bluuebunny/crossref_metadata_2025 huggingface dataset:
Multiscale Assessment of Random Polycrystalline Aggregates With Short Cracks
Authors: Cizelj L. , Simonovski I.
Published: September 2008

The complete understanding of the incubation and growth of microstructurally short cracks is still somewhat beyond the present state-of-the-art explanations. A good example is the intergranular stress corrosion cracking of Inconel 600 in high-temperature water. An effort was therefore made by the authors to construct a computational model of the crack growth kinetics at the grain-size scale. The main idea is to divide continuum (e.g., polycrystalline aggregate) into a set of sub-continua (grains). Random grain structure is modelled using Voronoi-Dirichlet tessellation. Each grain is assumed to be a monocrystal with random orientation of the crystal lattice. Elastic behaviour of grains is assumed to be anisotropic. Crystal plasticity is used to describe (small to moderate) plastic deformation of monocrystal grains. Explicit geometrical modelling of grain boundaries and triple points allows for the development of the incompatible strains along the grain boundaries and at triple points. Finite element method (ABAQUS) is used to obtain numerical solutions of strain and stress fields. The analysis is currently limited to two-dimensional models. Numerical examples illustrate analysis of about one grain boundary long transgranular cracks. In particular, the dependence of crack tip displacements on the random orientation of neighbouring grains is studied. The limited number of calculations performed indicates that the incompatibility strains, which develop along the boundaries of randomly oriented grains, significantly influence the local stress fields and therefore also the crack tip displacements. First attempts are also made to quantify the preferential growth directions of cracks crossing the discontinuities (e.g., grain boundary).
A Review of the PSA Model Truncation Convergence Issues at Some ENS Plants
Authors: Yilmaz Fatma
Published: September 2008

This study provides the insights gained from the Probabilistic Safety Assessment (PSA) model update of several Entergy Nuclear South (ENS) plants with respect to truncation convergence based on the limited guidance on the issue in the industry. The industry rule of thumb, the ASME and NRC guidance and requirements on the subject have been reviewed. The recent model updates performed at some of the ENS plants (River Bend, ANO 1 and 2) considered these criteria. Based on the current criteria used in the industry for truncation convergence, the recent PSA model update results for the River Bend Station (RBS) and ANO-1 are not converging even at a low truncation limit of 1E−11/reactor-year (yr). Many improvements were introduced in the recent model updates and convergence was expected at higher truncation values. This paper discusses the issues identified that are related to the convergence of the PSA results at low truncation limits.
A Review of the Initiating Event Fault Tree Development and Usage at the ENS Sites
Authors: Yilmaz Fatma , Bedell Loys
Published: September 2008

This paper provides an overview of the experience in developing and using the initiating event fault trees (IEFTs) in the base and EOOS Probabilistic Safety Assessment (PSA) models at the Entergy Nuclear South (ENS) sites. Simple fault tree models can be developed and linked to the master fault tree in order to provide a better estimate of plant-specific initiating event frequencies. If developed considering certain aspects, there can be significant advantages to using IEFT methodology for some initiators, especially in the plant EOOS models. The most important advantage is to have EOOS re-calculate and use such initiating event frequencies for different alignments taking the components that are out of service into consideration while keeping the quantification time and complexity of the cutsets at an acceptable level. The important steps for developing the IEFTs, some insights gained through the experience at the ENS sites, and the advantages of using IEFTs will be discussed in this paper.
An Example of an INPRO Assessment of an INS in the Area of Economics
Authors: Florido P. , Allan C. , Depisch F.
Published: September 2008

Following a resolution of the General Conference of the IAEA in the year 2000 an International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contribution to meeting the energy needs of the 21st century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. This paper discusses the example presented in the manual for performing an INPRO assessment in the area of economics. The example considers a private utility, operating in a liberalized market that is planning for an additional supply of power of about 600 MWe within a time frame of 10 years. Two nuclear options, an LWR and a HWR, are considered as well as a gas-fired plant using liquefied gas as fuel.
An Example of an INPRO Assessment of an INS in the Area of Waste Management
Authors: Allan C. , Busurin Y. , Depisch F.
Published: September 2008

Following a resolution of the General Conference of the IAEA in the year 2000 the International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. INPRO has defined requirements organized in a hierarchy of Basic Principles, User Requirements and Criteria (consisting of an indicator and an acceptance limit) to be met by innovative nuclear reactor systems (INS) in six areas, namely: economics, safety, waste management, environment, proliferation resistance, and infrastructure. If an INS meets all requirements in all areas it represents a sustainable system for the supply of energy, capable of making a significant contribution to meeting the energy needs of the 21st century. Draft manuals have been developed, for each INPRO area, to provide guidance for performing an assessment of whether an INS meets the INPRO requirements in a given area. The manuals set out the information that needs to be assembled to perform an assessment and provide guidance on selecting the acceptance limits and, for a given INS, for determining the value of the indicators for comparison with the associated acceptance limits. Each manual also includes an example of a specific assessment to illustrate the guidance. This paper discusses the example presented in the manual for performing an INPRO assessment in the area of waste management. The example, chosen solely for the purpose of illustrating the INPRO methodology, describes an assessment of an INS based on the DUPIC fuel cycle. It is assumed that uranium is mined, milled, converted, enriched, and fabricated into LWR fuel in Canada. The LWR fuel is assumed to be leased to a utility in the USA. The spent LWR fuel is assumed to be returned to Canada where it is processed into CANDU DUPIC fuel, which is then burned in CANDU reactors. The assessment steps and the results are presented in detail in the paper. The example illustrates an assessment performed for an INS at an early stage of development.
Modified Phenomena Identification and Ranking Table (PIRT) for Uncertainty Analysis
Authors: Gol-Mohamad Mohammad P. , Modarres Mohammad , Mosleh Ali
Published: September 2008

This paper describes a methodology of characterizing important phenomena, which is also part of a broader research by the authors called “Modified PIRT”. The methodology provides robust process of phenomena identification and ranking process for more precise quantification of uncertainty. It is a two-step process of identifying and ranking methodology based on thermal-hydraulics (TH) importance as well as uncertainty importance. Analytical Hierarchical Process (AHP) has been used for as a formal approach for TH identification and ranking. Formal uncertainty importance technique is used to estimate the degree of credibility of the TH model(s) used to represent the important phenomena. This part uses subjective justification by evaluating available information and data from experiments, and code predictions. The proposed methodology was demonstrated by developing a PIRT for large break loss of coolant accident LBLOCA for the LOFT integral facility with highest core power (test LB-1).
The Worlds’ First Ever Cooling Tower Acceptance Test Using Process Data Reconciliation
Authors: Langenstein Magnus , Hansen-Schmidt Jan
Published: September 2008

The cooling capacity of cooling towers is influenced by multiple constructive and atmospheric parameters in a very complex way. This leads to strong variations of the measured cold-water temperature and causes unacceptable unreliability of conventional acceptance tests, which are based on single point measurements. In order to overcome this lack of accuracy a new approach to acceptance test based on process data reconciliation has been developed by BTB Jansky and applied at a nuclear power plant. This approach uses process data reconciliation according to VDI 2048 [1, 2] to evaluate datasets over a long period covering different operating conditions of the cooling tower. Data reconciliation is a statistical method to determine the true process parameters with a statistical probability of 95% by considering closed material-, mass- and energy balances. Datasets which are not suitable for the evaluation due to strong transient gradients are excluded beforehand, according to well-defined criteria. The reconciled cold-water temperature is then compared, within a wet bulb temperature range of 5°C to 20°C to the manufacturer’s guaranteed temperature. Finally, if the average deviation between reconciled and guaranteed value over the evaluated period is below zero, the cooling tower guarantee is fulfilled.
Applicability of Loss of Offsite Power (LOSP) Events in NUREG/CR-6890 for Entergy Nuclear South (ENS) Plants’ LOSP Calculations
Authors: Li Yunlong , Yilmaz Fatma , Bedell Loys
Published: September 2008

Significant differences have been identified in loss of offsite power (LOSP or LOOP) event description, category, duration, and applicability between the LOSP events used in NUREG/CR-6890 and ENS’ LOSP packages, which were based on EPRI LOSP reports with plant-specific applicability analysis. Thus it is appropriate to reconcile the LOSP data listed in the subject NUREG and EPRI reports. A cross comparison showed that 62 LOSP events in NUREG/CR-6890 were not included in the EPRI reports while 4 events in EPRI reports were missing in the NUREG. Among the 62 events missing in EPRI reports, the majority were applicable to shutdown conditions, which could be classified as category IV events in EPRI reports if included. Detailed reviews of LERs concluded that some events did not result in total loss of offsite power. Some LOSP events were caused by subsequent component failures after a turbine/plant trip, which have been modeled specifically in most ENS plant PRA models. Moreover, ENS has modeled (or is going to model) the partial loss of offsite power events with partial LOSP initiating events. While the direct use of NUREG/CR-6890 results in SPAR models may be appropriate, its direct use in ENS’ plant PRA models may not be appropriate because of modeling details in ENS’ plant-specific PRA models. Therefore, this paper lists all the differences between the data in NUREG/CR-6890 and EPRI reports and evaluates the applicability of the LOSP events to ENS plant-specific PRA models. The refined LOSP data will characterize the LOSP risk in a more realistic fashion.
Application of RFID to High-Reliability Nuclear Power Plant Construction
Authors: Akagi Kenji , Ishiwata Masayuki , Araki Kenji , Kawahata Jun-Ichi
Published: September 2008

In nuclear power plant construction, countless variety of parts, products, and jigs more than one million are treated under construction. Furthermore, strict traceability to the history of material, manufacturing, and installation is required for all products from the start to finish of the construction, which enforce much workforce and many costs at every project. In an addition, the operational efficiency improvement is absolutely essential for the effective construction to reduce the initial investment for construction. As one solution, RFID (Radio Frequent Identification) application technology, one of the fundamental technologies to realize a ubiquitous society, currently expands its functionality and general versatility at an accelerating pace in mass-production industry. Hitachi believes RFID technology can be useful of one of the key solutions for the issues in non-mass production industry as well. Under this situation, Hitachi initiated the development of next generation plant concept (ubiquitous plant construction technology) which utilizes information and RFID technologies. In this paper, our application plans of RFID technology to nuclear power is described.
Anisotropic Turbulence Modeling for Accurate Rod Bundle Simulations
Authors: Baglietto Emilio
Published: September 2008

An improved anisotropic eddy viscosity model has been developed for accurate predictions of the thermal hydraulic performances of nuclear reactor fuel assemblies. The proposed model adopts a non-linear formulation of the stress-strain relationship in order to include the reproduction of the anisotropic phenomena, and in combination with an optimized low-Reynolds-number formulation based on Direct Numerical Simulation (DNS) to produce correct damping of the turbulent viscosity in the near wall region. This work underlines the importance of accurate anisotropic modeling to faithfully reproduce the scale of the turbulence driven secondary flows inside the bundle subchannels, by comparison with various isothermal and heated experimental cases. The very low scale secondary motion is responsible for the increased turbulence transport which produces a noticeable homogenization of the velocity distribution and consequently of the circumferential cladding temperature distribution, which is of main interest in bundle design. Various fully developed bare bundles test cases are shown for different geometrical and flow conditions, where the proposed model shows clearly improved predictions, in close agreement with experimental findings, for regular as well as distorted geometries. Finally the applicability of the model for practical bundle calculations is evaluated through its application in the high-Reynolds form on coarse grids, with excellent results.
Materials Research Project to Support Code Changes for GEN IV: A DOE/ASME Cooperative Effort
Authors: Ramirez James , Erler Bryan A. , Jetter Robert
Published: September 2008

For the last four years as reported in ICONE 13 Paper 13-50638, the ASME Board of Nuclear Codes and Standards (BNCS) has been leading an effort to identify code changes necessary to support the future nuclear plants of the world. In that paper the authors identified the results of meetings with NSSS suppliers, government regulators, engineers/constructors, and owner operators to ascertain the status of their future designs and what modifications are necessary so the right rules and materials are in ASME Nuclear Codes and Standards.
Design of Solid Form Xenon-124 Target for Producing I-123 Radioisotope Using Computer Simulation Techniques
Authors: Kamali Moghaddam K. , Kakavand T. , Sadeghi M. , Shokri Bonab S.
Published: September 2008

One of the most famous radioisotopes used in nuclear medicine is Iodine-123. Among the various types of nuclear reactions for producing I-123, the following reactions are favored due to the absence of I-124 and I-125 impurities: 54124Xe(P,2n)55123Cs5.9minβ+54123Xe2.08hrβ+53123I54124Xe(P,pn)54123Xe2.08hrβ+53123I Recently in Cyclotron and Nuclear Medicine Department of NRCAM, at Atomic Energy organization of Iran (AEOI), a system for producing I-123 via Xe-124 gas target technology, has been constructed and installed. One of the major problems in this system is the highly expensive cost of the enriched Xenon-124 gas. Therefore, saving this gas inside the system is very important. Unfortunately, by accidental rupture of the window foil or bad function of O-rings, the whole Xenon gas will escape from the system immediately. In this paper, by using computer codes; ALICE91, SRIM and doing some calculations we are going to demonstrate our latest effort for feasibility study of producing I-123 with the above mentioned reactions, but using Xe-124 solid target instead. According to our suggested design, a conical shaped irradiation vessel made of copper with 1mm thickness, 1cm outlet diameter, 5cm length and 12° angle at summit can be fixed inside a liquid nitrogen housing chamber. The Xenon-124 gas will be sent to the inside of this very cold conical trap and eventually deposited on its surface in solid form. Our calculation shows that during bombardment with 17–28 MeV proton energy, the thickness of solidified Xenon layer will remain around .28 mm. Likewise; thermo-dynamical calculation shows that in order to prevent the evaporation of solidified Xenon, the maximum permissible proton beam current for this system should be less than 1.4 μA. According to these working conditions, the production yield of I-123 can be predicted to be around 150 mCi/μAh.
Primary-Side Two-Phase Flow and Heat Transfer Characteristics of a Horizontal-Tube PCCS Condenser
Authors: Kondo Masaya , Nakamura Hideo , Kukita Yutaka , Kurita Tomohisa , Arai Kenji , Okazaki Toshihiko , Inoue Ryousuke
Published: September 2008

An experiment was performed using a horizontal condenser tube for the PCCS horizontal heat exchanger to confirm the performance and to estimate the thermal-hydraulic behavior in the tube thermally and visually. The PCCS, driven by the pressure difference between the drywell and wetwell, is a safety equipment to prevent the containment break for more than 1 day due to overpressurization at the containment spray failure situation. An annular flow, a wavy flow, a supercritical flow, a stratified flow and the liquid slug propagation are observed in the tube at the nominal condition. The flow regime transition, induced by the condensation, roughly agrees to the existing models. The observation also agrees with the condensate distribution in the tube circumference direction, which is suggested by the local thermal characteristics.
Modelling Heat Transfer in Very High Temperature Gas Cooled Reactor
Authors: Singh Suneet , Uddin Rizwan
Published: September 2008

Very High Temperature Reactor (VHTR) is one of the six reactor concepts selected for further consideration by Generation IV International Forum (GIF) members. Adequate heat removal under steady-state and transient condition is important for the safe operation of VHTR. In one of the proposed designs for reactor cavity cooling system, under off normal conditions, coolant comes down through a downcomer and is then distributed into riser channels. The purpose of this design is to enhance heat transfer rates. The heat transfer characteristics in such a cavity are modeled by natural convection in a square cavity with plate. The plate divides the cavity into channels. The cavity is open at the top and the left side wall is isothermal. The plate is parallel to the side walls. The plate is isothermal and is maintained at a lower temperature than the left wall. The plate is shorter than the height of the cavity so that flow down in one channel turns up in the next channel. The right wall has a temperature lower than the plate, and bottom of the cavity is adiabatic. The flow in the cavity is simulated by recently developed Modified Nodal Integral Method (MNIM). Simulations are done for two different heights of the plate. The cavity without plate is also studied for comparison purposes. The numerical investigation is done for Rayleigh numbers of 103, 104 and 105. Nusselt numbers at hot wall and volumetric flow rate into the cavity are reported for all nine cases. Results show that there is significant enhancement in heat transfer in the presence of a plate for higher Rayleigh numbers.
A Root Cause Study on AOA-PWR and CDA-VVER: A Point of View of “Long-Cell Action” Corrosion Mechanism
Authors: Saji Genn
Published: September 2008

The author has investigated the characteristics of boron co-deposition with crud experienced in AOA and iron ferrite deposition in CDA. Corrosion product deposits found in cores with appreciable AOA have been reported in mostly nickel-based (as NiO or elemental nickel) as opposed to nickel ferrite deposits common to non-boiling cores. Significant quantities of meta-ZrO2 and nickel iron oxyborates (bonaccordite), notably Ni2FeBO5 have also been found in deposits on cores with AOA. On the basis of this general characterization information, the author has constructed a potential-pH diagram of Ni2FeB(OH)10, which is a hydrated state of FeNi2(BO3)O2 as summarized in this paper. Although preliminary, the estimated E-pH diagram suggests some interesting observation, including: growth of bonaccordite “needles” on the clad is associated with a local anodic electrochemical reaction necessary to remove excess electrons from the system to a cathode. During the AOA cycle, the concentration of nickel and iron ions must have been unusually high as they should be for a significant amount of crud deposits. The author thinks such an acceleration of the anodic dissolution of metal cations is due to the effect of the long cell action corrosion mechanism. As early as 1949, an Italian scientist Petracchi demonstrated that electrochemical effects significantly influence the erosion rate. He constructed a flow nozzle with specimens kept under external electrical potential. Upon inducing as low as 0.1 mA/cm2 of the positive current, the erosion rates were reported drastically increased. No erosion was observed by reversing the polarity of the potential. As discussed in a companion paper also presented at this conference [1], the author discusses various mechanisms (electrochemical cell configurations) that induce potential differences, including those differences in ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author discusses how these potential differences are related to the AOA/CDA issues in PWR/VVER plants. The author is calling for further verification experiments regarding this corrosion mechanism as a joint international project.
Installation of Valvtechnologies Ball Valves in Severe Service Applications Within the Advanced Test Reactor
Authors: Howard Robert C.
Published: September 2008

The Advanced Test Reactor (ATR) located at the Department of Energy’s Idaho National Laboratory, is the most powerful test reactor operating in the United States rated at a design power of 250 MW(t). Operating cycles are nominally seven per year with outages that last 7 to 14 days, allowing time for routine plant maintenance and experiment insertions and manipulations. While the ATR pressurized water loops can operate at the same temperature and pressure requirements of a pressurized water reactor, the loops also have the ability to operate at higher conditions. Hence, it is critical to ensure that when component replacements are called for, they can meet or exceed design requirements of a typical power reactor, while continuing to satisfy the design requirements of the ATR experiment loops.
A Novel, Safe, and Environmentally Friendly Technology for Water Production Through Recovery of Rejected Thermal Energy From Nuclear Power Plants
Authors: Khalil Yehia F. , Elimelech Menachem
Published: September 2008

In this work, we describe a novel design that utilizes seawater and a portion of rejected heat from a nuclear plant’s steam cycle to operate a water desalination system using forward osmosis technology. Water produced from this process is of sufficient quality to be readily used to supply plant demands for continuous makeup water. The proposed process minimizes the environmental concerns associated with thermal pollution of public waters and the resulting adverse impact on marine ecology. To demonstrate the technical feasibility of this conceptual design of a water treatment process, we discuss a case study as an example to describe how the proposed design can be implemented in a nuclear power station with a once–through cooling system that discharges rejected heat to an open sound seawater as its ultimate heat sink. In this case study, the station uses a leased (vendor owned and operated) onsite water treatment system that demineralizes and polishes up to 500-gpm of city water (at 100 ppm TDS) to supply high-quality makeup water (< 0.01 ppm TDS) to the plant steam system. The objectives of implementing the new design are three fold: 1) forego current practice of using city water as the source of plant makeup water, thereby reducing the nuclear station’s impact on the region’s potable water supply by roughly 100 million gallons/year, 2) minimize the adverse impact of discharging rejected heat into the open sound seawater and, hence, protect the marine ecology, and 3) eliminate the reliance on external vendor that owns and operates the onsite water treatment system, thereby saving an annual fixed cost of $600K plus 6 cents per 1,000 gallons of pure water. The design will also eliminate the need for using two double-path reverse osmosis (RO) units that consume 425 kW/h of electric power to operate two RO pumps (480V, 281.6 HP, and 317.4 amps).
Optimum Channel Inclination for Gas Venting Under Countercurrent Flow Limitations
Authors: Liao Y. , Vierow K.
Published: September 2008

Countercurrent flow limitation (CCFL) in the pressurizer surge line of future Pressurized Water Reactors (PWR) with passive safety systems is an important phenomenon in reactor safety analysis. The pressurizer surge line is typically comprised of several sections with various inclination angles. Under certain accident conditions, countercurrent flow takes place in the surge line with liquid flowing down from the pressurizer and steam flowing up from the hot leg. The steam venting rate as well as the liquid draining rate may affect the Emergency Core Cooling System (ECCS) actuation. The objective herein is to develop a physics-based model for evaluating the effect of inclination angle on CCFL. For a given liquid superficial velocity in the countercurrent flow system of the pressurizer surge line, the gas superficial velocity should be as large as possible at the onset of flooding, so that the steam can vent as fast as possible without inhibiting the pressurizer drain rate. Thus the system could depressurize in a timely manner to initiate the ECCS actuation. As indicated by CCFL experiments, for a given liquid superficial velocity, the gas superficial velocity attains a greatest value at a certain channel inclination, which is defined as the optimum channel inclination. In the present work, an analytical model is proposed to predict the optimum channel inclination under simplified conditions. The model predictions compare favorably with experimental data.
Design of Tellurium-123 Target for Producing Iodine-123 Radioisotope Using Computer Simulation Techniques
Authors: Kakavand T. , Kamali Moghaddam K. , Sadeghi M. , Ghasemi R.
Published: September 2008

Iodine-123 is one of the most famous radioisotopes for Single Photon Emission Computed Tomography (SPECT) use, so, for 123I production, the 123Te has been chosen as a target through 123Te (p,n) 123I reaction. The various enriched targets (%99.9, %91, %85.4 and %70.1) have been used for the present calculations. In the current work, by using computer codes; ALICE & SRIM and doing a sort of calculations, we are going to demonstrate our latest effort for feasibility study of producing 123I by the above mentioned reaction. By using proton beam energy of less than 30 MeV, the mentioned codes give more accurate results. The cross section of all Tellurium reactions with proton has been calculated at 0–30 MeV proton beam energy with ALICE code. In the present work, the yield of 123I has been calculated by analytical methods. Our prediction for producing 123I yield via bombardment of 123Te (%99.9) with proton beam energy at 5–15 MeV is about 7.2 mCi/μAh. The present work shows that, the 123I yield is proportional to abundance of 123Te. Thermodynamical calculations with various current beams of up to 900 μA have been done, and the proper cooling system for the above purpose has been designed.
Main Features of JSME Design and Construction Code for Fast Reactors
Authors: Morishita Masaki , Sukekawa Masayuki , Otani Tomomi
Published: September 2008

Since its foundation in 1997, the Main Committee on Power Generation Facility Codes, MC-PGFC, of the Japan Society of Mechanical Engineers, JSME, has issued a number of nuclear codes including the rules on design and construction and the rules on fitness-for-service for nuclear power plants. Some of these JSME nuclear codes have been endorsed by the regulatory body, and are now utilized in the regulatory processes of the actual plants. Among these nuclear codes recently published is the “rules on design and construction for fast reactors”. It includes as its main body design rules on class 1 components for elevated temperature services. This paper overview the main features of the code.
Study on Transient Void Behavior During Reactivity Initiated Accidents Under Low Pressure Condition: Development and Application of Measurement Technique for Void Fraction in Bundle Geometry
Authors: Satou Akira , Maruyama Yu , Asaka Hideaki , Nakamura Hideo
Published: September 2008

Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 2 × 2 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.
Numerical Analysis on Air Ingress Behavior in GTHTR300H
Authors: Takeda Tetsuaki , Yan Xing , Kunitomi Kazuhiko
Published: September 2008

Japan Atomic Energy Agency (JAEA) has been developing the analytical code for the safety characteristics of the HTGR and carrying out design study of the gas turbine high temperature reactor of 300MWe nominal-capacity for hydrogen production, the GTHTR300H (Gas Turbine High Temperature Reactor 300 for Hydrogen). The objective of this study is to clarify safety characteristics of the GTHTR300H for the pipe rupture accident. A numerical analysis of heat and mass transfer fluid flow with multi-component gas mixture has been performed to obtain the variation of the density of the gas mixture, and the onset time of natural circulation of air. From the results obtained in this analysis, it was found that the duration time of the air ingress by molecular diffusion would increase due to the existence of the recuperator in the GTHTR300H system.
Numerical Analysis of Solitary Wave Influence on the Filmwise Condensation in Presence of Non-Condensable Gases
Authors: Karkoszka Krzysztof , Anglart Henryk
Published: September 2008

This paper is dealing with the analysis of condensation in presence of non-condensable gas on a laminar liquid film falling down on a vertical smooth surface. Particular interest is focused on the influence of solitary waves on the condensation process. Solutions to the pressure, velocity, temperature and additional scalar variable fields are obtained numerically by solving two-dimensional Navier–Stokes equations formulated in a general coordinate system and applying the artificial compressibility method. The whole system of equations together with adequate boundary conditions is implemented using the finite difference method and solved in the Matlab® code. Both implicit Crank–Nicolson and Euler schemes for the time derivatives are initially used and the latter one is chosen as a more stable. All computations are carried out with prescribed geometry for a film and gas domains and a special attention is focused mainly on the modelling of the influence of the interfacial boundary conditions on the heat transfer process between gaseous mixture and liquid phases. Description of the physical, mathematical and numerical models and several examples of the solutions are presented. Conclusions on the wave hydrodynamics influence on the heat transfer during phase change process are drawn.
Strategy for the Development of a DNB Local Predictive Approach Based on Neptune CFD Software
Authors: Haynes Pierre-Antoine , Pe´turaud Pierre , Montout Michae¨l , Hervieu Eric
Published: September 2008

The NEPTUNE project constitutes the thermal-hydraulics part of a long-term joint development program for the next generation of nuclear reactor simulation tools. This project is being carried through by EDF (Electricite´ de France) and CEA (Commissariat a` l’Energie Atomique), with the co-sponsorship of IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and AREVA NP. NEPTUNE is a multi-phase flow software platform that includes advanced physical models and numerical methods for each simulation scale (CFD, component, system). NEPTUNE also provides new multi-scale and multi-disciplinary coupling functionalities. This new generation of two-phase flow simulation tools aims at meeting major industrial needs. DNB (Departure from Nucleate Boiling) prediction in PWRs is one of the high priority needs, and this paper focuses on its anticipated improvement by means of a so-called “Local Predictive Approach” using the NEPTUNE CFD code. We firstly present the ambitious “Local Predictive Approach” anticipated for a better prediction of DNB, i.e. an approach that intends to result in CHF correlations based on relevant local parameters as provided by the CFD modeling. The associated requirements for the two-phase flow modeling are underlined as well as those for the good level of performance of the NEPTUNE CFD code; hence, the code validation strategy based on different experimental data base types (including separated effect and integral-type tests data) is depicted. Secondly, we present comparisons between low pressure adiabatic bubbly flow experimental data obtained on the DEDALE experiment and the associated numerical simulation results. This study anew shows the high potential of NEPTUNE CFD code, even if, with respect to the aforementioned DNB-related aim, there is still a need for some modeling improvements involving new validation data obtained in thermal-hydraulics conditions representative of PWR ones. Finally, we deal with one of these new experimental data needs and present a scaling method for the design of the associated experimentation devoted to the analysis of the dynamics-related modeling of a bubbly flow in PWR representative conditions.
A Fuel Channel Design for CANDU-SCWR
Authors: Chow C. K. , Bushby S. J. , Khartabil H. F.
Published: September 2008

The CANDU®-Supercritical Water Reactor (CANDU-SCWR) is one of the six reactor concepts being considered by the Generation-IV International Forum (GIF) for international collaborative R&D. With SCW coolant, the thermodynamic efficiency is increased to over 40%. The CANDU-SCWR is moderated using heavy water, and it has fuel bundles residing inside horizontal pressure tubes, similar to the current CANDU design. The coolant, however, is light water at 25 MPa, with an inlet temperature of 350°C and an outlet temperature of 625°C. Because of the high temperature and high pressure of the coolant, the standard CANDU pressure tube design cannot be used. This paper presents one of the insulated pressure tube designs being considered for the CANDU-SCWR fuel channels. Unlike current CANDU reactors, the proposed CANDU-SCWR fuel channel does not use calandria tubes to separate the pressure tubes from the moderator. Each pressure tube is in direct contact with the moderator, which operates at an average temperature of about 80°C. The pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator. A perforated metal liner protects the insulator from being damaged by the fuel bundles and erosion by the coolant. The coolant pressure is transmitted through the perforated metal liner and insulator and applied directly to the relatively cold pressure tube. The material selection for each fuel channel component depends on its function. The fuel sheaths and the perforated liner must have high corrosion resistance in SCW, although their resident times are significantly different. The insulator must have high thermal resistance and corrosion resistance in SCW, plus sufficient strength to bear the weight of the fuel bundles without significant thickness reduction during its design life. The pressure tube is the pressure boundary material, so it must have high strength to contain the coolant. One common requirement for all in-core fuel channel components is that they should be as neutron transparent as possible. The irradiation deformation of all these components must also be considered in their design. This paper presents the design of this fuel channel, reviews existing data for materials, indicates where more data are required, and summarizes our plans to obtain these data.
Measuring the Cost of Obsolescence
Authors: White Mark
Published: September 2008

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) has provided the capability for irradiation testing of nuclear fuels and materials since 1967, and is expected to operate for several more years. Within the scope of extending the life of a nuclear plant is dealing with aging and obsolescence issues. A component can be considered obsolete if the manufacturer no longer supports the component, or if the manufacturer does not even exist anymore. Though these components can be considered obsolete, the cost of obsolescence may or may not be significant; it may be more cost-effective to leave and/or repair the component rather than to replace it. The project at hand is to develop a tool that will not only identify these components, series of components, or entire systems that are obsolete, but to quantify the cost of obsolescence. This engineering tool will be based on empirical formulas created from data collected from factors that deal with obsolescence. These factors are primarily, the cost of item replacement, current cost of maintenance, cost of maintenance of the replacement, cost of failure, risk of failure, safety, increase in performance/efficiency, length of manufacturer’s support, and so forth. The objective is to be able to look at the outcome of this engineering tool and clearly see what needs to be replaced, be it a component, series of components, or an entire system. If there are several such replacements needed, which one(s) have the greatest priority for replacement. Therefore the engineering tool will identify, quantify, and prioritize the cost of obsolescence in the plant. An engineering tool of this type should find application in a number of nuclear and non-nuclear facilities. While the engineering tool is being developed, the first stage of development will be on system components. Once the foundation is set it will be used to evaluate other systems and eventually expand and develop the engineering tool for the entire plant.
Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor
Authors: Cheng Lap Y. , Ludewig Hans , Jo Jae
Published: September 2008

A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645m2) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800kPa.
Application of Standard Maintenance Windows in PHWR Outage
Authors: Jiang Fuming
Published: September 2008

The concept of Standard Maintenance Windows has been widely used in the planned outage of light water reactor in the world. However, due to the specific feature of Pressurized Heavy Water Reactor (PHWR), it has not come to a consensus for the PHWR owners to adopt Standard Maintenance Windows for planned outage aiming at the optimization of outage duration. Third Qinshan Nuclear Power Company (TQNPC), with their experience gained in the previous outages and with reference to other PHWR power plants, has identified a set of Standard Maintenance Windows for planned outage. It can be applied to similar PHWR plants and with a few windows that are specific to Qinshan Phase III NPP. The use of these Standard Maintenance Windows in planned outage has been proved to be effective in control shutdown nuclear safety, minimize the unavailability of safety system, improve the efficient utilization of outage duration, and improved the flexibility of outage schedule in the case of emergency issue, which forced the revision of outage schedule. It has also formed a solid foundation for benchmarking. The identification of Standard Maintenance Windows and its application will be discussed with relevant cases for the common improvement of outage duration.
A Computer Program for Simulating Transient Behavior in Steam Turbine Stage Pressure of AHWR
Authors: Dutta Anu , Thangamani I. , Chakraborty G. , Ghosh A. K. , Kushwaha H. S.
Published: September 2008

It is proposed to couple the Advanced Heavy water reactor (AHWR), which is being developed by Bhabha Atomic Research Centre, India, with a desalination plant. The objective of this coupling is to produce system make-up and domestic water. The proposed desalination plant needs about 1.9 kg/sec of steam and the minimum pressure requirement is 3 bars. The desalination plant can be fed with bled steam extracted from a suitable stage in low pressure turbine. As the turbine stage pressure changes with the load, it is essential to know the availability of bled steam at aforesaid pressure for various load condition. The objective of the present study is to identify a suitable extraction point so as to ensure availability of steam at desired condition for desalination plant, even at part load conditions. In order to fulfill the above objective a steam and feed system analysis code was developed which incorporates the mathematical formulation of different components of the steam and feed system such as, high pressure (HP) and low pressure (LP) turbines, re-heater, feed heaters etc. The dynamic equations are solved simultaneously to obtain the stage pressure at various load conditions. Based on the results obtained, the suitable extraction stage in LP turbine was selected. This enables to determine the lowest possible part load operation up to which availability of desalination plant could be ensured.
Evaluation of Potential to Air Ingestion From RWT Following RAS
Authors: Bang Young S. , Kim Ingoo , Woo Sweng W.
Published: September 2008

At the Recirculation Actuation Signal (RAS) when the Refueling Water Tank (RWT) water level decreased to a certain value following Loss-of-Coolant Accident (LOCA), the isolation valves of Containment Recirculation Sump (CRS) of the Korean Standard Nuclear Power Plants (KSNP) are open automatically while the RWT isolation valves would be closed manually. It was concerned whether the design has a potential to air ingestion to Emergency Core Cooling System (ECCS) pumps before completion of the manual action to close RWT isolation valves. To support the safety evaluation on this issue including the evaluation of design adequacy, an analysis of the hydraulic transient within the ECCS piping following the RAS in KSNP is performed. RELAP5/MOD3.3 code is used to calculate the transient behavior of the piping network. The code was known to have capability to calculate one-dimensional two-phase transient flow with noncondensible gas in the complex piping. Substantial portion of ECCS are modeled including RWT, CRS, each pipe line from RWT and CRS to connection point with its own isolation valve and check valve, a common pipe line to ECCS header, each pipe line from the header to High Pressure Safety Injection (HPSI) pump, Low Pressure Safety Injection (LPSI) pump, and Containment Spray (CS) pump. Transient hydraulic behavior in the piping system following RAS after LOCA is calculated. It is found that the RWT water level was always higher than the elevation of the check valve at the connecting point by more than 15 ft. It indicates the air intrusion to the check valve can be sufficiently prevented by this amount of water head.
Target Study for the RACE HP Experiment
Authors: Agostini P. , Ciotti M. , Petrovich C. , Beneamati G. , Bergeron A. , Elmi N. , Krakowiak C. , Sansone L. , Beller D. , Carta M.
Published: September 2008

The full development of a multi-disks uranium target is described, including neutron production analysis, power deposition distributions, thermo-mechanical simulations and cooling schemes. A detailed material choice and a geometrical optimization were carried on in order to maximize the neutron production.