bluuebunny/crossref_metadata_2025 huggingface dataset:
Building an Application-Specific Grid Computing Environment Using ITBL for Nuclear Material Engineering
Authors: Tsujita Yuichi , Arima Tatsumi , Idemitsu Kazuya , Suzuki Yoshio , Kimura Hideo
Published: July 2009

Effective use of nuclear fuel is an important issue in nuclear material engineering. Pu recycle is refocused for effective use of nuclear fuel. MOX and inert matrix fuels (e.g., ZrO2-PuO2) are expected for effective burning Pu, however, Pu material is difficult to manage due to its radiotoxicity. As a result, cost of experimental facilities is very expensive. As computer simulation not only cuts cost for experimental researches but also provides deep understandings in atomic behavior, we have performed molecular dynamics (MD) simulations to research its material characteristics. In order to obtain realistic results, many atoms and many time steps are essential, however, such computation takes too much long time. So a parallelized program code is executed on a grid computing environment provided by an Information Technology Based Laboratory (ITBL) project. Its grid computing infrastructure (hereafter an ITBL system) provides users a seamless computing environment and many kinds of software tools such as a file manager, a program execution manager, and a cooperative tool for AVS/Express on a grid computing environment. Furthermore, a client application program interface (API) is provided to build a variety of grid applications on a client terminal PC for accessing the ITBL’s functionalities. As there is a strong request from users in our material simulation research to utilize their native visualization software, we have selected the client API to build an application-specific grid computing environment which cooperates with the visualization software. We have built a customized graphical user interface (GUI) computing environment on a client terminal PC by using the Java client API. It provides a seamless access to ITBL’s computational resources from a user’s terminal PC. It also assists choosing parameters for the computation in parameter survey runs. Moreover, it enables successive processing of computation on remote parallel computers and visualization on a user’s terminal PC in a single operation. The environment also provides a user friendly GUI interface for parameter controls and monitoring of submitted jobs. As a result, this computing environment removes difficulties in manual operations for parallel computations and visualizations in parameter survey runs. So, it prevents users from mistakes in the operations. This environment is expected to accelerate finding procedure for good nuclear fuel.
Assessment of Film Drop Release From Liquid Pools by an Empirical Correlation Approach
Authors: Dapper Maik , Wagner Hermann-Josef , Koch Marco K.
Published: July 2009

The present work deals with the topic of wet resuspension, particularly with regard to the basics of film drop release from bubbles and its impact on the aerosol source term as well as with the development of an empirical correlation approach adapted to the containment code system COCOSYS at low atmosphere motion. Film drops are discharged from the lamella of a bubble during the disruption process, while the bubble is resting at the fluid surface. Besides the description of the bubble disruption process, factors which have an influence on the mass and size distribution of the drops released from the bubble lamella are discussed. To analyse the distribution of the film drops of different bubble sizes, measured film drop distributions of several bubble diameters were collected from the literature. The analysis shows that with the presence of surfactants (surface-active agents) a log-normal count distribution can be used for the approximation of the drop distribution. By the evaporation of the liquid of the released film drops the solved and/or suspended materials remain as particles. In dependence of their size the drops or particles are airborne or fall back onto the liquid pool surface. The remaining airborne drops/particles are able to contribute in the late phase of a severe accident to the source term, if they are radioactive.
Prediction of Post-Dryout Heat Transfer in Vertical Annular Channels Using Artificial Neural Network Method
Authors: Zhao D. W. , Su G. H. , Qiu S. Z. , Tian W. X.
Published: July 2009

Experimental investigations on post-dryout heat transfer in 10×8.1, 10×7 and 10×6mm annular test sections have been carried out under low-pressure and low mass flow rate conditions. An Artificial Neural Network (ANN) was trained successfully based on the experimental data for predicting the average post-dryout Nusselt number. Based on the ANN, the effects of gap size, pressure, steam Reynolds number, Reg, inlet quality, xi, Prandtl number, (Prg)W, and the ratio of heat flux of inner-tube to that of outer-tube, qi/qo, on post-dryout heat transfer were analyzed, respectively. In present study, Nusselt number in annular channels with big gap size is larger than that in annular channels with small gap size. Nusselt number increases significantly in 1.5mm and 2.0mm annular channels while it is almost constant in 0.95mm annular channel with increasing pressure or qi/qo. Nusselt number increases with Reg in case of 0.95mm and 1.5mm gap sizes. However, Nusselt number in 2.0mm annular channel firstly increases and then decreases with increasing Reg. Nusselt number decreases with increasing inlet quality under all three annular channels condition. Nusselt number decreases significantly with increasing (Prg)W when (Prg)W is less than 1.5. The changes of Nusselt number in 1.5mm or 2.0mm annular channels are larger than that in 0.95mm annular channel.
Usability Evaluation Research on Operation and Monitoring Interface of Main Control Room
Authors: Yan Shengyuan , Zhang Zhijian , Peng Mingjun , Yang Ming
Published: July 2009

Deciding the evaluation index weight is a key technique in estimating human factor design of operation and monitoring interfaces (OMI). It is very hard to avoid the influences caused by the individualities of valuators and the randomicity factors by using the traditional OMI evaluation methods. An RBF networks based subjective evaluation method is proposed in this paper, which has the properties of self-organizing, self-learning, self-adapting. In addition, because the RBF networks can be trained to study and learn the regularity of index weights of subjective evaluation concealed in the training data, and therefore the influence of randomicity factors can be overcome by means of RBF networks automatically adjusting index weights of subjective evaluation. The subjective evaluation indexes and models of OMI are established. The number of samples, the value of spread, the accuracy of RBF subjective evaluation network model, and the relationship between them are researched Error analyses of OMI subjective evaluation model based on RBF network are carried out by using different training samples, respectively. The analysis results show that the OMI subjective evaluation model by using 80 training samples is of satisfied accuracy.
Reliability and Safety Assessment of a Conceptual Thermochemical Plant for Nuclear-Based Hydrogen Generation
Authors: Zhang Yuepeng , Lu Lixuan , Naterer Greg F.
Published: July 2009

Hydrogen is a clean fuel that can help to reduce greenhouse gas emissions, as its oxidation does not emit carbon dioxide (a primary greenhouse gas). Generation of hydrogen has attracted much recent worldwide attention. A promising method to generate hydrogen is to use heat from nuclear power plants. The advantages of using nuclear heat are capabilities of large-scale generation of hydrogen and zero greenhouse gas emissions. Nuclear energy is expected to have an important role for hydrogen generation in the future. In this paper, reliability and probabilistic safety assessments of a conceptual nuclear-hydrogen plant will be analyzed. There are two main methods to generate hydrogen from nuclear energy. They include: 1) thermochemical processes and 2) electrochemical processes. The conceptual plant of this paper is based on a Cu-Cl thermocycle developed by Atomic Energy of Canada Limited (AECL) and the Argonne National Laboratory (ANL). Using a flowsheet of the hydrogen plant created by an Aspen Plus simulation by ANL, four fault-trees are constructed for potential risk scenarios. Based on the results from the fault tree analyses (FTA), the risk levels of the hydrogen generation plant under different accident scenarios can be calculated. Based on the results, potential problems encountered in Cu-Cl cycle are identified and possible solutions will be recommended for future improvements.
Experiments and Analysis on Film Boiling Heat Transfer Around a Finite-Length Vertical Cylinder With a Convex Surface Swelling Downward
Authors: Momoki Satoru , Toyoda Kaoru , Yamada Takashi , Shigechi Toru , Yamaguchi Tomohiko
Published: July 2009

A method of predicting the overall heat transfer coefficient and the temperature at the lower limit of film boiling for a finite-length cylinder with flat top and bottom surfaces has been researched and proposed in a previous paper. This paper presents and compares an analysis in the case of a cylinder with a hemispherical bottom. The film boiling heat transfer around a vertical silver cylinder with a convex hemispherical bottom surface is investigated both experimentally and analytically in the present study. The obtained results are also compared and discussed with the authors’ previous results for a finite-length cylinder with flat top and bottom surfaces. Quenching experiments were performed using silver cylinders in saturated water. The diameter and length of the test cylinders are 32mm and 48mm, respectively. The test cylinder was heated up to about 600°C in an electric furnace and then cooled down in saturated quiescent water at atmospheric pressure. The resultant cooling and boiling curves and photographs of the film boiling phenomena are presented and discussed. The average heat transfer performance of the hemispherically bottomed cylinder is about 20% higher than that of the flat bottomed cylinder. The degree of wall superheating at the lower limit of film boiling is about 133K. The saturated film boiling heat transfer around the vertical finite-length cylinder with a convex hemispherical bottom was analyzed by taking into account the convective heat transfers from the bottom, side and top surfaces of the cylinder. The resulting analytical data correlated closely with the experimental data in the present study.
Rationalization of Radiation Shielding Design for Spent Nuclear Fuel Storage Building
Authors: Nemoto Yuji , Tsukiyama Toshihisa , Nemezawa Shigeki , Nakano Hideo
Published: July 2009

A spent nuclear fuel storage building is generally a structure provided with heat removal, radiation shielding functions, and a seismic design. The facility has air supply ducts and a chimney exhaust duct for removing heat through natural cooling, with the structural shapes determined by the demands of the above functions. In radiation protection design, radiation levels must be kept below acceptable levels for the general public. The radiation dose can be lowered by increasing the concrete thickness, as typically applied in radiation shielding design, or by increasing the distance. The setting of additional shielding plates also helps reduce the scattered radiation escaping from the air supply and exhaust ducts. However, such protective structures against the scattered radiation challenges in effective heat removal and seismic design. Therefore, determining a building structure that can satisfy all safety demands requires a great deal of time. This study aims to effectively achieve radiation shielding. The height and width of the exhaust duct were considered, and the correlation of these parameters was studied. In the calculation, two-dimensional Sn code (DORT) was used to examine the validity of the results. Monte Carlo N-Particle Transport Code (MCNP) calculations were also made for comparison with the DORT results. The cross-section libraries of JENDL3.3 and DLC23F were used in this calculation, with the difference being clarified.
Response Analysis From Transient Signal for Variation of R-C Passive Elements Under High Temperature Accident Condition in NPPs
Authors: Koo Kil Mo , Ahan Kwang Il , Song Yong Mann , Yang Joon Eon , Park Eung Seok
Published: July 2009

The electrical signal could be checked whether it lies within its expected electrical range when it is doubtful condition. In the case of an abnormal transient signal from an instrument under a severe accident condition, it needs to define a signal validation to operate at control room in NPPs. Diagnostics and analysis for some abnormal transient signals have been also performed through the important equivalent circuits modeling for the passive elements under the conditions. Even under such a condition, the electrical signal should be within its expected range so that some mitigation actions can be taken through it in control room. Whereas, in case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response signals have been evaluated with an equivalent circuit model to real instruments, which is regarded as the best way. Main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for an equivalent circuit modeling. The program links a signal analysis tool code to an abnormal signal simulation engine not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis) module developed to obtain a varying range of the elements in high temperature conditions. As a result, a special function for abnormal signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output signals through a response characteristic of a 4∼20mA circuit model and a range of the elements changing with temperature under an accident condition.
Physics Principles to Achieve Comparable Fission Power From Fertile and Fissile Rods of the Conceptual ATBR/FTBR Reactors
Authors: Pal Usha , Jagannathan V.
Published: July 2009

Loading of seedless fertile rods has been used as the central principle to maximize fertile to fissile conversion in the two thorium breeder reactor concepts, viz. ATBR and FTBR [1, 2]. At fresh state the seedless thoria rods will produce practically no fission power, or nearly thousand times less fission rate compared to the seed fuel rods. Hence it is conceived that the fuel assembly would be constituted by assembling the fresh seed rods with one fuel cycle irradiated fertile thoria rods. Even in this state there is a wide disparity between the fissile content of these rods. By judicious choice of the rod dimensions and their relative locations, a degree of balance in the fission rate is achieved in the fresh state of seeded rods. Remarkably as the burnup proceeds the initially seedless fertile rods have a continuous growth of fissile content up to an asymptotic value for a given spectrum and the fissile content in seeded rods monotonically decreases. If the discharge burnup is sufficiently large by design, it is seen that the power share of the initially seedless fertile rods can even exceed that of the seed fuel rods. The physics principles of achieving this characteristic are presented in this paper.
Non-Linear BWR Stability Analysis
Authors: Wehle F. , Schmidt A. , Opel S. , Velten R.
Published: July 2009

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants already during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge oN BWR instabilities. The appropriate representation of the physical processes in the non-linear regime requires typically time domain stability analysis. The objective of this paper is to present a physical model, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. The application of this model gives a deeper insight into the physical reasons for the prevention of the uncontrolled divergence of BWR oscillations. The mechanisms that have a stabilizing effect are demonstrated.
Assessment of the Airborne Release Coefficient During Cutting Operations in the Frame Dismantling Studies
Authors: Pilot Guy
Published: July 2009

In the frame of the decommissioning of nuclear facilities, many cutting techniques have been studied in different configurations, in order to qualify the performance of the tools implemented and also the quantity and characteristics of the wastes and secondary emissions produced. The present paper, calling the operational conditions of each tool, presents a synthesis of the results obtained from the suspension or airborne release point of view. The knowledge of the suspension fraction of radioactive aerosol contamination is needed for the safety analysis of nuclear installation and for the radioprotection of workers as well. This synthesis shows that the suspension fraction can vary over several orders of magnitude according to the configurations studied and provides relevant information for the dismantlers.
Design and Development of a Reliability Analysis Tool Based on Multilevel Flow Models
Authors: Yang Ming , Zhang Zhijian , Yan Shengyuan , Peng Minjun
Published: July 2009

This paper proposes a universal graphical tool for the modeling and reliability analysis of complex industrial process system based on Multilevel Flow Models (MFM). An Extensible Markup Language (XML) is used for structuring the MFM model. An editor is developed and an executor can implement reliability analysis in terms of the established MFM models. The proposed reliability analysis tool is meaningful for further research of the MFM based system reliability analysis and will be useful for more practical applications such as online risk monitoring by integrating the existed algorithms of alarm analysis and fault diagnosis based on MFM.
Analysis of Pressurized Water Reactor With Pinch, Exergy and Thermoeconomic Methods
Authors: Khoshgoftar Manesh Mohammad Hasan , Amidpour Majid
Published: July 2009

Exergetic, pinch and thermoeconomic analyses were performed for a 1000 MW PWR power plant similar to BUSHEHR power plant. In these analyses, mass and energy conservation laws were applied to each component of the system. Quantitative balances of the exergy and exergetic cost for each component, and for the whole system was carefully considered. The exergoeconomic model, which represented the productive structure of the system considered, was used to visualize the cost formation process and the productive interaction between components. Simulation of power plant is performed in STEAM PRO 13.0. The computer program developed in this study can determine the production costs of each component of PWR power plant such as electricity production in steam turbines. In addition, this code can provide combined exergy and pinch representation for each component of system. This graphical representation helps us to show energy integration in the nuclear power plant. The code can be also be used to study plant characteristics, namely, thermodynamic performance and sensitivity to changes in process and/or component design variables.
Motion of Small Bubbles Near a Grid Spacer in a Two by Three Rod Bundle
Authors: Zhang Zheng , Suzuki Kana , Hosokawa Shigeo , Tomiyama Akio
Published: July 2009

Several experiments have reported that departure from nucleate boiling (DNB) in a pressurized water reactor (PWR) fuel assembly often takes place just upstream of a grid spacer. Since DNB inception is to be tightly related with an extremely high void fraction in the vicinity of a heated wall, it is speculated that one of the causes of DNB inception is lateral migration of bubbles toward a fuel rod due to a flow induced by the grid spacer. Therefore, experiments are conducted to investigate the motion of small bubbles and liquid velocity profiles just upstream of a grid spacer in a two by three rod bundle. Tap water and air are used for the liquid and gas phases, respectively. The bulk liquid velocity ranges from 1.0 to 4.0 m/s and experiments are conducted at atmospheric pressure and room temperature. Trajectories of single small bubbles are recorded using a high-speed video camera. Mean liquid velocities and turbulence distributions upstream of the spacer are calculated based on 10,000 instantaneous velocity distributions measured by Particle Image Velocimetry (PIV). Discussion on the effects of the gird spacer on bubble motion is made based on the difference between the measured bubble trajectories and liquid streamlines. Numerical simulations using a three-dimensional two-way bubble tracking method are also carried out. The calculated trajectories of single bubbles and liquid velocity distributions are compared with the measured data. The main conclusions obtained are as follows: (1) just upstream of the grid spacer, bubbles are apt to migrate toward the rods and accumulate in the vicinity of the rod surface, (2) a centrifugal force caused by a curved streamline just upstream of the grid spacer is the main cause of lateral migration of bubbles toward the rods, and (3) the two-way bubble tracking method gives good predictions for trajectories of small bubbles upstream of the grid spacer and for liquid velocity profiles.
Study on Enhanced Performance Sodium-Cooled Metal Fuel Core Concepts by Adopting Advanced Fuel and Flexible Design Criteria
Authors: Kobayashi Noboru , Ogawa Takashi , Ohki Shigeo , Mizuno Tomoyasu , Ogata Takanari
Published: July 2009

The metal fuel core is superior to the mixed oxide fuel core because of its higher breeding ratio and compact core size resulting from neutron economics, hard neutron spectrum, and high content of heavy metal nuclides. Meanwhile, the metal fuel core exhibits the characteristic of a lower allowable maximum cladding temperature. Utilizing the advantage of the metal fuel core, conceptual sodium-cooled fast breeder reactor designs have been pursued for the attractive core properties of high breeding ratio, small inventory, compact size, low sodium void reactivity, and high transmutation ratio of the minor actinides. Among attractive cores, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void reactivity of less than 8 $, a core height of less than 150 cm, a maximum cladding temperature of 650 °C, and a fuel pin bundle pressure drop of 0.4MPa. The breeding ratio of the resultant core was 1.34 without blanket fuels.
R&D of the Next Generation Safety Analysis Methods for Fast Reactors With New Computational Science and Technology: Part 2 — Development and Verification of Thermo-Hydrodynamics Module of the COMPASS Code
Authors: Yamamoto Yuichi , Hirano Etsujo , Oue Masaya , Shimizu Sensuke , Shirakawa Noriyuki , Koshizuka Seiichi , Morita Koji
Published: July 2009

A computer code, named COMPASS, is being developed employing the Moving Particle Semi-implicit (MPS) method for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In FY2006, development of the basic functions of the COMPASS was completed and fundamental verification calculations are being carried out at the moment. In this paper, we describe the conceptual design of COMPASS and the outline of the formulation of MPS method, and show the results of the basic verification calculations for thermo -hydrodynamics module.
CANDU 600 Severe Accident Assessment
Authors: Negut Gheorghe , Prisecaru Ilie , Catana Alexandru , Dupleac Daniel
Published: July 2009

Romania is now a UE member since January first 2007. New challenges are for our country that includes, also, their nuclear power reactors. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU are operated PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model severe accidents for these types of reactors. Starting from previous studies a CANDU degraded core thermalhydraulic model was developed. The initiating event is a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.
Research and Analysis on Damage of Fretting Wear and Fretting Fatigue for Components in NPPs
Authors: Ning Dong , Zhang Min , Xie Yongchen , He Yinbiao , Yao Weida
Published: July 2009

In this paper, from the point of view of study and analysis on fretting wear and fretting fatigue mechanism of materials, the main differences between fretting fatigue and ordinary fatigue are summarized; further more, the damage reasons on a component in a NPP are analyzed and semi-quantitative relationship between fretting fatigue and ordinary fatigue for the component material is obtained to provide with referenced basis for structural improvement and failure analysis on original structure.
Analysis of the Chimney Effect During the Reflooding Phase of a Large Break LOCA Transient With the 3D Module of the CATHARE2 Code
Authors: Tamburini Isabelle , Guillard Valia , Seiler Nathalie
Published: July 2009

This study deals with the ability of the three-dimensional module of the CATHARE2 code to simulate the thermalhydraulic behavior of a 900MWe Pressurized Water Reactor (PWR) in Large Break Loss of Coolant Accident (LOCA) conditions. The CATHARE2 code is a “Best-Estimate” system code, developed by the CEA, in collaboration with EDF, AREVA NP and IRSN, used in France in the frame of realistic methodology to evaluate safety margins. Particularly, the realistic simulation of the so-called “chimney effect”, which occurs during the reflooding phase of a Large Break LOCA is of primary importance. Observed during experiments, this effect is indeed characteristic of the hydraulic behavior of a nuclear core presenting a non-uniform radial power profile. Several separate effect tests such as PERICLES 2D reflood and CCTF/SCTF experiments have demonstrated the existence of cross-flows between the hot assembly and the mean assemblies of the core during this reflooding phase. Liquid goes from the mean assemblies toward the hot one beneath the quench front leading to an increase in the heat transfer coefficient in the hot assembly compared to the one in the mean assemblies, and hence to a better cooling of the hot rod. After a literary survey on the “chimney effect”, quantitative information has been found in several publications concerning SCTF and CCTF tests. More precisely, a correlation has been established from the results of these tests providing the increase rate of the heat transfer coefficient in the hot assembly compared to the one in the mean assemblies depending on the power features of the core. The assumptions related to the establishment of this correlation are first validated in case of the PERICLES 2D Reflood test configuration. Then the simulation of the “chimney effect” by the three-dimensional module of the CATHARE2 code is analyzed by comparing simulation and experimental results in the PERICLES 2D Reflood test configuration. Finally, the same kind of study is performed with the simulation of a 900MWe PWR core in reflooding conditions typical of a Large Break LOCA transient. In both cases, the difference between the heat transfer coefficients of the hot assembly and the mean ones obtained during the CATHARE2 simulations is compared to the correlation derived from the SCTF and CCTF experimental results. While the simulation of a Large Break LOCA in a 900MWe PWR has quite well reproduced SCTF/CCTF experimental evidences, the study performed with PERICLES configuration has not given such satisfying results, probably due to the lack of representativeness of the device (only three aligned assemblies).
Influence of the CATHARE2 Three-Dimensional Meshing of a 900MWe PWR Core in LOCA Conditions
Authors: Guillard Valia , Seiler Nathalie , Tamburini Isabelle , Ducros Marie , Massin Ste´phanie , Glantz Tony
Published: July 2009

The CATHARE2 code is a “Best-Estimate” system code, developed by the CEA, EDF, AREVA-ANP and IRSN, mainly used in France in the frame of realistic methodology to evaluate safety margins for Pressurized Water Reactor (PWR). Since a three-dimensional thermal hydraulic module, based on a two-fluid 6-equations model, is now available in the most recent versions of the code, a new challenge consists in analyzing whether the 3D modeling of the vessel allows to describe more accurately complex 3D phenomena occurring during Loss of Coolant Accident (LOCA) transients. This document specially studies the sensitivity of the 3D meshing of a 900MWe PWR vessel in Large and Intermediate Breaks LOCA conditions. The development of such a meshing, suitable for the CATHARE2 code, is a long and meticulous task that implies important knowledge of the PWR vessel features and the evaluation of numerous CATHARE2 parameters such as repartition of the guide tubes, volumetric and surfacic porosities or hydraulic diameters. This study also requires a good understanding of the main specific phenomena occurring during the various types of LOCA transients in order to check if the chosen CATHARE2 meshing is well adapted. These studies are successively focused on the consideration of homogenous parameters in the vessel lower head, on axial meshing refinement in the lower head as well as in the active part of the core and finally on radial and azimuthal meshing refinement in the vessel. In order to assess its results and since any experimental evolutions of hydraulics core parameters are available for such multi-dimensional transients, IRSN has considered the physical experimental results of separate effect tests, characteristic of LOCA transients and available in the literature. It has also taken into account the results of previous reference calculations performed with the CATHARE2 code. Considering different three-dimensional meshing of the vessel, the evolution of the hydraulics is observed in the whole reactor and more accurately in the core, which is submitted to a non-uniform radial power profile. In case of Large Break LOCA transient analysis, more attention is paid to the simulation of cross-flows between hot and cold channels during the reflooding phase. This study allows us to verify the strong impact of the axial refinement of the core meshing on the simulation results, leading to greater cross-flows under the quench front. The study of the Intermediate Break LOCA transient, with a delayed pumps stop, has also shown a strong impact of the vessel schematization on the hydraulic evolution. Contrary to the reference case water falling back from hot legs to the upper plenum is observed at the centre of the core and flow is less disturbed in the lower plenum when the CATHARE2 parameters are homogenized. Moreover these sensitivity tests to the 3D meshing show that cladding temperatures are dependent on axial meshing refinement.
Investigation on Header Manifold Model Effect at CANDU6 Large Loss of Cooling Accident
Authors: Dupleac Daniel , Prisecaru Ilie , Mladin Mirea , Negut Gheorghe , Ghitescu Petre
Published: July 2009

In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pressure tubes tube through which the heavy-water coolant flows. 95 pressure tubes are connected by individual feeders to a common header. For CANDU 6 safety analyses, even when multiple channels model is employed, only one node is used for header. In this approach, all the channels are subjected to the same boundary condition. However, site inlet and outlet header pressure measurements and ultrasonic feeder flow data, confirm the existence of axial pressure gradients along the inlet and outlet headers. These axial pressure gradients would give rise to individual header-to-header pressure drops for each channel and also to flow distribution throughout both the inlet and outlet headers. In this paper, the header manifold model effect on the large break loss of coolant accident analyses of CANDU reactors has been performed by RELAP5/ mod 3.4 code. The 35% reactor inlet header break was selected for this study, as this break size produce the highest fuel clad temperature among all postulated breaks size. The results obtained considering the header manifold model, show that location of fuel channel upon break location has a strong impact on peak clad temperature calculation.
Study of Heat Mass and Mass Transfers in a Spray for Containment Application: Analysis of the Influence of the Spray Mass Flow Rate
Authors: Lemaitre Pascal , Porcheron Emmanuel
Published: July 2009

During the course of a hypothetical accident in a nuclear power plant, spray might be activated in order to reduce static pressure in the containment. To have a better understanding of the heat and mass transfers between a spray and the surrounding confined gas, the IRSN has developed the TOSQAN experiment. This article is devoted to analyse the influence of the injected spray mass flow rate on the thermal-hydraulic of spray tests. In order to perform this analysis, two other tests are performed with exactly the same boundary and initial conditions, except the spray mass flow rate that ranges from 10 to 50 g.s−1. First, the scenario of these three tests and the associated results used for this analysis are presented. Then, we focus our analysis on the inter-comparison of the thermal-hydraulic behaviour induced by spray mass flow rates variations. This inter-comparison is divided into two parts: a global and a local one.
Problems of Development of a Prediction Technique of a Residual Lifetime of Heat Exchanging Tubes of WWER NPPs
Authors: Bakirov Murat , Chubarov Sergei , Morozov Evgeny
Published: July 2009

Efficiency of steam generators (SG) is one of the major components of safety operation of NPP units. The history of design evolution and operation of SG as in Russia, and abroad, shows that main for SG resource is problem of heat exchanging tubes (HET) integrity. It is replaced more than 150 SG on NPP with PWR to 2005 year. It has happened because of heat exchanging tubes have been damaged. This reason has led to replacement of 6 SG on the NPP with WWER [1].
Uranium Recovery From LWR Spent Fuel for the Future FR Deployment
Authors: Fukasawa Tetsuo , Yamashita Junichi , Hoshino Kuniyoshi , Fujimura Koji , Sasahira Akira
Published: July 2009

Nuclear energy systems are necessary to assure sufficient energy resources without harming the environment. Fast reactor (FR) systems are especially important taking into account the limited uranium resources and the nuclear sustainability. As the FR system is still under development, FR deployment start-time and rate are unclear. On the other hand, it is desirable to reduce light water reactor (LWR) spent fuel due to the difficulties of storage and disposal (retrievable) site determination. Reprocessing is one of the effective methods to reduce LWR spent fuel but the recovery and long-term storage of plutonium, even with uranium, is undesirable for the aspect of proliferation resistance. The authors propose the new system named Flexible Fuel Cycle Initiative (FFCI), which recovers only uranium (∼90%) from LWR spent fuel and stores the residual material (∼5% U, ∼1% Pu, ∼4% other nuclides) for the future FR deployment. Residual material named recycle material (RM) is suitable for FR fresh fuel preparation due to its high Pu concentration and similar Pu/U ratio to FR core fuel, and for proliferation resistance due to its high concentrations of fission products (FP) and minor actinides (MA). The volume of RM is about 1/10 of that of LWR spent fuel. However RM needs sufficient heat removal, radiation shielding and criticality safety. After the FR development is finished and several years before the commercial FR deployment start-time, Pu and U will be recovered from the RM that might be stored liquid or solid state. Many well known methods can be applied for U recovery such as solvent extraction, crystallization, precipitation, electro refining, and fluoride volatilization. As recovered U has slightly higher U-235 concentration than natural U, its re-enrichment and recycling in LWRs seems to be effective for ultimate utilization of nuclear resources. In this case fluoride volatility U recovery method is most preferable because the product is UF6 that is the supply material for enrichment. Quantitative evaluations have been carried out for several fuel cycle systems including FFCI with parameters such as spent fuel amounts, facility capacity and Pu balance, which revealed the feasibility and flexibility of FFCI for LWR spent fuel reduction, high facility capacity factors and sufficient (no excess) Pu supply to FR.
Technical Development for IASCC Irradiation Experiments at the JMTR
Authors: Shibata Akira , Nakano Junichi , Ohmi Masao , Kawamata Kazuo , Saito Takashi , Hayashi Kouji , Saito Junichi , Nakagawa Tetsuya , Tsukada Takashi
Published: July 2009

Irradiation assisted stress corrosion cracking (IASCC) is considered to be one of the key issues from a viewpoint of the life management of core components in the aged Light Water Reactors (LWRs). To simulate IASCC behavior by the in-pile IASCC experiment or post-irradiation experiment (PIE), it is necessary to irradiate specimens up to a neutron fluence that is higher than the so-called IASCC threshold fluence in a test reactor. There are, however, some technical hurdles to overcome for the experiments. For the in-pile IASCC test, techniques assembling pre-irradiated specimens into an in-pile test capsule in a hot cell by remote handling are necessary, and the Japan Atomic Energy Agency (JAEA) developed the techniques for the in-pile test to be carried out in the Japan Material Testing Reactor (JMTR). To examine crack growth and crack initiation behaviors under neutron irradiation, pre-irradiated specimens were relocated from pre-irradiation capsules to an in-pile capsule. Hence, a remote welding machine has been newly developed and welding work for inner and outer tubes of capsule are carried out with rotating of the capsule. The other hurdle is the material integrity of the capsule of the capsule housing for a long term irradiation. Since the changes in microstructure, micro chemistry and mechanical properties of materials increase with neutron fluence, the integrity for capsules of long irradiation period was evaluated by tensile tests in the air and slow strain rate test (SSRT) in oxygenated water. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 1026 n/m2 (E> 1 MeV) previously. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.
Research on Flow Instability in Vertical Narrow Annuli
Authors: Wu G. P.
Published: July 2009

A narrow annular test section of 1.5mm gap and 1800mm length was designed and manufactured, with good tightness and insulation. Experiments were carried out to investigate characteristics of flow instability of forced-convection in vertical narrow annuli. Using distilled water as work fluid, the experiments were conducted at pressures of 1.0∼3.0MPa, mass flow rates of 3.0∼25kg/h, heating power of 3.0∼ 6.5kW and inlet fluid temperature of 20 °C, 40 °C or 60°C. It was found that flow instability occured with fixed inlet condition and heating power when mass flow rate was below a special value. Effects of inlet subcooling, system pressure and mass flow rate on the system behavior were studied. The instability region is given, which has satisfactory agreement with results of the calculation based on the assumption of homogenous two-phase flow and thermodynamic equilibrium of the phases.
Using Artificial Neural Networks in the Reactor Safety Research
Authors: Kratzsch Alexander , Ka¨stner Wolfgang , Hampel Rainer
Published: July 2009

The paper deals with the creation of a differential pressure model with artificial neural networks (ANN). Particular, model development and verification tests are considered. One of the main features in reactor safety research is the safe heat dissipation from the reactor core and the reactor containment of light-water reactors. In the case of loss of coolant accident (LOCA) the possibility of the entry of isolation material into the reactor containment and the building sump of the reactor containment and into the associated systems to the residual heat exhaust is a serious problem. This can lead to a handicap of the system functions. To ensure the residual heat exhaust it is necessary the emergency cooling systems to put in operation which transport the water from the sump to the condensation chamber and directly to the reactor pressure vessel. A high allocation of the sieves with fractionated isolation material, in the sump can lead to a blockage of the strainers, inadmissibly increase of differential pressure, build-up at the sieves and to malfunctioning pumps. Hence, the scaling and retention of fractionated isolation material in the building sump of the reactor containment must be estimated. This allows the potential plant status in case of incidents to be assessed. The differential pressure is the essential parameter for the assessment of allocation of the strainers. For modelling we use artificial neural networks. To build up the ANN, the available experimental data are used to train the ANN.
RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project
Authors: Nakamura Hideo , Watanabe Tadashi , Takeda Takeshi , Asaka Hideaki , Kondo Masaya , Maruyama Yu , Ohtsu Iwao , Suzuki Mitsuhiro
Published: July 2009

The Japan Atomic Energy Agency (JAEA) started OECD/NEA ROSA Project in 2005 to resolve issues in the thermal-hydraulic analyses relevant to LWR safety through the experiments of ROSA/LSTF in JAEA. More than 17 organizations from 14 NEA member countries have joined the Project. The ROSA Project intends to focus on the validation of simulation models and methods for complex phenomena that may occur during DBEs and beyond-DBE transients. Twelve experiments are to be conducted in the six types. By utilizing the obtained data, the predictability of codes is validated. Nine experiments have been performed so far in the ROSA Project to date. The results of two out of these experiments; PV top and bottom small-break (SB) LOCA simulations are studied here, through comparisons with the results from pre-test and post-test analyses by using the RELAP5/MOD3.2 code as a typical and well-utilized/improved best estimate (BE) code. The experimental conditions were defined based on the pre-test (blind) analysis. The comparison with the experiment results may clearly indicate a state of the art of the code to deal with relevant reactor accidents. The code predictive capability was verified further through the post-test analysis. The obtained issues in the utilization of the RELAP5 code are summarized as well as the outline of the ROSA Project.
LIVE Experiments on Melt Behavior in the RPV Lower Head
Authors: Miassoedov Alexei , Cron Thomas , Foit Jerzy , Gaus-Liu Xiaoyang , Schmidt-Stiefel Silke , Wenz Thomas
Published: July 2009

Behavior of the corium pool in the lower head is still a critical issue in understanding of PWR core meltdown accidents. One of the key parameter for assessing the vessel mechanical strength is the resulting heat flux at the pool-vessel interface. A number of studies [1]–[3] have already been performed to pursue the understanding of a severe accident with core melting, its course, major critical phases and timing and the influence of these processes on the accident progression. Uncertainties in modeling these phenomena and in the application to reactor scale will undoubtedly persist. These include e.g. formation and growth of the in-core melt pool, relocation of molten material after the failure of the surrounding crust, characteristics of corium arrival in residual water in the lower head, corium stratifications in the lower head after the debris re-melting [4]. These phenomena have a strong impact on a potential termination of a severe accident. The main objective of the LIVE program [5] at FZK is to study the core melt phenomena both experimentally in large-scale 3D geometry and in supporting separate-effects tests, and analytically using CFD codes in order to provide a reasonable estimate of the remaining uncertainty band under the aspect of safety assessment. Within the LIVE experimental program several tests have been performed with water and with non-eutectic melts (mixture of KNO3 and NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behavior. The information obtained from the LIVE experiments includes heat flux distribution through the reactor pressure vessel wall in transient and steady state conditions, crust growth velocity and dependence of the crust formation on the heat flux distribution through the vessel wall. Supporting post-test analysis contributes to characterization of solidification processes of binary non-eutectic melts. Complimentary to other international programs with real corium melts, the results of the LIVE activities provide data for a better understanding of in-core corium pool behavior. The experimental results are being used for development of mechanistic models to describe the in-core molten pool behavior and their implementation in the severe accident codes like ASTEC. The paper summarizes the objectives of the LIVE program and presents the main results obtained in the LIVE experiments up to now.