bluuebunny/crossref_metadata_2025 huggingface dataset:
Gas Driven Circulation in an Adiabatic Air-Water Loop With Reference to Accelerator Driven Systems
Authors: De Salve Mario , Malandrone Mario , Panella Bruno
Published: March 2009

In order to investigate the fluid dynamic mechanisms of the gas driven circulation, with reference to the accelerator driven systems (ADS), an experimental research with water driven by air injected at the inlet of the vertical riser in an adiabatic test rig has been carried out. The present experimental apparatus consists essentially of two vertical plexiglass pipes 3.8 m long, 0.08 m I.D., that are connected at the bottom by a horizontal pipe and at the top by a large open tank. The test rig is equipped with differential pressure transducers and two quick closing valves to measure the pressure drops and the average void fraction. The air flow rate ranges up to 5000 Nl/h, at 7 bar pressure, and the liquid flow rate can reach the value of more than 5 kg/s. The research aim is to investigate the effect of the superficial gas velocity on the superficial liquid velocity, on the void fraction and on the pressure drops, for a given geometry. The test data have been compared with some two-phase pressure drops and void fraction correlations. The geometry effect has been investigated too.
Neutronic Analysis for Tehran Research Reactor Upgrading
Authors: Afshar Ebrahim , Shahidi Alireza
Published: March 2009

This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of Tehran Research Reactor (TRR) from 5 to 10 MW (th), keeping the same core configuration and with minimum changes in the primary cooling circuit. The main aim of TRR upgrade is to increase the volume of radioisotope production. The neutronic analysis was carried out for a fresh core with 22 Standard Fuel Elements (SFE) under normal operating conditions. Two different calculational lines were used to simulate the neutronic behavior in the core and perform the necessary neutronic calculations. First, combination of cell calculation transport code WIMS-D/4 [1] and three dimensional core calculation diffusion code CITATION [2] were used to and next a Monte Carlo code MCNP-4B [3] together with point depletion code ORIGEN-2 [4] were used. The results obtained show good agreement between these two different schemes.
BWR 9×9 Fuel Assembly Thermal-Hydraulic Tests: 1 — New Formula for Post Boiling Transition Heat Transfer Correlation and Rewet Correlation
Authors: Tsukuda Yoshiaki , Hayashi Hiroshi , Isaka Koujun , Kaneko Hirohisa , Mitsutake Toru , Akiba Miyuki , Abe Nobuaki , Morooka Shinichi , Masuhara Yasuhiro , Suzuki Riichiro , Nishino Yuji
Published: March 2009

Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8×8 and 9×9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, post boiling transition (post-BT) tests by the use of full-scale test assemblies simulating 9×9 fuel were carried out to evaluate BWR fuel integrity. Two kinds of post-BT tests, the steady-state post-BT test and transient post-BT test were performed, and the test results were applied to development of new heat transfer and rewet correlations for the estimation of fuel rod surface temperature.
Study of Mechanism of Initial Crud Formation on Fuel Cladding in Subcooled Boiling Region in PWR
Authors: Rao Qinyang , Jones Barclay G.
Published: March 2009

A hemispherical vapor bubble model has been constructed, and thermocapillary flow driven by surface tension gradient, due to temperature difference, around bubble attached to a plane heated surface, and it has been examined using numerical methods. LiOH solute concentration distribution around the bubble has been obtained. Some parameters under PWR operating conditions have been discussed: bubble size, mass diffusion coefficient and vapor heat transfer coefficient. Effects of some important parameters, such as heat flux and bubble size, have been examined. The study shows that the thermocapillary flow around a static hemispherical vapor bubble can not generate sufficient solute concentrations to initialize crud formation under PWR operating condition.
Remote-Controlled Inspection Robot for Nuclear Facilities in Underwater Environment
Authors: Miwa Yasuhiro , Satoh Syuichi , Hirose Naoya
Published: March 2009

A remote-controlled inspection robot for nuclear facilities was developed. This is a underwater robot technology combined with inspection and flaw removal technologies. This report will describe the structure and performance of this robot. The inspection robot consists of two parts. The one is driving equipment, and the other is inspection and grinding units. It can swim in the tank, move around the tank wall, and stay on the inspection area. After that it starts inspection and flaw removal with a special grinding wheel. This technology had been developed to inspect some Radioactive Waste (RW) tanks in operating nuclear power plants. There are many RW tanks in these plants, which human workers can be hard to access because of a high level dose. This technology is too useful for inspection works of human-inaccessible areas. And also, in conventional inspection process, some worker go into the tank and set up scaffolding after full drainage and decontamination. It spends too much time for these preparations. If tank inspection and flaw removal can be performed in underwater, the outage period will be reduced. Remote-controlled process can be performed in underwater. This is the great advantage for plant owners. Since 1999 we have been applying this inspection robot to operating nuclear 11 facilities in Japan.
Large-Scale Parallel Finite Element Analysis of the Stress Singular Problems
Authors: Kushida Noriyuki , Okuda Hiroshi , Yagawa Genki
Published: March 2009

In this paper, the convergence behavior of large-scale parallel finite element method for the stress singular problems was investigated. The convergence behavior of iterative solvers depends on the efficiency of the preconditioners. However, efficiency of preconditioners may be influenced by the domain decomposition that is necessary for parallel FEM. In this study the following results were obtained: Conjugate gradient method without preconditioning and the diagonal scaling preconditioned conjugate gradient method were not influenced by the domain decomposition as expected. symmetric successive over relaxation method preconditioned conjugate gradient method converged 6% faster as maximum if the stress singular area was contained in one sub-domain.
Pore-Scale Simulation for Predicting Material Transport Through Porous Media
Authors: Itoh Goichi , Nakamura Jinya , Kono Koji , Watanabe Tadashi , Ohashi Hirotada , Chen Yu , Nagasaki Shinya
Published: March 2009

Microscopic models of real-coded lattice gas automata (RLG) method with a special boundary condition and lattice Boltzmann method (LBM) are developed for simulating three-dimensional fluid dynamics in complex geometry. Those models enable us to simulate pore-scale fluid dynamics that is an essential part for predicting material transport in porous media precisely. For large-scale simulation of porous media with high resolution, the RLG and LBM programs are designed for parallel computation. Simulation results of porous media flow by the LBM with different pressure gradient conditions show quantitative agreements with macroscopic relations of Darcy’s law and Kozeny-Carman equation. As for the efficiency of parallel computing, a standard parallel computation by using MPI (Message Passing Interface) is compared with the hybrid parallel computation of MPI-node parallel technique. The benchmark tests conclude that in case of using large number of computing node, the parallel performance declines due to increase of data communication between nodes and the hybrid parallel computation totally shows better performance in comparison with the standard parallel computation.
Feasibility Investigation of the Decay Heat Removal Capability Using the Concept of a Thermosyphon in the Liquid Metal Reactor
Authors: Kim Yeon-Sik , Sim Yoon-Sub , Kim Eui-Kwang
Published: March 2009

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced LIquid MEtal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20∼40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design.
The IRIS Spool-Type Reactor Coolant Pump
Authors: Kujawski J. M. , Kitch D. M. , Conway L. E.
Published: March 2009

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called “spool” pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper.
Technology Perspectives on the Management of Spent-Resin Wastes Generated From Nuclear Power Reactor Operations
Authors: Vijayan Shiv , Kikuchi Makoto , Komatsu Akihiro
Published: March 2009

Organic-resin wastes (spent resins) are generated by different purification systems employed in all types of nuclear power reactors during routine and non-routine operations. The quantities of such resin wastes, and their inventories of contaminants vary depend on the operational goals of the individual power plant. Depending on the regulatory target in the particular jurisdiction where the reactor is located, the type and amounts of radionuclides, metals and other chemical contaminants in the resin waste determine the extent of treatment required for interim storage or final disposal of the waste. Resin-waste treatment comprises different operations such as pretreatment, conditioning/stabilization and containerization that produce a waste package suitable for handling, transport, storage and disposal. One aspect of the contaminants that has significant impact on waste conditioning and the overall cost of managing such wastes are the concentrations of short half-life (arbitrarily less than approximately 30 years) radionuclides, and long half-life radionuclides, in particular carbon-14, and toxic metals present in the waste. A spectrum of resin-waste conditioning methods is available. Some methods have been applied to specific situations while others are being developed for future applications to meet the need for reducing worker dose, environmental releases, and waste-storage and disposal costs. This paper describes waste treatment options for low-level radioactive resin wastes and potential options of resin wastes containing appreciable amounts of carbon-14. Indications are that drying of the resin waste containing long half-life radionuclides such as carbon-14 and compaction or pelletizing can be favourable to allow interim dry-storage of the waste and to provide sufficient flexibility in the preparation of a suitable waste form to meet applicable waste acceptance criteria for the eventual disposal of such wastes.
Development of Plus Ultrasonic Doppler Method for Flow Rate Measurements of Power Plant: Multiline Flow Rate Measurement for Non Developed Flow
Authors: Kikura Hiroshige , Wada Sanehiro , Aritomi Masanori , Mori Michitsugu , Takeda Yasushi
Published: March 2009

Flow metering system is being developed using plus ultrasonic Doppler method. The principle is an integration of instantaneous velocity profile over a pipe diameter so that it is expected to be able to eliminate installation problems such as entry length as well as to follow transient flow rate precisely. Flow metering principle by plus ultrasonic Doppler method in a circular pipe depends on the alignment of measuring lines. In this paper, influence of number of measuring lines on the flow rate measurements for power plant have been investigated for non developed flows in a vertical pipe.
Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components
Authors: Shah V. N. , Liu Y. Y.
Published: March 2009

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of the AMPs, which do not require further evaluation, do need enhancements to allow for an extended period of reactor operation. The paper concludes that the GALL report systematically evaluates the generic AMPs for their effectiveness during an extended period of operation and provides a technical basis for their acceptance. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.
Prediction of MIT Pressurizer Data Using RELAP5 and TRAC-M
Authors: Shumway Rex , Bolander Mark , Aktas Birol
Published: March 2009

Tests simulating Pressurized Water Reactor pressurizers under inflow and outflow conditions have been performed at MIT. Prediction of pressurizer pressure requires accurate models of wall heat transfer as well as interfacial liquid-steam heat transfer. The US NRC has two computer programs used for predicting thermal hydraulic behavior in reactors; RELAP5 and TRAC-M. TRAC-M is the Consolidated Thermal-hydraulics Code developed by combining models from TRAC-B and RELAP5 into a modernized version of TRAC-P. The component models from RELAP5 and TRAC-B have been ported to TRAC-M but not the constitutive models. A suite of assessment cases are being developed to guide the constitutive model improvement process. Assessment against data will determine which constitutive relations need to be ported to TRAC-M. This paper compares the RELAP5 and TRAC-M codes against MIT pressurizer data. As water is injected into the bottom of the pressurizer the steam pressure in the top of the pressurizer rises. The pressure increase rate is controlled by wall and interface condensation rates. Both codes predict this complex compression process reasonably well. The effect of time step size and code options are explored in this paper. The benefits of using two codes to analyze thermal-hydraulic processes are evident from the results.
First Core and Refueling Options for IRIS
Authors: Petrovic Bojan , Carelli Mario D. , Greenspan Ehud , Milosˇevic Miodrag , Vujic Jasmina , Padovani Enrico , Ganda Francesco
Published: March 2009

The International Reactor Innovative and Secure (IRIS) is being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. The IRIS design addresses key requirements associated with advanced reactors, including improved safety, enhanced proliferation resistance, competitive electricity production cost, and improved waste management. IRIS is a modular, small/medium size (100 to 335 MWe) PWR with integral vessel configuration. Its design is based on proven LWR technology, so that no new technology development is needed and near term deployment is possible. At the same time, aim was to introduce improvements as compared to present PWRs. These opposing requirements resulted in an evolutionary approach to fuel and core design, balancing new features and the need to avoid extensive testing and demonstration programs. A path forward was devised by selecting the current fuel technology for the first IRIS core, but keeping future upgrades possible through the variable moderation fuel assembly design. This paper describes this approach and discusses core fueling options that enable achieving four-year and eight-year core lifetime.
Isotope Enrichment by Laser Stimulation Causing Condensation Repression
Authors: Kunze Jay F. , Islam K. T. S. , Nieschmidt Ernest
Published: March 2009

Sulfur isotopes in SF6 molecules have been enriched, in a laboratory experiment, using tuned laser radiation to excite a particular sulfur isotopic molecule and inhibit its condensation on the cooled annulus inside of the chamber. The evidence of enrichment was determined by examining the residual gas with a Fourier Transform lnfra Red Spectrometer. The enrichment was observed during a transient experiment in which the temperature of the condensing surface was gradually decreased, and gas pressure of the SF6 molecules was in the range of 8 Torr (0.01 atmospheres). These results show that excitation to a single excited level can create differential rates of condensation so as to achieve an enrichment factor of approximately 2.0 in a single stage.
Risk Based Inspection Pilot Study of Ignalina Nuclear Power Plant, Unit 2
Authors: Brickstad Bjorn , Letzter Adam , Klimasauskas Arturas , Alzbutas Robertas , Nedzinskas Linas , Kopustinskas Vytis
Published: March 2009

A project with the acronym IRBIS (Ignalina Risk Based Inspection pilot Study) has been performed with the objective to perform a quantitative risk analysis of a total of 1240 stainless steel welds in Ignalina Nuclear Power Plant, unit 2 (INPP-2). The damage mechanism is IGSCC and the failure probabilities are quantified by using probabilistic fracture mechanics. The conditional core damage probabilities are taken from the plant PSA.
Feasibility Study on Nitrogen-15 Enrichment and Recycling System for Innovative FR Cycle System With Nitride Fuel
Authors: Inoue Masaki , Ono Kiyoshi , Fujioka Tsuna-aki , Sato Koji , Asaga Takeo
Published: March 2009

Highly-isotopically-enriched nitrogen (HE-N2; 15N abundance ≥ 99.9%) is indispensable for a nitride fueled fast reactor (FR) cycle to minimize the effect of carbon-14 (14C) generated mainly by 14N(n,p)14C reaction in the core on environmental burden. Thus, the development of inexpensive 15N enrichment and recycling technology is one of the key aspects for the commercialization of a nitride fueled FR cycle. Nitrogen isotope separation by the gas adsorption technique was experimentally confirmed in order to obtain its technological perspective. A conventional pressure swing adsorption technique, which is already commercialized for recovering the nitrogen gas from multi-composition gas-mixture, would be suitable for recovering in both reprocessing and fuel fabrication to recycle the HE-N2 gas. A couple of the nitride fuel cycle system concepts including the reprocessing and fuel fabrication process flow diagrams with the HE-N2 gas recycling were newly designed for both aqueous and non-aqueous (pyrochemical) nitride fuel recycle plants, and also the effect of the HE-N2 gas recycling on the economics of each concept was evaluated.
Mechanistic Two Stage Fission Gas Release Model
Authors: Kim Yong-Soo , Lee Chan-Bok
Published: March 2009

In this study, a mechanistic two stages model is developed which analytically simulates the two-step diffusion processes, grain lattice diffusion and grain boundary diffusion, coupled with the bubbles trap/resolution. Mathematical manipulation reveals that the release at high burn-up depend on the ratio of the diffusivities in the both processes, i.e., α ≅ Dveff/Dgbeff where Dveff and Dgbeff are effective volume and grain boundary diffusion coefficients, respectively. Thus, the ratio α is incorporated in the time-dependent third kind boundary condition at the equivalent grain surface. This model brings forth analytical solutions of the fractional release which are identical to that of either ANS5.4 or modified ANS5.4 model when α goes to the infinity. It turns out that this model describes the release behavior well in the high burn-up fuel and puts out a comparable prediction to the solution of FRAPCON-3 model under the same condition. It is also demonstrated that the new factor α not only ease the computational treatment for the high burn-up fuel performance evaluation, but also enables us to possibly separate the burn-up enhancement from the diffusion coefficients and to easily simulate the bubble-related phenomena in the grain boundary.
An Improved Method for Postulating Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluation
Authors: Simonen F. A. , Dickson T. L.
Published: March 2009

This paper presents an improved model for postulating fabrication flaws in reactor pressure vessels (RPVs) and for the treatment of measured flaw data by probabilistic fracture mechanics (PFM) codes that are used for structural integrity evaluations. The model used to develop the current pressurized thermal shock (PTS) regulations conservatively postulated that all fabrication flaws were inner-surface breaking flaws. To reduce conservatisms and uncertainties in flaw-related inputs, the United States Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) that has resulted in data on fabrication flaws from non-destructive and destructive examinations of actual RPV material. Statistical distributions have been developed to characterize the number and sizes of flaws in the various material regions of a vessel. The regions include the main seam welds, repair welds, base metal of plates and forgings, and the cladding that is applied to the inner surface of the vessel. Flaws are also characterized as being located within the interior of these regions or along the weld fusion lines that join the regions. Flaws are taken that occur at random locations relative to the embrittled inner region of the vessel. The probabilistic fracture mechanics model associates each of the simulated flaw types with the fracture properties of the region being addressed.
Uncertainty Characterization of Reactor Vessel Fracture Toughness
Authors: Li Fei , Modarres Mohammad
Published: March 2009

To perform fracture mechanics analysis of reactor vessel, fracture toughness (KIc) at various temperatures would be necessary. In a best estimate approach, KIc uncertainties resulting from both lack of sufficient knowledge and randomness in some of the variables of KIc must be characterized. Although it may be argued that there is only one type of uncertainty, which is lack of perfect knowledge about the subject under study, as a matter of practice KIc uncertainties can be divided into two types: aleatory and epistemic. Aleatory uncertainty is related to uncertainty that is very difficult to reduce, if not impossible; epistemic uncertainty, on the other hand, can be practically reduced. Distinction between aleatory and epistemic uncertainties facilitates decision-making under uncertainty and allows for proper propagation of uncertainties in the computation process. Typically, epistemic uncertainties representing, for example, parameters of a model are sampled (to generate a “snapshot,” single-value of the parameters), but the totality of aleatory uncertainties is carried through the calculation as available. In this paper a description of an approach to account for these two types of uncertainties associated with KIc has been provided.
Receipt and Storage Issues at the TMI-2 Irradiated Fuel Storage Installation
Authors: Christensen Allan B. , Custer Kenneth , Gardner Rick , Kaylor James , Stalnaker James
Published: March 2009

In less than a year, up to 12 canisters of TMI-2 reactor fuel debris were loaded into each of 28 Dry Storage Containers (DSCs), and placed into interim storage at an Irradiated Spent Fuel Storage Facility (ISFSI) at the Idaho National Engineering and Environmental Laboratory (INEEL). Draining and drying the canisters, loading and welding the DSCs, shipping the DSCs 25 miles, and storing in the ISFSI initially required up to 3 weeks per DSC. Significant time efficiencies were achieved during the early stages, reducing the time to less than one week per DSC. These efficiencies were achieved mostly in canister draining and drying and DSC lid welding, and despite several occurrences that had to be resolved before continuing work. The ISFSI has been operated without issue since, with the exception that license basis monitoring has indicated an unusual pattern of season- and position-dependent hydrogen generation. This paper discusses some of the innovations and storage experiences for the first ISFSI designed for the storage of severely defected fuel.
Visual Observation of Vapor Film Collapse Behavior During Vapor Explosion
Authors: Abe Yutaka , Nariai Hideki
Published: March 2009

During severe accident of a light water reactor, various thermal hydraulic phenomena including vapor explosion could threaten the integrity of the containment vessel. Thermal detonation model is proposed to describe the vapor explosion. According to the model, several processes should be sequentially satisfied for the trigger phenomena of the vapor explosion. One of the most important processes for the trigger phenomena is the vapor film collapse around high temperature molten material droplets. In the present study, the vapor film collapse behavior around high temperature solid particle submerged into water was experimentally investigated by attacking a pressure pulse to the vapor film on a high temperature sold particle. The interfacial phenomena between vapor and water were measure by using a high-speed video camera of the maximum speed of 40,500 fps. The visual data obtained were processed with visual data processing techniques. That is, the average vapor film thickness was estimated, dynamic behaviors of the interfaces were analyzed with PIV technique and the interface movement was estimated with the digital auto correlation techniques from the visual data obtained. Furthermore, the transients of the temperature and pressure were simultaneously measured. The interfacial temperatures between vapor and water, and between molted liquid and water are analytically estimated by solving the heat conduction equation with the data obtained as the boundary conditions. It is clarified that vapor collapse by pressure pulse occurs homogeneously around the vapor film surface on a high temperature particle. Microscopic information are obtained from the visual data by using visual data processing technique, PIV technique and digital auto-correlation technique. At the time the vapor film surface changes to white, the saturation temperature exceeds the interfacial temperature. The microscopic vapor film collapse behavior indicates the possibility of the phase change at the vapor film collapse.
Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project
Authors: Dickson Terry L. , Malik Shah N. , Kirk Mark T. , Jackson Deborah A.
Published: March 2009

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.
Restructuring the Uranium Mining Industry in Romania: Actual Situation and Prospects
Authors: Georgescu P. D. , Petrescu S. T. , Iuhas T. F.
Published: March 2009

Uranium prospecting in Romania has started some 50 years ago, when a bilateral agreement between Romania and the former Soviet Union had been concluded and a joint Romanian-Soviet enterprise was created. The production started in 1952 by the opening of some deposits from western Transylvania (Bihor and Ciudanovita). From 1962 the production has continued only with Romanian participation on the ore deposit Avram Iancu and from 1985 on the deposits from Eastern Carpathians (Crucea and Botusana). Starting with 1978 the extracted ores have been completely processed in the Uranium Ore Processing Plant from Feldioara, Brasov. Complying with the initial stipulations of the Nuclear National Program launched at the beginning of the 1980s, the construction of a nuclear power station in Cernavoda has started in Romania, using natural uranium and heavy water (CANDU type), having five units of 650 MW installed capacity. After 1989 this initial Nuclear National Program was revised and the construction of the first unit (number 1) was finalized and put in operation in 1996. In 2001 the works at the unit number 2 were resumed, having the year 2005 as the scheduled activating date. The future of the other 3 units, being in different construction phases, hasn’t been clearly decided. Taking into consideration the exhaustion degree of some ore deposits and from the prospect of exploiting other ore deposits, the uranium industry will be subject of an ample restructuring process. This process includes workings of modernization of the mines in operation and of the processing plant, increasing the profitableness, lowering of the production costs by closing out and ecological rehabilitation of some areas affected by mining works and even new openings of some uraniferous exploitations. This paper presents the actual situation and the prospects of uranium mining industry on the base of some new technical and economical strategic concepts in accordance with the actual Romanian Program for Nuclear Energetics.
Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements
Authors: Wilkowski G. , Scott P. , Rahman S. , Rudland D. , Wolterman R. , Krishnaswamy P. , Fairbanks C.
Published: March 2009

The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors”. These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the containment building main crane dropping a heavy load on the pipe system when the reactor is in operation; gasket or seal failure; etc. All three major contributors to probabilistic failure need to be considered for redefinition of the LB-LOCA requirements.
Are We Forgetting the Lessons From the Accident at Three Mile Island Unit 2, March 1979: A Case Study
Authors: Christie Bob , Johnson David H.
Published: March 2009

The accident at Three Mile Island Unit 2 in March 1979 resulted in major changes to the way emergency procedures were written and operators were trained at nuclear commercial electric generating units. These changes had a major impact on the public health risk of nuclear electric generating units. The record over the last 20 years has been excellent. For approximately 2000 reactor years of operation since 1979, there have been no accidents equivalent to TMI Unit 2 in the USA. Other factors have had an influence on this excellent record but it is clear that more efficient emergency procedures and better operator training had a significant impact on the excellent record achieved over the last 20 plus years. Abnormal events still occur at the nuclear commercial electric generating units in the USA and these events have the potential for causing damage to the reactor core. In some cases, the emergency procedures used in abnormal events and the training received by the operators of the nuclear units have not been based on the lessons learned from the accident at Three Mile Island. The following paper describes one such case. It is clear to the authors of this paper that further changes should be made to make sure that the lessons learned from the accident at Three Mile Island Unit 2 in 1979 are implemented and not forgotten.
Development of Thermal Analysis Capability of Dry Storage Cask for Spend Fuel Interim Storage
Authors: Ko Fu-Kuang , Liang Thomas K. S. , Yang Chung-Yu
Published: March 2009

As most of the nuclear power plants, on-site spent fuel pools (SFP) of Taiwan’s plants were not originally designed with a storage capacity for all the spent fuel generated over the operating life by their reactors. For interim spent fuel storage, dry casks are one of the most reliable measures to on-site store over-filled assemblies from SFPs. The NUHOMS®-52B System consisting of a canister stored horizontally in a concrete module is selected for thermal evaluation in this paper. The performance of each cask in criticality, radioactive, material and thermal needs to be carefully addressed to ensure its enduring safety. Regarding the thermal features of dry storage casks, three different kinds of heat transfer mechanisms are involved, which include natural convection heat transfer outside and/or inside the canister, radiation heat transfer inside and outside the canister, and conduction heat transfer inside the canister. To analyze the thermal performance of dry storage casks, RELAP5-3D is adopted to calculate the natural air convection and radiation heat transfer outside the canister to the ambient environment, and ANSYS is applied to calculate the internal conduction and radiation heat transfer. During coupling iteration between codes, the heat energy across the canister wall needs to be conserved, and the inner wall temperature of the canister needs to be converged. By the coupling of RELAP5-3D and ANSYS, the temperature distribution within each fuel assembly inside canisters can be calculated and the peaking cladding temperature can be identified.
Radiotoxicity of Actinides During Transmutation in Final Stage of Atomic Power
Authors: Gerasimov Aleksander S. , Bergelson Boris R. , Myrtsymova Lidia A. , Tikhomirov Georgy V.
Published: March 2009

Characteristics of a transmutation mode in final stage of atomic power are analyzed. In this stage, transmutation of actinides accumulated in transmutation reactors is performed without feed by actinides from other reactors. The radiotoxicity during first 20 years of transmutation is caused mainly by 244Cm. During following period of time, 252Cf is main nuclide. Contribution of 246Cm and 250Cf is 5–7 times less than that of 252Cf. During 50 years of a transmutation, the total radiotoxicity falls by 50 times. Long-lived radiotoxicity decreases slowly. During the period between T = 50 years and T = 100 years, long-lived radiotoxicity falls by 3.7 times. For each following 50 years after this period, long-lived radiotoxicity falls by 3.2 times. These results corresponding to neutron flux density 1014 neutr/(cm2s) in transmutation reactor demonstrate that the final stage of a transmutation should be performed with use of high flux transmutation facilities which provide shorter time of transmutation.
New Cryogenic Steels and Design Approach for ITER Superconducting Magnet System
Authors: Nakajima H. , Hamada K. , Okuno K. , Hada K. , Tada E.
Published: March 2009

A new design code has been developed for construction and operation/maintenance of the International Thermonuclear Experimental Reactor (ITER). A superconducting magnet system is one of the key components of ITER and its design code includes new cryogenic materials and design approach with taking account of unique features of a performance of the superconducting magnet. The new materials are nitrogen strengthened austenitic stainless steels, which have a yield strength (Sy) of over 1000 MPa and fracture toughness (KIc) of over 200 MP√m at liquid helium temperature (4K). The feature of the design approach is use of the allowable stress defined by only 2/3 Sy measured at 4K. A concept and reliability of the new design approach using new cryogenic materials for the ITER superconducting magnet system are discussed in this paper.
Main Features of ITER Vacuum Vessel and Approach to Code Application
Authors: Nakahira M. , Takeda N. , Hada K. , Tada E. , Miya K. , Asada Y.
Published: March 2009

The ITER Vacuum Vessel (VV) is a complex torus-shaped configuration with a double-walled structure and the main function is to provide a high degree of vacuum for DT operation and form an enclosure for containing radioactive materials. Such structural uniqueness requires quite different technologies and implementations from the existing plant, so that technical rules are to be evolved on design, fabrication and examination for assuring its structural reliability. A new code concept, namely system-based code for integrity, has been developed so as to optimize the total reliability to be attained for ITER operation, considering unique structural features and safety attractiveness. This paper describes structural features of the ITER VV from code standpoint and basic approach to code application, together with on-going research activities for supporting the code development.
BWR In-Core Monitor Housing Replacement Under Dry Condition of Reactor Pressure Vessel
Authors: Ishida Tatsuo , Yamamoto Shoji , Eguchi Fujitoshi , Fuse Motomasa , Kurosawa Kouichi , Shimizu Sadato , Masuda Minoru , Fujii Shinya , Tanaka Junji , Jacobson Bryce A.
Published: March 2009

A new method of In-Core Monitor Housing replacement has been successfully applied to Tokai Unit 2 (BWR with 1100 MWe) in April of 2001. It was designed to replace a housing under dry condition of reactor pressure vessel (RPV): this enabled the elimination of water filled-up and drained processes during the replacement procedure resulting in the reduction of implementation schedule. To realize the dry condition, the radiation shields were placed in the RPV and the hollow guide pipe (GP) was adopted to transfer the apparatuses from the top to the bottom work area.
Simulation of Bubble Dynamics in Sub-Cooled Boiling on Fuel Clad in PWRs
Authors: Wu Wen , Jones Barclay G.
Published: March 2009

The crud deposition on nuclear fuel assembly cladding generally increases the resistance to heat transfer, which may result in deterioration of thermal performance, degradation of the fuel cladding, and an axial power shift, i.e. Axial Offset Anomaly (AOA). Crud formation continues to elude prediction. An operational difficulty, of not being able to accurately determine power safety margin, then arises. In some cases, this condition has required decreasing the core power by as much as thirty percent, hence, resulting in considerable loss of revenue for the utility. The specific purpose of this study is to examine bubble dynamics, flow characteristics of the surrounding fluid, and its impact on the formation of the curd. The presence of a bubble on the clad surface affects the flow field around it , particularly in forming a stagnant flow region behind the bubble. The temperature difference between the bubble and the bulk coolant surrounding it causes vaporization at the bubble-clad interface and condensation at its apex. Pure water is thereby moved into the bubble through vaporization resulting in the concentration of solutes in the water at the bubble/wall surface region, which may cause their precipitation on and/or attachment to the clad surface, thereby initiating crud deposition. We investigate analytically and numerically, the growth of a bubble in the boundary layer and the influence of the bubble on the flow. Because of the small bubble size, a spherical model of the bubble is selected for our research. A two-step calculation is applied to this model. In the first step, bubble growth is estimated analytically with omission of the effect of the bulk fluid velocity, a reasonable approximation. In the second step, the flow field around the stationary bubble is obtained through numerical methods. Some parameters in PWR operating condition have been determined approximately e.g. size of the bubble, boundary layer thickness, flow velocity and drag forces on the bubble.
VVER 440 Reliability Data Used for PSA Studies
Authors: Cˇilli´k Ivan , Procha´ska Ja´n , Halada Peter
Published: March 2009

The paper presents the methods used for the reliability parameter calculations. The method includes full procedure of a plant specific data gathering and processing for the purposes for full and low power conditions and shutdown states for V1 WWER 440/V230 and V2 WWER 440/V213 located in Jaslovske Bohunice, Slovakia. Presented calculations include unavailability due to maintenance and test as well as some special cases. More detail description of sources plant specific data used for final elaboration as well as software used for reliability parameters calculation is given. Different software output files present data formats required not only by PSA models including calculation background for defined component groups. Based on the quality of the presented VVER 440 data VU´JE Trnava was contracted by U.S. DOE to support its PSA activities in Russia.