bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Oxygen Potential on Crack Growth in Alloys for Advanced Energy Systems
Authors: Benz Julian K. , Kim J. H. , Ballinger Ronald G.
Published: July 2009

The effect of oxygen partial pressure on fatigue and SCC growth rates in alloy 617 has been studied using both static and fatigue loading @ 650°C over the oxygen partial pressure range 10−19–10−2 atm. Tests were conducted at either constant stress intensity factor, K, for static conditions or constant ΔK in fatigue. Oxygen concentration was measured on both the inlet and outlet as well as in-situ with a probe located directly at the specimen surface. For fatigue loading the crack path was observed to be transgranular but crystallographic with a decreasing growth rate as the oxygen concentration decreased. However, for static loading the crack path shifted to intergranular and exhibited an increasing crack growth rate with decreasing oxygen concentration.
Creep Effects on Design Below the Temperature Limits of ASME Section III Subsection NB
Authors: Sham T.-L. , Jetter Robert I. , Eno Daniel R.
Published: July 2009

Some recent studies of material response have identified an issue that crosses over and blurs the boundary between ASME Boiler and Pressure Vessel Code Section III Subsection NB and Subsection NH. For very long design lives, the effects of creep show up at lower and lower temperature as the design life increases. Although true for the temperature at which the allowable stress is governed by creep properties, the effect is more apparent, e.g. creep effects show up sooner, at local structural discontinuities and peak thermal stress locations. This is because creep is a function of time, temperature and stress and the higher the localized stress, the lower in temperature creep begins to cause damage. If the threshold is below the Subsection NB to NH temperature boundary, 700°F for ferritic steels and 800°F for austenitic materials, then this potential failure mode will not be considered. Unfortunately, there is no experience base with very long lives at temperatures close to but under the Subsection NB to NH boundary to draw upon. This issue is of particular interest in the application of Subsection NB rules of construction to some High Temperature Gas Reactor (HTGR) concepts. The purpose of this paper is, thus, twofold; one part is about statistical treatment and extrapolation of sparse data for a specific material of interest, A533B; the other part is about how these results could impact current design procedures in Subsection NB.
High-Temperature Properties of Nuclear Graphite
Authors: Albers Tracy L. , Batty Lionel , Kaschak David M.
Published: July 2009

The unique combination of physical properties inherent to graphite makes it an attractive material for use as a moderator in high-temperature nuclear reactors (HTR’s). High-temperature physical properties of three nuclear grade graphites manufactured by GrafTech International Holdings Inc. (GrafTech) (PCEA, PCIB-SFG, and PPEA) have been determined experimentally and are presented here. Tensile strength, Young’s modulus, thermal conductivity, specific resistance, and coefficient of thermal expansion (CTE) data are collected at temperatures from 25 °C to as high as 2000 °C and are found to be consistent with classical graphite behavior.
Submerged Helium Blower Experience With Specific Reference to the Electro Magnetic Bearing Behaviour on the PBMR Helium Test Facility
Authors: Jacobs G. G. , Hammann G. J. F. , Rossouw M. A.
Published: July 2009

The Helium Test Facility (HTF) situated at Pelindaba in the North-West Province of South Africa is one of the test facilities that is owned and operated by the Pebble Bed Modular Reactor (Pty) Ltd. company (PBMR). The main purpose of the HTF is to test the main support systems for PBMR’s demonstration reactor. Helium circulation is achieved with a submerged centrifugal blower making use of electro magnetic bearings. The blower has to circulate helium through the plant at operating pressures ranging from 10 to 90 bar and operating temperatures ranging from ambient to 280 degrees Celsius. This blower has been in operation from 2006 until 2007 when it experienced an electro magnetic bearing failure. Valuable operating experience with this blower was gained. Many lessons on the requirements and operation of such a blower and the plant were learned. This paper presents the operating experience with regard to blower behaviour as well as the behaviour and failure of the electro magnetic bearings. Blower sizing and available pressure ratio is discussed. The problems associated with operation in nitrogen is also discussed. The results of a vibration investigation, that was undertaken to investigate the cause of a high pitched audible noise when the blower was operated at approximately 11,000 r/min, is presented. The electro magnetic bearing failure is discussed with photos presented of the damaged electro magnetic bearing coil. Rust was found inside the blower which is also discussed.
Development of Local Heat Transfer Models for the Safety Assessment of Prismatic Modular High Temperature Gas-Cooled Reactor Cores
Authors: Stainsby Richard , Worsley Matthew , Dawson Frances , Baker Joakim , Grief Andrew , Dennier Ana , Coddington Paul
Published: July 2009

This paper presents a model developed for determining fuel particle and fuel block temperatures of a prismatic core modular reactor during both normal operation and under fault conditions. The model is based on multi-scale modeling techniques and has been qualified by comparison with finite element simulations for both steady state and transient conditions. Further, a model for determining the effective conductivity of the block fuel elements — important for heat removal in loss of flow conditions — is presented and, again, qualified by comparison with finite element simulations. A numerical model for predicting conduction heat transfer both within and between block fuel elements has been developed which, when coupled with the above multi-scale model, allows simulations of whole cores to be carried out whilst retaining the ability to predict the temperatures of individual coolant channels and individual coated particles in the fuel if required. This ability to resolve heat transfer on length scales ranging from a few meters down to a few microns within the same model is very powerful and allows a complete assessment of the fuel and structural temperatures within a core to be made. More significantly, this level of resolution facilitates interactive coupling with neutronics models to enable the strong temperature/reactivity feedbacks, inherent in such cores, to be resolved correctly.
Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Depressurized Loss of Coolant Accidents
Authors: Zhengy Yanhua , Shi Lei
Published: July 2009

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.
Conceptual Design of Sulfur-Iodine Hydrogen Production Cycle of Korea Institute of Energy Research
Authors: Cho Wonchul , Bae Kikwang , Park Chusik , Kim Changhee , Kang Kyoungsoo
Published: July 2009

The Sulfur-Iodine thermochemical cycle offers a promising approach to the high efficiency production of hydrogen from nuclear power. Several SI cycles have been proposed by several research group. General Atomic (GA) studied I2 separation by extractive distillation using H3PO4. RWTH introduced the concept of reactive distillation. In this process, HIx stream coming from the Bunsen reaction is fed to the column. And HIx is distillated and decomposed at the same time to obtain hydrogen. Korea Institute of Energy Research (KIER) and Japan Atomic Energy Agency (JAEA) concentrate HIx using electro-dialysis cell and concentrated HIx is fed to the column to produce HI vapor, which is decomposed to produce hydrogen. HI was separated from HIx solution by an extractive distillation using H3PO4. However, a large amount of electric energy was required to recycle H3PO4. Most of SI processes have difficulties producing hydrogen because it has excess iodine in HI decomposition Section. SI cycle with electrodialysis cell uses membrane reactor to separate H2 and HIx. The current state of the membrane technology is not compatible with the process needs. This study examined several cases of flowsheets to overcome the problems mentioned above. The flowsheets were revised by adding the iodine separator and excluding membrane reactor. The thermal efficiency of SI process was analyzed using the revised flowsheet.
Meeting the Thermal-Hydraulic Analysis Needs for Advanced Gas Reactor Systems
Authors: Schultz Richard R.
Published: July 2009

Simulation of some fluid phenomena associated with Generation IV reactors requires the capability of modeling mixing in two- or three-dimensional flow. At the same time, the flow condition of interest is often transient and depends upon boundary conditions dictated by the system behavior as a whole. Computational fluid dynamics (CFD) is an ideal tool for simulating mixing and three-dimensional flow in system components, whereas a system analysis tool is ideal for modeling the entire system. This paper presents the reasoning which has led to coupled CFD and systems analysis code software to analyze the behavior of advanced reactor fluid system behavior. In addition, the kinds of scenarios where this capability is important are identified. The important role of a coupled CFD/systems analysis code tool in the overall calculation scheme for a Very High Temperature Reactor is described. The manner in which coupled systems analysis and CFD codes will be used to evaluate the mixing behavior in a plenum for transient boundary conditions is described. The calculation methodology forms the basis for future coupled calculations that will examine the behavior of such systems at a spectrum of conditions, including transient accident conditions, that define the operational and accident envelope of the subject system. The methodology and analysis techniques demonstrated herein are a key technology that in part forms the backbone of the advanced techniques employed in the evaluation of advanced designs and their operational characteristics for the Generation IV advanced reactor systems.
The Bench-Scale Facility for Fabrication of TRISO-Coated Plutonium Fuels
Authors: Kadarmentov Igor M. , Makarov Vladimir M. , Vakhrushev Yevgeny , Baranov Sergey , Koshelev Yuri , McEachern Donald W. , Noren Robert C.
Published: July 2009

The program to develop the GT-MHR [1] and its coated particle fuel for disposal of excess Russian weapons plutonium is being carried out by Russian nuclear labs and industrial organizations shown in Table 1 with support from US specialists at General Atomics and Oak Ridge National Laboratory. Rosatom of Russia and the U. S. National Nuclear Security Agency each provide 50% of the funds for the program.
Experimental Study of Jet Flow Impingement Within a Rod Bundle Using PIV Technique
Authors: Amini Noushin , Hassan Yassin A.
Published: July 2009

In this investigation Particle Image Velocimetry technique was implemented to a matched refractive index facility which was placed in a rectangular channel of L:1016 mm×W:76.2 mm×H:76.2 mm. Water was pumped into either one or both of the inlet jets which were entering the channel’s top wall with several different Reynolds numbers. The instantaneous and time-resolved velocity fields were successfully obtained from which several flow characteristics such as vorticity, turbulence instabilities and Reynolds stresses can be calculated.
EREC-STRESA: Web Database of Thermohydraulic Experiments on NPPs Safety
Authors: Blinkov Vladimir N. , Davydov Mikhail V.
Published: July 2009

There is a considerable amount of resources devoted to experimental databases in the field of NPPs safety at the international level. Most of them have no data concerning light water reactors of Russian type. EREC has gained a large scope of experimental data for such reactors. This paper presents the EREC-STRESA web-based database. It is the unique most complete database resulted from thermohydraulic experiments dedicated to analysis of Russian LWR NPPs safety.
Simulations of Metal Oxidation in LBE at a Mesoscopic Level
Authors: Tan Taide , Chen Yitung
Published: July 2009

The oxidation process of metals in lead bismuth eutectic (LBE) environment is studied at a mesoscopic scale. An improved stochastic cellular automaton model based on an improved Moore neighborhood is proposed to investigate the development of a continuous oxide layer of metals in LBE. The ionization of metal and the oxidation reaction were simulated with consideration of the transport of oxygen along the grain boundary and the diffusion of metallic ions. The growth of oxide layer in two directions is observed and the volume expansion effect can be realized by changing the volume control parameter. The model was benchmarked with a pure diffusion process, both with the analytical solution and with the previous work. Significant agreement was reached between the data. The developed model is also mapped with the experimental data from an LBE loop. A parametric study was conducted in order to check the importance of the main explicit parameters of the mesoscopic model.
Evaluation of Cross-Weld Properties of Austenitic Stainless Steels Before and After PWHT
Authors: Ibrahim M. M. , Mohamed H. G. , Tawfik Y. E. , Taha Ibrahim
Published: July 2009

Different types of austenitic stainless steels, which are commonly used for piping systems, tanks, and vessels, required postweld heat treatment (PWHT), at temperatures between 540 and 590 °C, regardless of the plate thickness. This paper reports on the weld procedures and cross-weld performance evalution of weldments in 6 mm AISI 304L, 316L, and 347 steel plates before and after PWHT. This welds were produced by SMAW and GTAW techniques using a single vee preparation and multiple weld beads, and welded by various types of consumables. After PWHT, tensile tests indicated a reduction in the ultimate tensile strength of all samples and a decrease in the yield strength for some cases only. The hardness results were consistent with the tensile test results because they both revealed significant softening in the HAZ and WM as a result of PWHT. In spite of the fact that PWHT exerts a beneficial effect on reducing residual stresses, it is concluded that the ductility of the weld region was satisfactory without PWHT, and PWHT decreased the cross-weld tensile strength.
Oxidation of Advanced Zirconium Alloys in Oxygen in the Temperature Range 600-1600°C
Authors: Steinbru¨ck Martin
Published: July 2009

The oxidation kinetics of the classic pressurized water reactors (PWR) cladding alloy Zircaloy-4 has been extensively investigated over a wide temperature range. In recent years, new cladding alloys optimized for longer operation and higher burn-up are being increasingly used in Western light water reactors (LWR). These alloys were naturally optimized regarding their corrosion behavior for operational conditions. The publicly available data on high temperature oxidation of the various cladding materials are very scarce. Therefore, at FZK an experimental program on the high-temperature behavior of different cladding alloys presently used has been started. This paper presents the results of a first test series with Zircaloy-4 as reference material, Framatome Duplex cladding, Framatome M5 and the Russian E110 alloy. The first two are Zr-Sn, the latter two Zr-Nb alloys. All materials were investigated in isothermal and transient tests in a thermal balance under argon-oxygen atmosphere. Strong and varying differences (up to 500%) of oxidation kinetics between the alloys were found till 1000 °C, where the breakaway effect plays a role. Smaller but still significant differences (20–30%) were observed at higher temperatures. Generally one can say that the advanced cladding alloys here studied show also a favorable behavior at high temperatures during accident scenarios.
Corrosion Behavior of Carbon Steel for the Overpack in the Groundwater Containing Bicarbonate Ions
Authors: Nishimura Toshiyasu , Dong Junpha
Published: July 2009

Carbon steel is considered in Japan the candidate material for overpacks in high-level radioactive waste disposal. Effects of bicarbonate solutions on the corrosion behavior and corrosion products of carbon steel were investigated by electrochemical measurements, FT-IR and XRD analyses. The anodic polarization measurements showed that bicarbonate ions (HCO3−) accelerated the anodic dissolution and the outer layer film formation of carbon steel in the case of high concentrations, on the other hand, it inhibited these processes in the case of low concentrations. The FT-IR and XRD analyses of the anodized film showed that siderite (FeCO3) was formed in 0.5 to 1.0mol/L bicarbonate solution, and Fe2(OH)2CO3 in 0.1 to 0.2mol/L bicarbonate solution, while Fe6(OH)12CO3 was formed in 0.02 to 0.05mol/L bicarbonate solutions. The stability of these corrosion products was able to be explained by using the actual potential – pH diagrams for the Fe-H2O-CO2 system.
Study on Condensation Heat Transfer Under High-Temperature, High-Pressure Conditions
Authors: Yuasa H. , Abe N. , Ono H. , Shirakawa K. , Morooka S.
Published: July 2009

Knowing the predicted overpressure rate under anticipated operational occurrences (AOOs) is very important when evaluating the integrity of a BWR reactor pressure vessel. One of the factors that influence the overpressure rate is the wall condensing performance. Many condensing studies have been done under low-pressure conditions without vapor flow, but few condensing test results under BWR conditions have been reported. Therefore, the purposes of this study were to extend the vapor condensing data base for high-pressure, high-temperature conditions, to improve the heat transfer correlation and to evaluate the condensing effect on the overpressure rate. Condensation heat transfer tests have been performed with the pressure range from 0.5 to 8 MPa under upward and downward vapor flow. The test section consisted of a condensing tube and a water-cooling jacket. The condensing tube was a circular tube. The test results showed that the condensation heat transfer coefficient increased with the velocity of vapor flow due to enhancement of heat transfer caused by turbulence of the liquid film. We obtained a new correlation for condensation heat transfer that considered vapor shear force and condensate film Reynolds number. This new correlation agreed well with experimental data over a wide range of pressure. New correlation was incorporated into TRACG02modT1 code. When the condensation heat transfer tests were analyzed using this modified TRACG02modT1 code, the calculated condensation heat transfer coefficients were found to be in considerable agreement with the measured data. Furthermore, when the main steam isolation valve AOO (safety relief valve capacity design) of the BWR plant was evaluated by this modified TRACG02modT1 code, we found that the vapor condensation effect appeared under relatively high-pressure conditions and the pressure with improved condensation model was lower than that without vapor condensation. In summary, the condensation heat transfer model of TRACG02modT1 code has been improved based on high-pressure, high-temperature condensation test data with vapor flow. The vapor condensation effect was found to be strong, especially in the pressure increase AOO of the actual plant.
The Main Changes and Consequences by Burnup Extensions in PWRs: Some Recommendations for Negative Plant Performance Issue Limitations
Authors: Dobrevski Ivan D. , Zaharieva Neli N.
Published: July 2009

The higher duty cores are always attended with the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of core. The impacts of these factors on the chemical environments of the nuclear fuel cladding- and reactor coolant system- surfaces are discussed: The SNB at the cladding surfaces strongly influences the behavior of gases dissolved in cladding water layer phase, causing their stripping out which will favor the creation of oxidizing conditions. It is considered that the creation of oxidizing conditions in the nuclear fuel cladding environment is not a direct boiling consequence but it is a result of the synergic impact of the boiling- and water radiolysis- processes on the Pressurized Water Reactor fuel cladding surface areas. This requires new criteria for corrosion resistance properties of fuel cladding materials. These materials must have high corrosion resistance to oxidizing conditions. If only boric acid as neutron absorber is implemented, the critical boron concentration in coolant increases significantly when implementing an extended fuel cycle. This effect can be limited through application of 10B enriched boric acid or by application of such solid burnable absorbers as Gd2O3 integrated in UO2 fuel. Some recommendations to address negative plant performance issues are given in addition to the results of these investigations.
Optimal Preventive Maintenance Policy Under a Budget Constraint
Authors: Singh Inderjeet , Popova Elmira , Kee Ernie
Published: July 2009

We design an optimal preventive maintenance policy for a system of N items that minimizes the total expected maintenance cost. We assume that the budget for preventive maintenance is limited and constrained. The problem has a finite time horizon and we consider constant inter-preventive maintenance times for every item. The resulting nonlinear optimization problem is reformulated as a binary integer program and computation results are presented on a real data set from South Texas Project Nuclear Operating Company in Bay City, Texas, USA.
Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads Due to Aircraft Engine Crash
Authors: Shirai Koji , Namba Kosuke , Saegusa Toshiari
Published: July 2009

In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters and seismic tests subjected to strong earthquake motions. Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001. This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are developed and calculated. Main criteria for estimating the maximum leakage rate for the lid metallic seal system are no loss of the pre-stress of the lid bolts, no appearance of the plastic region between the metal seal flanges, and no large relative deformation of the lid seals. Finally, in both cases, the low leakage rate for the metal cask lid closure system under the impulsive loads due to aircraft engine crash will be proved thoroughly.
A Modal Analysis Procedure for PWR Fuel Assemblies
Authors: Carrilho Le´o A. , de Gouveˆa Jayme P.
Published: July 2009

The analysis of vibration effects on the structure of Nuclear Fuel Assemblies was performed by using a 3-D finite element model. A 16×16 Standard Fuel Assembly design was used for this analysis. Since fuel assemblies have a nonlinear behavior during normal operation, due to the interaction between fuel rods and spacer grids, the methodology consisted basically of four steps: a modal analysis for obtaining the natural frequencies, a linear structural analysis for the no sliding stiffness calculation, a nonlinear structural analysis for the sliding stiffness calculation, and a correction factor estimate for natural frequencies obtained directly from the modal analysis. Also, it includes the major calculated stresses associated with the most external thimble tube.
Implementation and Validation of Downcomer and Upper Tie Plate CCFL Correlations in a Two-Fluid Code
Authors: Ha¨nninen Markku
Published: July 2009

At the moment the two-fluid system code APROS has CCFL (Counter Current Flow Limitation) correlations that are designed only for a single pipe or for bundle geometry. In the reactor pressure vessel the downcomer and the upper tie plate of the core are components that call for special CCFL correlations. In the present task the Glaeser CCFL correlations for a downcomer and an upper tie plate have been implemented in the code. The implemented correlations have been validated by calculating several downcomer and upper tie plate test cases of the UPTF (Upper Plenum Test Facility). In the tests the steam flow, the emergency core cooling flows, injection locations and the pressure levels were varied. In the validation the UPTF facility has been modeled with a nodalization, which is normally used in the corresponding calculations. Because the calculation results do not depend merely on the CCFL correlations the new Kutateladze coefficients of the Glaeser correlation had to be specified during the validation. With the selected coefficients, good, slightly conservative results were obtained. In the paper the work done with the correlations and the results of the validation calculations are described.
Strengths and Weaknesses of Japan’s Nuclear Knowledge Management: From Field Survey of All Nuclear Power Stations in Japan
Authors: Yamamoto Shinji , Kitamura Toshiro
Published: July 2009

Until around 1990 Japanese nuclear power generation maintained the world’s top-level safety and high capacity factors. However, repeated accidents and troubles in the latest decade (including those experienced earlier and newly uncovered) caused the capacity factors decline, and simply the regulatory authorities reinforce their rules. Nevertheless, Japan has maintained its nuclear technologies and knowledge by continued constructions of new nuclear power plants (NPP) and the technology and knowledge were handed down and accumulated., while the construction decreased drastically worldwide after the Chernobyl accident in 1986. It should be meaningful to see how Nuclear Knowledge Management (NKM) worked over this period. All 17 nuclear power stations in Japan were visited from January to September 2007 for field survey of maintenance practices, human resources development programs and quality assurance activities. From the analyses of the collected information, the strengths and weaknesses in nuclear knowledge management have been induced. The lessons we concluded are elaborated in the following subjects. • Pluses (+) and minuses (-) of multi-layered structures of nuclear industry: its correlation with NKM (accumulation, preservation and creation of technologies and knowledge); • Technology improvement and knowledge accumulation in the new quality assurance systems; • “Olds” and “News” in transferring technologies and knowledge, and in developing human resources; • Rebuilding of on the job training (OJT) and off the job training (Off - JT): its contents of lessons and training; • Appropriate use of the manuals: to know its usefulness and limits; • Appropriate deployment of the inspection enforcement: incentives and formalities; • Motivation of active information sharing by the adequate intellectual property management; • Lateral development of best practices, and new value creation by the shared knowledge. The Japanese electric utilities and nuclear industry should further advance their information sharing of valuable experiences such as troubles and good practices, and strengthen their self-management. Thus, they will be able to contribute, with their advanced strengths of accumulated/preserved nuclear technologies and knowledge, to meeting the growing needs of new NPP construction worldwide in the Nuclear Renaissance. NKM will in turn be further advanced on the so-called spiral-up track.
Seismic Analysis Approach Applied to a Small Size Next Generation Nuclear Reactor
Authors: Forasassi G. , Lo Frano R.
Published: July 2009

The aim of the paper is to evaluate the behaviour of a Near Term nuclear energy system example with reference to IRIS (International Reactor Innovative and Safety) project. As it is well known the development of new and future-generation nuclear power plant (Gen IV NPP) is strictly related to the sustainability, safety and reliability as well as to the proliferation resistance. In this paper, the safety aspects related to the effects of a severe earthquake (Safe Shutdown Earthquake) as well as to the induced loads are treated by means the Substructure and Time History Approaches, assuming a free field Peak Ground Acceleration equal to 0.3 g as input motion. The analyses and upgrading of the geometry structures with highest probability of criticality are performed on rather complex and detailed 3D finite element (FE) models. The main goals were: the evaluation of the dynamic characteristics of each considered structure, the verification of the load bearing structures in order to obtain a preliminary assessment of the adopted methodological approach and structural models. The analyses results and dynamic response of internal components (e.g. Nuclear Buildings, etc.) seem to confirm the possibility to upgrade the geometry and the performances of the proposed design choices.
Risk Managed Technical Specification Implementation at South Texas Project Units 1 and 2
Authors: Kee Ernie , Richards Drew , Grantom C. Rick , Liming James K.
Published: July 2009

Implementation of Risk Managed Technical Specification (RMTS) Initiative 4b, Risk-Informed Completion Times, for the two 3800 MW pressurized water reactors operated at the South Texas Project is described. Implementation issues encountered are presented and their resolutions described. The STP approach using a software application RICTCal to calculate risk informed completion times (RICTs) on the existing enterprise database to calculate risk informed completion times (RICTs) is described. Advantages, disadvantages, and lessons learned of the RMTS implementation are discussed. The methodology used to calculate completion times and maintain the database is described and a brief description of the developed documentation (procedures and guidance documents) for RMTS implementation at STP is given.
Heat Transfer Boundary Conditions in the RELAP5-3D Code
Authors: Riemke Richard A. , Davis Cliff B. , Schultz Richard R.
Published: July 2009

The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.
Local Flow Structure of Subcooled Boiling Flow of Water in a Heated Annulus
Authors: Lee Tae-Ho , Yun Byong-Jo , Park Goon-Cherl , Hibiki Takashi , Kim Seong-O
Published: July 2009

Local measurements of flow parameters were performed for vertical upward subcooled boiling flows in an internally heated annulus. The annulus channel consisted of an inner heater rod with a diameter of 19.0 mm and an outer round tube with an inner diameter of 37.5 mm, and the hydraulic equivalent diameter was 18.5 mm. The double-sensor conductivity probe method was used for measuring the local void fraction, interfacial area concentration, bubble Sauter mean diameter and gas velocity, whereas the miniature Pitot tube was used for measuring the local liquid velocity. A total of 32 data sets were acquired consisting of various combinations of heat flux, 88.1–350.9 kW/m2, mass flux, 469.7–1061.4 kg/(m2s) and inlet liquid temperature, 83.8–100.5 °C. Six existing drift-flux models and six existing correlations of the interfacial area concentration were evaluated by the data obtained in the experiment.
Liquid Metal Lithium Jet Experiment for IFMIF Target
Authors: Kanemura Takuji , Kondo Hiroo , Yamaoka Nobuo , Miyamoto Seiji , Ida Mizuho , Nakamura Hiroo , Matsushita Izuru , Muroga Takeo , Horiike Hiroshi
Published: July 2009

The present paper reports experimental study on the liquid lithium (Li) target flow for the International Fusion Materials Irradiation Facility (IFMIF). Experiments on the Li flow were conducted with a liquid lithium circulation facility, with producing a 10 mm thick and 70 mm wide flat plane jet flow along a 70 mm wide horizontal straight channel at velocity of up to 15 m/s. The nozzle has a cross section of 1/2.5 scale of the IFMIF design. Free surface of the Li jet was photographed with using a CCD camera of wide dynamic range, and a stroboscopic light source. Variation of surface shape against velocity was revealed. In order to investigate the effect of the free surface wake generated at the nozzle corner on the surface shape, measurement of the thickness distribution and numerical calculation were attempted. Comparison between experimenral results, numerical calculation and a simplified model based on a ship wake showed that the surface wake generated at the nozzle corner had an insignificant effect on the surface shape of the D+ beam irradiation region in IFMIF.
Use of Similarity Analysis on Experiments of Different Size to Predict Critical Cooling Rates for Large Ceramic Waste Forms
Authors: Bateman Kenneth J. , Solbrig Charles W.
Published: July 2009

Ceramic waste forms (CWF) are produced to store fission products for the long term. They are cast into cylindrical shape at high temperature (925°C). Rapid cooling of the product is desirable for product turnaround, but cooling has the potential to crack the coalesced product into many pieces due to thermal stress. This paper investigates the rapid-cooling process with a borosilicate-glass component of the CWF used as a surrogate. The critical cooling rate of formed cylinders (the rate which separates the damage from the no-damage region) has been determined. This paper extends previous experimental data and analysis to production temperature as a step in the extrapolation of the data to production CWF’s. The glass solidifies in the range of 650°C to 625°C. The previous tests (7.8-cm diameter) were all run starting from a solid (625°C or less) to provide a basis for the higher temperature cases. Thermal stress cannot build up until solidification begins to occur. The current tests (7.8 and 9.9cm diameter) were run from the liquid temperature of 925°C. A theoretical model has been developed to analyze the data. The model includes heat transfer and the stress developed from the thermal gradients. Similarity analysis based on this model is used to produce dimensionless charts which allow data of different initial temperatures and diameters to be analyzed. The new data corroborated the previous estimate of the critical cooling rate and analytical-model projection for the minimum in-furnace cooling times for two production size CWF’s that will be stored in Yucca Mountain (70 hours for the 52-cm diameter and 35 days for the 181.5-cm sizes). To further reduce these times, an analytical prediction was made which shows that the formed cylinder can be removed from a furnace at a temperature of 320°C without any danger of cracking.
Structural Optimization Under a Multiple Loading Condition Based on the ASME Code Section III Stress Categories
Authors: Choi Woo-Seok , Seok Ki-Seog
Published: July 2009

The structural optimization technique is well known as one of the tools to achieve a better design. Recently, it has been applied to the design of nuclear components for the purpose of a design evolution and enhancement. When it is applied to nuclear components, there are a couple of points to be considered. The ASME code section III, design by analysis, provides stress criteria for nuclear components, which is quite different from the general stress constraints for a typical optimization process. Based on the ASME code section III, stresses are categorized into three types with different stress limits. The stress limits change in accordance to corresponding service levels. Since the types of stresses and the service levels introduce different stress, the multiple loading conditions should be utilized during an optimization. And the general stress constraints during an optimization are the maximum nodal stresses applied to the concerned points rather than the concerned section. On the contrary, the failure condition of the ASME code section III can be evaluated by comparing the stress intensity, which was created by the stress linearization procedure at the concerned section, with the prescribed one. So, the linearized stresses such as the stress intensities are used as stress constraints during an optimization for nuclear components. In order to adopt the stress linearization procedure to the constraints during an optimization, an extra data manipulation such as calculating a linearized stress and a derivative of the linearized stress needs to be conducted. In this paper, a structural optimization technique is proposed for an optimization application of a nuclear-related structure. This proposed optimization technique mainly includes a multiple loading condition and a stress linearization procedure to satisfy the requirements of the ASME code section III. To validate the ideas, one of the nuclear-related structures is selected and optimized under multiple loading conditions with linearized stress constraints.
Steady-State Thermal-Hydraulic Analysis of a Plate Type Reactor
Authors: Lu Qing , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: July 2009

In the plate type fuel reactor, all the fuel elements are arranged in parallel in the subassembly and set in the core, which create many isolated parallel rectangular channels. The arrangements of the fuel elements may have significant impact on the thermal-hydraulics of the core. In order to fully understand the geometry impact on the steady-state thermal-hydraulic characteristics, based on the mass, momentum and energy conservation equations, a multi-channel code was developed to estimate the steady-state thermal-hydraulic characteristics. Two different types of fuel elements arrangements in the subassembly were proposed and studied, one was uniform distribution with the same gap sizes between each fuel elements, and the other was nonuniform distribution with different clearances among boundary fuel elements. The mass flux distribution, the temperature field and the DNBR were obtained by the developed code. It showed that the mass flux distribution was affected by the flow area of the channel and the given power distribution. The flow area of the channel has bigger influence than the second. In addition, the results of the two different types of fuel elements arrangements were compared, and it could be found that the asymmetrical cooling of the fuel elements occurred when fuel elements were nonuniformly arranged, which leaded to the decrease of the maximum temperature in the asymmetrically cooled fuel elements and the increase of the maximum temperature in the core. Furthermore the DNBR calculated by the Sudo model indicated that the safety margin of the reactor under the present conditions is sufficient. All the obtained results served some valuable information for the design of the new type research reactor.
Laser Peening Systems for Preventive Maintenance Against Stress Corrosion Cracking in Nuclear Power Reactors
Authors: Uehara Takuya , Yoda Masaki , Sano Yuji , Mukai Naruhiko , Chida Itaru , Kato Hiromi
Published: July 2009

Laser peening introduces compressive residual stress on metal surface by irradiating laser pulses underwater without any surface preparations. The process utilizes the impulsive effect of high-pressure plasma generated through ablative interaction of each laser pulse with material. Laser peening systems, which deliver laser pulses with mirrors or through an optical fiber, were developed and have been applied to preventive maintenance against stress corrosion cracking (SCC) in nuclear power reactors since 1999. Each system was composed of laser oscillators, a beam delivery system, a laser irradiation head, remote handling equipment and a monitor/control system. Beam delivery with mirrors was accomplished through alignment/tracking functions with sufficient accuracy. Reliable fiber-delivery was attained by the development of a novel input coupling optics and an irradiation head with auto-focusing. At present, we are developing a newer concept and the prototype system has been just completed, which is extremely small, reliable and easy-handled.
Digital Systems for Reactor Control: Design, Experience of Operation
Authors: Bakhmach Ievgenii S. , Siora Alexander A. , Bezsalyi Volodymyr T. , Yastrebenetsky Mikhail A.
Published: July 2009

Conversion of traditional analog NPP I&C systems to digital systems is a common tendency for many countries. Digital systems for reactor control designed by «Radiy» Company (Kirovograd, Ukraine) are described below. FPGA (Field Programmable Gates Arrays) were used for implementation of control algorithms. An equivalence between FPGA-projects implementation and schemes of control technological algorithms permitted to simplify development and verification processes and decrease the number of development errors. The platform was used for implementation of different safety important systems: reactor protection systems, automatic reactor power control and limitation systems, rods control systems, control safety systems. The main peculiarity of the reactor protection system is different types of diversity (apparatus and program diversity due to different hardware and different languages in main and diverse divisions; functional diversity; difference of CASE-tools). These systems have been used at 10 units. Reliability measures of systems and their components were determined using operational statistical data. Possible uses for these systems: modernization of different types of existing reactors (not only WWER); as full system or as subsystems; not only for Ukraine, but for other countries; for reactors III+ and IV generations.
Analytical and Computational Analysis of Turbulent Bouyant Jets in the Containment Atmosphere
Authors: Deri Enrico , Bucci Matteo , Studer Etienne , Abdo Daniele
Published: July 2009

In case of severe accident, complex thermal-hydraulics phenomena are expected to occur in the containment atmosphere. To investigate and understand these phenomena, fundamental for nuclear safety and design, major efforts are being spent all over the world. A new OECD project, named SETH-2, is conceived to generate relevant experimental data, useful to improve the modeling capabilities of the computer codes aimed to predict post-accident containment thermal-hydraulic conditions. The Commissariat a` l’Energie Atomique (CEA) contributes to the project performing experiments within the large scale MISTRA facility. Tests are proposed to investigate mixing phenomena promoted in a stratified containment. In particular, one of these test series concerns the interaction of buoyant jets with a stratified atmosphere. The present work is aimed to develop and validate computational tools useful to support the design of this experimental campaign and to analyze the actual MISTRA tests. In this aim, two different models have been implemented for turbulent buoyant jets in a stratified atmosphere: an engineering analytical model for a fast characterization of flow structures and a finite elements computational fluid dynamics (CFD) model that allows a detailed analysis of local phenomena. The models have been successfully validated for vertical buoyant jets in uniform atmosphere. Further experimental and numerical activities are illustrated, aimed to carry out the validation with stratified atmosphere and inclined injections.
Computational Fluid Dynamics of Supercritical Carbon Dioxide Turbine for Brayton Thermodynamic Cycle
Authors: Jeong Wi S. , Kim Tae W. , Suh Kune Y.
Published: July 2009

The supercritical gas turbine Brayton cycle has been adopted in the secondary loop of the Generation IV Nuclear Energy Systems, and also planned to be installed in power conversion cycles of the nuclear fusion reactors. Supercritical carbon dioxide (SCO2) is one of widely considered fluids for this application. The potential beneficiaries include the Secure Transportable Autonomous Reactor - Liquid Metal (STAR-LM), the Korea Advanced Liquid Metal Reactor (KALIMER), and the Battery Omnibus Reactor Integral System (BORIS) which is being developed at the Seoul National University. The reason for these welcomed applications is that the SCO2 Brayton cycle can possibly achieve higher energy conversion efficiency than the steam turbine Rankine cycle. Gas turbine design is crucial part in achieving this high efficiency. In this paper, a one-dimensional gas turbine analysis methodology is applied for optimal design of the component. Case study result shows that the entire turbine efficiency is increased as hub radius is increased for a same number of stage conditions. Comparing the efficiency which is applied the boundary condition, 4 stage turbines have optimal efficiency.