.138–0.915 m/s). The interfacial area transport mechanism under microgravity environment was discussed in detail based on the obtained data and the visual observation. These data can be used for the development of reliable constitutive relations which reflect the rigorous transfer mechanisms in two-phase flow under microgravity environment.
Experimental Study on Contact Angle of Water Droplet in High-Temperature Condition
Authors: Hirose Yoshinori , Hayashi Tsukasa , Hazuku Tatsuya , Takamasa Tomoji
Published: September 2008

In recent decades, a number of reports have pointed out the importance of the influence of wall wettability on gas-liquid two-phase flow with a gas-liquid or liquid-gas phase change. However, very limited experimental data can be obtained about the wall surface wettability in BWRs high-temperature condition, or in the range from 250 to 300°C. From this point of view, we conducted an experimental study of surface wettability of droplet using a pressure vessel, we measured contact angles of water droplet under various ambient pressures (7.7, 14 and 15 MPa), atmospheric gases (nitrogen and argon), and test plate materials (aluminum, stainless 304, Zircaloy). In addition, experimental correlation was derived based on energy balance. The data obtained in this study will provide more accurate assessment of heat transfer in subchannels of BWRs.
New French Regulation for NPPs and Code Consequences
Authors: Faidy Claude
Published: September 2008

On December 2005, the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment (PE). This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Different advantages are in these proposal, it’s more qualitative risk oriented and it’s an important link with non-nuclear industry. Only few components are nuclear specific. But, the general philosophy of the existing Codes (RCC-M [15], KTA [16] or ASME [17]) have to be improved. For foreign Codes, it’s plan to define the differences in the user specifications. In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems. In the same time a new concept has been developed to replace the well known “Leak Before Break methodology”: the “Break Exclusion” methodology. This paper will summarize the key aspects of these different topics.
Heat-and-Electricity-Supply-System for Manned Spacecraft Using Nuclear Power
Authors: Honma Yuji , Narabayashi Tadashi , Shimazu Yoichiro
Published: September 2008

In order to explore the deep space, such as Mars, Jupiter, Saturn, etc in the future, a spacecraft that will be driven by nuclear power should be developed.
THAI Multi-Compartment Containment Test Program
Authors: Kanzleiter T. , Poss G. , Funke F. , Allelein H.-J.
Published: September 2008

The THAI experimental programme includes combined-effect investigations on thermal hydraulics, hydrogen, and fission product (iodine and aerosols) behaviour in LWR containments under severe accident conditions. An overview on the experiments performed up to now and on the future test program is presented, in combination with a selection of typical results to illustrate the versatility of the test facility and the broad variety of topics investigated.
Measurements of Sheath Temperature Profiles in Bruce LVRF Bundles Under Post-Dryout Heat Transfer Conditions in Freon
Authors: Guo Y. , Bullock D. E. , Pioro I. L. , Martin J.
Published: September 2008

An experimental program has been completed to study the behaviour of sheath wall temperatures in the Bruce Power Station Low Void Reactivity Fuel (shortened hereafter to Bruce LVRF) bundles under post-dryout (PDO) heat-transfer conditions. The experiment was conducted with an electrically heated simulator of a string of nine Bruce LVRF bundles, installed in the MR-3 Freon heat transfer loop at the Chalk River Laboratories (CRL), Atomic Energy of Canada Limited (AECL). The loop used Freon R-134a as a coolant to simulate typical flow conditions in CANDU® nuclear power stations. The simulator had an axially uniform heat flux profile. Two radial heat flux profiles were tested: a fresh Bruce LVRF profile and a fresh natural uranium (NU) profile. For a given set of flow conditions, the channel power was set above the critical power to achieve dryout, while heater-element wall temperatures were recorded at various overpower levels using sliding thermocouples. The maximum experimental overpower achieved was 64%. For the conditions tested, the results showed that initial dryout occurred at an inner-ring element at low flows and an outer-ring element facing internal subchannels at high flows. Dry-patches (regions of dryout) spread with increasing channel power; maximum wall temperatures were observed at the downstream end of the simulator, and immediately upstream of the mid-bundle spacer plane. In general, maximum wall temperatures were observed at the outer-ring elements facing the internal subchannels. The maximum water-equivalent temperature obtained in the test, at an overpower level of 64%, was significantly below the acceptable maximum temperature, indicating that the integrity of the Bruce LVRF will be maintained at PDO conditions. Therefore, the Bruce LVRF exhibits good PDO heat transfer performance.