bluuebunny/crossref_metadata_2025 huggingface dataset:
SCWR Rod Bundle Thermal Analysis by a CFD Code
Authors: Sharabi M. , Ambrosini W. , Forgione N. , He S.
Published: July 2009

The present paper describes the results of the application of the FLUENT code in the analysis of rod bundle configurations proposed for high pressure supercritical water reactors. The model considers a 1/8 slice of a rod bundle. The details from CFD calculations offer predictions of the circumferential clad surface temperature and of the effect of axial power distribution on the mass exchange between subchannels and on the maximum surface rod temperature. Geometry and boundary conditions are adopted from a previous work that made use of subchannel programs, allowing for a direct comparison between the two techniques. Both the standard k-ε model and the Reynolds stress transport model are used. Conclusions are drawn about the present capabilities in predicting heat transfer behavior in fuel rod bundles proposed for supercritical water reactors.
Cable Aging Assessment and Condition Monitoring Using Line Resonance Analysis (LIRA)
Authors: Toman Gary J. , Fantoni Paolo F.
Published: July 2009

This paper describes a method for cable system condition monitoring, developed at the IFE Halden Reactor Project, which is based on Frequency Domain Reflectometry. This method resulted in the development of a system called LIRA (LIne Resonance Analysis), which can be used on-line to detect local or global changes in the cable electrical parameters as a consequence of insulation faults or degradation. This paper presents some results achieved in field experiments on signal cables for nuclear installations, in USA and Europe.
Water Hammer Characteristics for Parallel Pumps Water Supply Systems
Authors: Tian Wenxi , Su Guanghui , Qiu Suizheng , Wang Gaopeng , Lu Qing
Published: July 2009

The water hammer induced by abrupt velocity change of fluid flow is inevitable for nuclear power plant systems because of the sudden opening or closing of valves, the sudden startup or shutdown of the pumps and the rupture of pipes. The water hammer pressure wave can damage the pipes and cause the abnormal shutdown of Nuclear Power Plant (NPP). The object of this study is a Parallel Pumps Water Supply system (PPWS) adopted in a NPP. The PPWS is composed of two parallel mixed-flow pumps connected with a check valve separately, a container, a throttle flap and pipe lines. The Method of Characteristic line (MOC) was adopted to evaluate the water hammer behaviors of the PPWS during the alternate startup and shutoff conditions of two parallel pumps. A code was developed using Fortran language to compute the transient behaviors including he peak pressure, the flow velocity and the movement of the valve plate. The results indicate that the water hammer behaviors under low speed startup condition differ from that of high speed startup condition. The maximum pressure vibration amplitude is up to 5.0MPa occurring under high-high speed startup condition. The computation results are instructive for the optimization design of the PPWS so as to minimize the damage potential induced by water hammer.
Experimental Investigation of Pressure Drop of Flow Boiling at Vertical Narrow Rectangular Channel
Authors: Pan Liang-Ming , Chen De-Qi , Yuan De-Wen
Published: July 2009

Narrow channel heat transfer element has been extensive adopted in engineering applications, especially at electronics technology, this kind of elements often be used to construct compact heat exchanger. Pressure drop of flow boiling at vertical channel with gaps of 1.7, 2.2 and 3.6 mm was experimentally investigated in this paper. The variation of the two-phase frictional multiplier vs. heat flux at various operating conditions was gotten experimentally, possible mechanism of the two-phase frictional multiplier trends of narrow channel were analyzed. Experimental results revealed that the two-phase frictional multiplier increased at lower flow rate and heat flux, as well as higher vapor quality, and dropped at wider flow gap. The multiplier can not be estimated by commonly used method for ordinary gap, thus a modified model of pressure drop for narrow channel was proposed considering the size effects of channel. The error of the predicted two-phase frictional multiplier is within ±15.4% compared with experimental results.
Supercritical Water Heat Transfer in a Vertical Bare Tube
Authors: Gospodinov Yevgeniy , Mokry Sarah , Kirillov Pavel , Pioro Igor
Published: July 2009

This paper presents selected results on heat transfer to supercritical water flowing upward in a 4-m-long vertical bare tube. Supercritical water heat-transfer data were obtained at pressures of about 24 MPa, mass fluxes of 200 – 1500 kg/m2s, heat fluxes up to 884 kW/m2 and inlet temperatures from 320 to 350°C for several combinations of wall and bulk-fluid temperatures that were below, at or above the pseudocritical temperature. In general, the experiments confirmed that there are three heat-transfer regimes for forced convective heat transfer to water flowing inside tubes at supercritical pressures: (1) normal heat-transfer regime characterized in general with heat transfer coefficients (HTCs) similar to those of subcritical convective heat transfer far from critical or pseudocritical regions, which are calculated according to the Dittus-Boelter type correlations; (2) deteriorated heat-transfer regime with lower values of the HTC and hence higher values of wall temperature within some part of a test section compared to those of the normal heat-transfer regime; and (3) improved heat-transfer regime with higher values of the HTC and hence lower values of wall temperature within some part of a test section compared to those of normal heat-transfer regime. These new heat-transfer data are applicable as a reference dataset for future comparison with supercritical-water bundle data and for a verification of scaling parameters between water and modeling fluids. Also, these HTC data were compared to those calculated with the original Dittus-Boelter and Bishop et al. correlations. The comparison showed that the Bishop et al. correlation, which uses the cross-section average Prandtl number, represents HTC profiles more correctly along the heated length of the tube than the Dittus-Boelter correlation. In general, the Bishop et al. correlation shows a good agreement with the experimental HTCs outside the pseudocritical region, however, overpredicts the experimental HTCs within the pseudocritical region. The Dittus-Boelter correlation can also predict the experimental HTCs outside the pseudocritical region, but deviates significantly from the experimental data within the pseudocritical region. It should be noted that both these correlations cannot be used for a prediction of HTCs within the deteriorated heat-transfer regime.
SERB Devices for the Isolation of Buildings and Equipment Against Shocks, Vibrations and Seismic Movements
Authors: Serban Viorel , Panait Adrian , Androne Marian
Published: July 2009

The paper presents new types of devices to isolate buildings and equipment against shocks, vibrations and seismic movements, developed by SITON in cooperation with other units in Romania. The new devices, called SERB devices, can overtake very large static loads (over 3000KN) over which dynamic loads are overlapping and damped. The devices allow the “cutting-off” of the dynamic action coming from the environment to the isolated component (building or equipment) and the shocks and vibrations generated by the component to the environment. The relative displacements allowed by the devices in horizontal plane, reach up to ±200mm and their capacity to dissipate the energy is very high. The devices can overtake compression and tension loads up to half of the maximum compression load. The devices are made of materials that practically, are not affected by the aging phenomenon and they may be guaranteed for the entire service-life of a nuclear power plant, without maintenance and repair. The sizes of the devices are small if compared with the sizes of the existing devices and the performances are similar. The radiation resistance capacity is very good allowing thus their use in all the areas of a nuclear-power plant.
The Numerical Calculation of Two-Phase Pressure Drop During Flow Boiling in the Vertical Annular Channel
Authors: Wang Ruosu , Peng Changhong , Zhou Tao , Wang Shichao
Published: July 2009

In this paper, according to a separated phase flow model for annular two-phase flow in the vertical annular channel, the liquid film thickness, void fraction and two-phase pressure drop along the flow channel are predicted. The calculated pressure drop values are compared with that measured in experiment in the range of mass flux from 38.8 to 163.1kg/m2s; pressure from 1.5 to 6.0 MPa; and heat flux from 4.9 to 50.7kW/m2 for inside tube and from 4.2 to 78.8kW/m2 for outside tube. It can be found that the model can predict the pressure drop well. With the mass flux increasing and the gap of the annular channel decreasing, pressure drop increases.
Human Factors Verification and Validation: Tecnatom’s Experience
Authors: Alonso Pedro Trueba , Illobre Lui´s Ferna´ndez , Ferna´ndez-Sesma Alfonso Jime´nez , Pascual Fernando Ortega
Published: July 2009

Tecnatom has been performing Human Factors Engineering Verification and Validation (HFE V&V) from the mid-eighties. This activity started as one of the various activities of the post TMI requirements followed also in Spain: performing detailed Control Room Design Reviews (DCRDRs). All the existing Spanish Nuclear Power Plants (NPPs) were reviewed to identify Human Engineering Discrepancies (HEDs). DCRDRs were completed by the mid-nineties, and the following V&V activities have been related to new designs and plant modifications as part of the activities described in the Human Factors Program Review Model (HFE PRM), included in NUREG-0711 since 1994. The NRC recommends following the HFE PRM or an acceptable alternative method in the case of the HFE activities and the Spanish Regulatory Body (CSN) recommend the same approach for new designs, design modifications and even for conventional plants. The activities embedded in a HFE V&V process are Task Support Verification (TSV), HFE Design Verification (HFE DV) using NUREG-0700 HFE Guidelines, and Integrated System Validation (ISV), with the execution of performance based tests, mainly in simulator facilities. This paper describes some of the experience of Tecnatom during the past years regarding the execution of these V&V activities previously mentioned, and in relation to the applicability and methodological aspects of each of these activities. Methodological aspects regarding TSV are related to its execution when there is a no Task Analysis to use. Methodological aspects regarding the HFE DV are related to the type of HSI to verify (small or large), its development status (paper design or implemented), the selection and translation of applicable HFE guidelines, and the HED preparation. Methodological aspects regarding the ISV are related to the necessary crews, training, number of scenarios, issues to test, data collection and performance measures. The experience is mainly related to Tecnatom’s work is Spanish NPP like Jose´ Cabrera, Almaraz and Vandello´s and in the case of foreign plants Beznau and an advanced NPP in Taiwan amongst others.
Efforts Toward Risk Informed Regulation and Immediate Issues in Japan
Authors: Hidaka Akihide
Published: July 2009

Based on the basic policy for introduction of Risk Informed Regulation (RIR) expressed by the Nuclear Safety Commission (NSC) of Japan in 2003, NSC set up the taskforce in April 2004 to take an initiative for coordinated collaboration of related organizations for developing scheme of RIR. Since then the taskforce has reviewed the efforts of related organizations and discussed the issues on further utilization of risk information in Japan. According to the final report prepared by the taskforce in September 2007, the risk consideration in related organizations has made a progress in line mostly with the NSC’s basic policy. For example, the regulatory guide for seismic design was revised in 2006 including combination of deterministic and probabilistic approaches. The regulatory body will start a new inspection system in 2008 that considers the risk informed safety classification of structures, systems and components. However, the followings were identified as important issues in the future: i) promotion of advanced and preliminary trials, ii) comprehensive applications using PSA results for both of the representative and individual plants, iii) preparation of acceptance guidelines for safety and performance goals in risk informed decision making, iv) improvement of PSA quality, v) revision of regulatory guides considering risk information, vi) comprehensive evaluation and promotion by NSC for further utilization of risk information, vii) enhancement of infrastructure such as PSA experts and database, viii) promotion of safety research, ix) application to seismic design, x) Introduction of risk informed approaches into nuclear fuel cycle facilities, and xi) promotion of risk communication, etc.
Validation of RELAP5/3D Steam-Generator Analysis Model by Turbine Trip Test at the Prototype Fast Breeder Reactor MONJU
Authors: Shindo Yoshihisa , Endo Hiroshi , Ishizu Tomoko , Haga Kazuo
Published: July 2009

In order to develop a thermal-hydraulic analysis model of the SG (steam-generator) to simulate transient phenomena in the FBR (sodium cooled fast breeder reactor) MONJU, JNES (Japan Nuclear Energy Safety Organization) validated a SG analysis model using the RELAP5/3D code against the results of turbine trip test at a 40% power load. The modeling by using the code was performed to explain main observed behaviors of the pressure and the temperature of the EV (evaporator) steam outlet, and the temperature of water supply distributing piping (near the water supply chamber) until 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results on parameter study of the blow efficiency (release coefficient.). It was found that RELAP5/3D with a two-fluid model of water/steam and a sodium fluid model for tube outside predicted well the physical behaviors: the void of steam generated by the depressurization boiling moves upward in the down-comer tubes accompanied by the enthalpy, increases the temperature of the water supply chambers; and that the pressure change of a “shoulder” like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/3D to the SG modeling was confirmed by the well simulation of the actual FBR system.
Heat Transfer Studies on Lead-Bismuth Eutectic Flows in Circular Tubes
Authors: Borgohain A. , Maheshwari N. K. , Vijayan P. K. , Saha D. , Sinha R. K.
Published: July 2009

The use of accurate heat transfer model in liquid metal like Lead Bismuth Eutectic (LBE) flow is essential for the designing of the liquid metal cooled nuclear reactor systems. In the present study, the existing physical correlations for heat transfer in LBE flow through circular tube have been reviewed and assessed with the experimental results. In CFD analysis, PHOENICS-3.6 is used to carry out the evaluation of the various turbulence models in the tube geometry and to identify the difference between the numerical results and experimental ones in LBE flows. Based on the assessment of the existing correlations for heat transfer in LBE flow and the CFD results achieved, the best-suited correlation for turbulent Prandtl number is recommended in terms of Peclet number. This Prt can be incorporated in PHOENICS for LBE flow analysis.
Design Study of a Pilot Test Plant for Hydrogen Production by a Hybrid Thermochemical Process
Authors: Rivalier Patrick , Charton Sophie , Ode Denis , Duhamet Jean , Boisset Laurence , Pabion Jean-Louis , Gandi Florent , Croze Jean-Pierre
Published: July 2009

Mass production of hydrogen is a major issue for the coming decades, particularly for greenhouse gas lowering purpose. Thanks to fourth-generation nuclear reactors development, providing high-temperature energy, hydrogen production processes are currently revisited worldwide. In France, the CEA is particularly investigating the Westinghouse hybrid sulfur cycle. This cycle, which combines an electrolysis step (oxidization of SO2 to H2SO4 from a liquid-phase anode stream) and a thermochemical step (decomposition of H2SO4), was developed in the 1970s by Westinghouse Electric Corporation. Its production capacity was demonstrated at a scale of 150 NL·h−1 of hydrogen and a conceptual plant, designed to produce 90 Nm3·s−1, was developed in 1977. This plant was however never built, probably for both economic and technological reasons. Since 2000, the Westinghouse process is the subject of renewed interest. Indeed, the decomposition of sulfuric acid has been investigated in detail (many thermochemical cycles are based on sulfur) and it is now optimized from a thermodynamic standpoint. Moreover, the substantial research efforts devoted to fuel cells have led to the development of new materials and structures that are expected to yield significant performance improvements when applied to classical electrochemical processes. Therefore, the CEA has decided to build a pilot electrolysis facility with an hydrogen capacity of 100 NL·h−1, allowing implementation of the most promising material technologies.
Modelling of Pressure Fluctuation With Cross Flow in a Tight-Lattice Rod Bundle
Authors: Zhang Weizhong , Yoshida Hiroyuki , Takase Kazuyuki
Published: July 2009

An approximate model is presented which permits the prediction in detail of the unsteady differential pressure fluctuation behavior between subchannels in the nuclear reactor core. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature slug flow in each subchannel. The model is based on the detailed numerical simulation result of two-phase flow that pressure drop occurs mainly in liquid slug region and in the bubble region it is negligibly small. The instantaneous fluctuation of differential pressure between the two subchannels is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This model temporarily used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels.
First Step in the Investigation of Hydrogen Recombiners Efficiency According to Their Location in Nuclear Power Plants
Authors: Meynet N. , Bentaib A. , Bleyer A. , Caroli C.
Published: July 2009

Passive Autocatalytic Recombiners (PAR) are used as a mitigation measure for hydrogen risk in pressurized water reactors (PWR). They aim to remove hydrogen by its exothermic catalytic recombination into steam, and to maintain local hydrogen concentration below deflagration-detonation transition limit. This study deals with PAR efficiency for removing hydrogen and mixing containment atmosphere, according to their location inside the containment. The competition of physical phenomena as free and forced convection and condensation is considered on the basis of simplified geometry and scenarios.
Selecting HLR and SR Standards for PRA Applications
Authors: Farkas Steven E.
Published: July 2009

This white paper provides guidance for mapping risk informed applications to PRA Scope and Technical Adequacy requirements. The discussion considers regulatory guidance contained in SRP 19.1 “Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities” and supplements industry guidance on application mapping contained in Section 3 “Risk Assessment Application Process of the ASME Standard for the PRA for Nuclear Power Plant Applications,” of the ASME PRA Standard with Regulatory Guide 1.200 clarifications. The general processes identified in this paper should also be applicable to mapping applications to any of the forthcoming ANS standard requirements. The aim of this white paper is to help identify which ASME PRA Standard High Level Requirements (HLRs) fall in the scope of a particular “risk informed” application to the NRC that will, in part, justify the requested change with risk metrics. To obtain values for risk metrics, it is necessary to manipulate the plant PRA model. The question becomes what should be changed in the model to represent the “risk informed” topic at hand? Each HLR has a set of supporting requirements (SRs) graded by “Capability Categories.” For many supporting requirements (SRs) under each HLR, there is no difference between the three capability categories. At issue is what features of the model have to conform to Capability Category 2 when Capability Category 2 differs from Capability Category 1.
Improvement of the CATHARE 3D Code Prediction on PIERO Transient
Authors: Glantz Tony , Freitas Roberto
Published: July 2009

The PIERO experiment has been carried out to study phase’s separation in the lower plenum and the downcomer of a Pressurized Water Reactor (PWR) during the end of the depressurization of a large break loss of coolant accident (LB-LOCA). This experiment has been used for the validation and assessment of the 3D module of CATHARE code [1] but the results are not good because of an overestimation of the liquid entrainment in the lower plenum in one hand and the use of a coarsed meshing for modelling the PIERO experiment in the other hand. Two ways of improvement are possible: the first one and the most complicated is to introduce a stratification model in the 3D module of CATHARE. The other one is a possibility to use a refining meshing in order to simulate PIERO experiment. This second way has been performed and the computations results are greatly improve. Nevertheless, PIERO experiment is not on a reactor scale and a direct application of the meshing recommendations made on PIERO is impossible to translate directly on the reactor case. So, the strategy of validation applied to the reactor case consisted in reproducing a PIERO transient with a full scale lower plenum in a first step. In a second step, a converged meshing for the full scale modelling has been determined. In a last step, results obtained with this kind of modelling have been validated via two correlations developed by Wallis and al., that define boundaries conditions between which the water level remaining in the lower head is allowed to vary. This strategy of validation led to model the reactor’s lower plenum with the more axial meshes in order to have good results.
Numerical Study on Turbulent Mixed Convection in Plane Channel
Authors: Xie Z. R. , Yang Y. H.
Published: July 2009

Numerical study on turbulent mixed convection in plane channel is carried out with low-Reynolds number k-ε model coupled with radiation model. Results are compared with experimental data in open literatures. Traditional LS low-Reynolds model and advanced low-Reynolds model SZS low-Reynolds model have been used to compare with experimental data. Results show that there is no general privilege of SZS model over LS model. Although SZS gives better results of mean velocity, it gives too large Nusselt number due to the too larger budget of turbulent shear stress. It is also shown from numerical study that radiation plays a great part especially under the great buoyancy influenced condition.
Flow-Accelerated Corrosion Susceptibility Prediction of Recirculating Steam Generator Internals
Authors: Pietralik John M. , Heppner Kevin L.
Published: July 2009

Steam generator (SG) components are subjected to corrosive solutions in turbulent flow. Under such conditions, actual component lifetimes may be significantly reduced from their original design lifetimes. Premature replacement of steam generator components before their expected lifetime can be very expensive. Furthermore, degradation of essential components can reduce the steam generator efficiency, thus reducing net profits. Moreover, a SG failure can also be a safety issue. One of the degradation mechanisms affecting secondary-side SG internal structural elements, which are referred to as internals, is Flow-Accelerated Corrosion (FAC). The susceptibility to FAC depends on flow parameters, water chemistry, and materials. All SG internals made of carbon steel are susceptible to FAC to varying degrees. For FAC susceptibility prediction, flow velocity, pH, and oxygen distributions are needed. SG codes, including THIRST (Thermal Hydraulic analysis In STeam generators, a computer code developed by AECL), traditionally solve for thermalhydraulic parameters. A new chemistry module has been added to THIRST, which now makes this code useful for the prediction of local water chemistry parameters in the SG. The THIRST chemistry module is comprised of a multicomponent, multiphase mass transport model coupled with a multiphase chemical equilibrium model. As input, the module requires amine concentrations in the feedwater and reheater drains. The module predicts local distributions of amine concentration in the secondary side. The concentrations predicted by the module are used to compute the pH. The chemistry module was verified against results of other work in the literature and against station blowdown data. Flow and chemistry predictions of THIRST were used to predict FAC susceptibility for internals of a SG with an integral preheater and a SG without it. Ranking of SG locations in order of FAC susceptibility was estimated from an empirical, Kastner-Riedle model. The most susceptible internals are predicted to be those in the upper section of the hot side and those on the cold side that are near the SG centre, while SG lower regions, including the integral preheater, if one exists, are better protected.
Research on the Control System of Thorium-Based Long-Life Core
Authors: Yu Ganglin , Wang Kan
Published: July 2009

In the thorium-based long-life reactor, the thorium-spent fuel plutonium mixed fuel and lead-bismuth coolant are used to design a fast neutron spectrum core. This paper studies the control system in the long-life core and gets some useful conclusion. The control system includes two parts: the central control rod and a peripheral control unit that surrounds the core. The paper studies the thickness of the absorber, which influences the neutron absorb capability and economy, and gives the relation between the thickness and the reactivity. A peripheral reflector is used in most small fast neutron spectrum critical equipment to control the reactivity. The paper discusses the absorber mode and the reflector mode in peripheral control unit design. The paper gives the final design details of the control system in the thorium-based long-life core reactor. The control system has high control efficiency and achieves the requirement of the long-life core.
Effect of Nature Convection on Heat Transfer in the Liquid LiPb Blanket
Authors: Wang Hongyan , Zhang Xidong
Published: July 2009

In some liquid blankets (or liquid blanket module) of fusion reactor, the liquid metal, i.e. LiPb flow, as only tritium breeder is characterized by lower outlet temperature and slower flow velocity that allows the utilization of relatively mature material technology [1–2]. The magnetohydrodynamic (MHD) flow and heat transfer become very complex resulting from the differential heating of walls of the channels, especially adjacent to the First Wall (FW), and internal heat sources inside of the liquid LiPb. The nature convection of the liquid LiPb, due to thermal diffusion, in the poloidal channel adjacent to the FW in the presence of the strong magnetic field of the blanket has been considered and studied. The temperature distribution is changed and there is a strong thermal coupling, modifying importantly the magnitude of the flow. The effect of the buoyancy on pressure driven duct flows has been investigated. The buoyant convection was found to be sufficiently strong to impose its flow pattern on the cross flow in the region of intense volumetric heating.
Transient 3D Stability Analysis of SCWR Rod Bundle Subchannels by a CFD Code
Authors: Sharabi M. , Ambrosini W. , He S. , Jiang Pei-Xue , Zhao Chen-Ru
Published: July 2009

The paper describes the application of CFD in simulating density wave oscillations in triangular and square pitch rod bundles. The FLUENT code is used for this purpose, addressing typical conditions proposed for SCWR conceptual design. The RELAP5 code and an in-house 1D linear stability code are also adopted to compare the results for instability thresholds obtained by different techniques. Transient analyses are performed both by the CFD code and RELAP5, with increasing heating rates and constant pressure drop across the channel, up to the occurrence of unstable behaviour. The obtained results confirm that the density wave mechanism is similar in rod bundle and in axisymmetric configurations.
Numerical Simulation of Boiling Two-Phase Flow in Tight-Lattice Rod Bundle by 3-Dimensional Two-Fluid Model Code ACE-3D
Authors: Yoshida Hiroyuki , Misawa Takeharu , Takase Kazuyuki
Published: July 2009

Two-fluid model can simulate two phase flow less computational cost than inter-face tracking method and particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system in order to simulate complex shape channel flow. In this paper, boiling two-phase flow analysis in a tight lattice rod bundle is performed by ACE-3D code. The parallel computation using 126CPUs is applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. At height z = 0.5 m, void fraction in the gap region is higher in comparison with that in center region of the subchannel. However, at height of z = 1.1m, higher void fraction distribution exists in center region of the subchannel in comparison with the gap region. The tendency of void fraction to concentrate in the gap region at vicinity of boiling starting point, and to move into subchannel as water goes through rod bundle, is qualitatively agreement with the measurement results by neutron radiography. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight lattice rod bundle with no lift force model (neglecting lift force acting on bubbles) is also performed. From the comparison of numerical results, it is concluded that the effects of lift force model are not so large on overall void fraction distribution in tight lattice rod bundle. However, higher void fraction distribution in center region of the subchannel was not observed in this simulation. It is concluded that the lift force model is important for local void fraction distribution in rod bundles.
The Rulemaking Process and Its Effect on the Use of ASME Nuclear Codes and Standards in the USA
Authors: Ferlisi Mark J. , West Raymond A. , Lewis David R.
Published: July 2009

The ASME Board on Nuclear Codes & Standards (BNCS) has a standing Task Group on Regulatory Endorsement (TG-RE) that is actively involved with obtaining endorsement of ASME Nuclear Codes and Standards (NC&S) used in the USA. However, all ASME NC&S that are required to be met by law in the USA must first be endorsed in the Code of Federal Regulations (CFR), specifically in Title 10, Section 50.55a, “Codes and standards.” Similarly, all Code Cases permitted by law to be used, as alternatives, must be approved for use in other United States Nuclear Regulatory Commission (NRC) documents such as NRC Regulatory Guides endorsed in the NRC rules. The NRC uses a rulemaking process to amend Section 50.55a to endorse and approve new or updated NC&S, and thus controls and mandates the use of ASME NC&S in the USA. The rulemaking process is very structured and is not very flexible. This paper describes how the TG-RE is working to improve and expedite the process by which the NRC endorses our ASME NC&S. Because the NRC influences international regulatory action, this paper is written to inform the international nuclear engineering community about the process by which endorsement of NC&S occurs in the USA and what actions are being taken to improve that process.
Autoregressive Multivariate Analysis of BWR Bistable Flow
Authors: Castillo-Dura´n Rogelio , Ortiz-Villafuerte Javier , Go´mez-Herrera Raymundo , Calleros-Micheland Gabriel
Published: July 2009

Bi-stable flow patterns can induce flow oscillations possibly leading to power fluctuations. In other cases, bi-stable flow can generate loads on components, so that structural stresses may become a potential cause of failure. This type of flow occurs in BWRs at different operation conditions, so there is no an absolute methodology for detection and prediction of such phenomenon. In this work, a multivariate autoregressive (MAR) analysis is performed to different signals related to a bi-stable flow event that occurred in one of the BWR Units at the Laguna Verde Nuclear Power Plant. The signal analysis was performed with the home-developed NOISE computer program, which, among several other applications, computes the autoregressive coefficients which contain the information of the dynamics of the signal, and that later are used to determine the relative power contribution (RPC) ratio, which in turn allows establishing the influence of the different signals on each other. From the signal analysis, among the important results obtained, it was found that no new frequencies appeared during the event. Also, it was determined through the Relative Power Contribution ratios that the most probable cause of reactor power change was the flow variation in the recirculation flow of loop B. Maximum variations (both above and below) from the initial average reactor power were 0.5%, so the bi-stable flow impact was of no safety concern.
PTV Experiments of Subcooled Boiling Flow Through a Rectangular Channel
Authors: Estrada-Perez Carlos E. , Dominguez-Ontiveros Elvis E. , Ahn Hee Seok , Amini Noushin , Hassan Yassin A.
Published: July 2009

Experiments were carried out to investigate turbulent sub-cooled boiling flow of Novec-2000 [1] refrigerant through a vertical square channel with one heated wall. Channel dimensions were selected to be similar to those encountered on a Boiling Water Reactor (BWR) channel flow, with an hydraulic diameter of Dh = 8.2mm. Flow visualization techniques such as Particle Tracking Velocimetry (PTV) and Shadowgraphy were used to measure time-average axial and normal velocities, axial and normal turbulence intensities, and Reynolds Stresses. Results are reported for hydraulic Reynolds numbers at channel inlet of 4638, 14513 and 24188 for up to thirteen wall heat fluxes (q″) ranging from 0.0 to 64.0 kW/m2. This work is an attempt to enrich the database already collected on turbulent subcooled boiling flow, with the hope that it will be useful in turbulence modeling efforts.
Review and Status of the Gen-IV CANDU-SCWR Passive Moderator Core Cooling System
Authors: Khartabil H. F.
Published: July 2009

Enhanced safety is an important priority in the development of Generation IV reactors, which can be accomplished through the use of improved passive heat removal systems. In CANDU® reactors, the separation between the low-pressure moderator and high-pressure coolant provides a unique passive heat sink for decay heat removal during accident scenarios. Methods for enhancing this passive heat sink for the GenIV CANDU-SCWR (supercritical water cooled reactor) have been under investigation for the past several years to support a “no core melt” reactor design concept (1, 2). Initially, to test feasibility, tests and analysis at AECL studied a full-height passive cooling loop and showed that a flashing-driven natural circulation system was possible in principle. However, flow oscillations were observed at low powers and could not be readily explained through analysis. While these oscillations were not considered to be detrimental to the heat removal capability, additional separate-effects experiments were conducted and causal mechanisms proposed for the oscillations. In addition, these separate effects tests suggested that oscillations could be avoided at any power level by suitable design. A new test loop with a more representative geometry was recently constructed and commissioned. Preliminary commissioning tests confirmed conclusions from the separate effects tests. In this paper, the new tests are compared to the past tests to explain the improved and more stable loop operation. This comparison suggests that a complete system coupled to an ultimate heat sink has the potential to improve loop operation even more by eliminating or significantly reducing flow oscillations at low powers. Plans for validating this conclusion will be provided.
Risk-Informed Safety Classification for New Plant Designs
Authors: Palm Nathan A.
Published: July 2009

Risk-informed safety classification can be used to identify power plant components that are important to plant safety. Alternative treatment requirements can then be allowed for select components based on their risk-informed safety classification. This concept has been embraced in Section XI of the ASME Boiler and Pressure Vessel Code through the approval of Code Cases N-660 and N-662. Draft Code Case N-720 provides an approach to applying risk-informed safety classification to new plant designs in Section III of the ASME Boiler and Pressure Vessel Code. For new plant designs some unique challenges exist compared to those encountered in the application of the concept to an operating plant. This paper will discuss the risk-informed safety classification approach in Code Case N-720 and the key differences in its application to new plant designs. The alternative treatment requirements proposed in the Code Case and their benefits will also be discussed.
Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project
Authors: Suzuki Mitsuhiro , Takeda Takeshi , Nakamura Hideo
Published: July 2009

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.
Roadmap for Research, Development, and Demonstration of Instrumentation, Controls, and Human-Machine Interface Technologies
Authors: Miller Don W. , Arndt Steven A. , Bond Leonard J. , Dudenhoeffer Donald D. , Hallbert Bruce P. , Holcomb David E. , Wood Richard T. , Naser Joseph A. , O’Hara John , Quinn Edward L.
Published: July 2009

Instrumentation, controls, and human-machine interfaces are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The nuclear power industry is currently engaged in a transition from traditional analog-based instrumentation, controls, and human-machine interface (ICHMI) systems to implementations employing digital technologies. This transition has primarily occurred in an ad hoc fashion through individual system upgrades at existing plants and has been constrained by a number of concerns. Although international implementation of evolutionary nuclear power plants and the progression toward new plants in the United States have spurred design of more fully digital plantwide ICHMI systems, the experience base in the nuclear power application domain is limited. Additionally, design and development programs by the U.S. Department of Energy (DOE) for advanced reactor concepts, such as the Generation IV Program and Next Generation Nuclear Plant (NGNP), introduce different plant conditions and unique plant configurations that increase the need for enhanced ICHMI capabilities to fully achieve programmatic goals related to economic competitiveness, safety and reliability, sustainability, and proliferation resistance and physical protection. As a result, there are challenges that need to be addressed to enable the nuclear power industry to effectively and efficiently complete the transition to safe and comprehensive use of digital technology.
Management of Component Fatigue in CANDU Stations for Life Extension
Authors: Yetisir M. , Stevens G. L. , Robertson S.
Published: July 2009

CANDU® nuclear generating stations and their components were designed for 30 effective full power years (EFPY) of operation. Many CANDU plants are now approaching their design end-of-life and are being considered for extended operation beyond their design life. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC), has asked utilities to consider component fatigue issues in plant life extension (PLEX) applications. In particular, environmental effects on fatigue is identified as an issue that needs to be addressed, similar to that being addressed for license renewal for U.S. nuclear power plants. To address CNSC concerns, CANDU stations have initiated a program to develop component fatigue management programs for PLEX operation. A pilot study conducted in a typical CANDU plant showed that: • Only 10 to 15% of the numbers of design transients have been used after 25 EFPY of operation. Hence, a significant amount of original design fatigue usage margin remains available for PLEX operation. • Environmental fatigue considerations in heavy water (D2O) were included in the assessment. Only warm-up transients are assessed to have dissolved oxygen concentrations that can result in a significant environmental effect for the ferritic steels used in the CANDU primary and secondary systems. • Due to the low accumulation of transients, and the relative absence of thermal stratification mechanisms, thermal fatigue is not as significant an issue in CANDU plants as in pressurized water reactor (PWR) and boiling water reactor (BWR) plants. This paper summarizes the results of the pilot study conducted for the Canadian CANDU plants.
PBMR Fuel Kernel Model for the Prediction of Accurate Temperature Profiles
Authors: Ubbink Onno , du Toit Pieter S. , Lourens Pierre , Joubert Wessel R.
Published: July 2009

PBMR (Pebble Bed Modular Reactor) is a High-Temperature Gas-cooled Reactor (HTGR) utilizing spherical pebble like fuel elements. A pebble is a moulded graphite sphere about the size of a tennis ball that contains approximately 15000 homogeneously distributed, triso-coated low-enriched uranium dioxide (UO2) particles, about 1mm in diameter. In the case of fast reactivity transients the accurate time-dependent calculation of the uranium temperature is essential as the neutron balances in the nuclear reactor are strongly influenced by the actual fuel temperature. This paper presents a calculation model that calculates the temperature profile through a representative fuel kernel, its coating layers and the associated graphite moderator. The local nuclear fission heat is deposited directly in the fuel itself. Great care is taken with the definition of the boundary conditions and implementation thereof to ensure that the kernel temperature calculation model describes the physics as accurately as possible. This paper reports on this in detail. A sample calculation is included to illustrate the effect of and need for a more accurate model.
RELAP5-3D Analysis of OECD-NEA/NRC BFBT Benchmark
Authors: Kovtonyuk Andriy , Petruzzi Alessandro , Parisi Carlo , D’Auria Francesco
Published: July 2009

OECD-NEA and the NRC organized and sponsored the BWR Fuel Bundle Test (BFBT) Benchmark with the main purpose of assessing sub-channel and Computational Fluid Dynamic codes capabilities in estimating relevant thermal-hydraulics parameters like void fraction and critical power for a BWR boiling channel. The assessment activity is performed comparing the code calculation results with the experimental data at steady state and transient conditions available through the Japanese Nuclear Power Engineering Corporation (NUPEC). In this framework, the San Piero a Grado Nuclear Research Group (GRNSPG) of the University of Pisa (UNIPI) developed a RELAP5-3D© thermal-hydraulic nodalization of the experimental bundle. The main purpose of this activity was the assessment of the capability of the well-known three-dimensional system thermal-hydraulic code RELAP5-3D© for the prediction of relevant parameters at the fuel assembly scale. In order to exploit the large amount of experimental data available, a three dimensional thermal-hydraulic nodalization was developed, simulating all sub-channels with MULTI-D component. The overall activities resulted in challenges for the code and the code users because of the necessary large number of nodes and heat structures used and because of the different solutions that had to be found for performing a typical sub-channel analysis with a system thermal-hydraulic code. Several calculations simulating steady state conditions were performed for different fuel assembly configurations. For each of them the void fraction distributions in all fuel assembly sub-channels and pressure drops along different part of an assembly were compared with the available experimental data. In the next exercises, BWR transients were executed (turbine trip and recirculation pump trip, respectively), calculating again the void distributions and critical power conditions. The results of the activity demonstrated the capability of the RELAP5-3D© code to perform calculations using a sub-channel approach. The code was able to calculate several thermal-hydraulics parameters with high accuracy at “fuel bundle” level of resolution; the results of “sub-channel” level are instead affected by a higher error (e.g. deviation of around 20% in the prediction of void distribution). Sub-channel results showed a better agreement when considering high quality tests compared to the lower quality ones.