bluuebunny/crossref_metadata_2025 huggingface dataset:
A Short Method to Compute Nusselt Numbers in Rectangular and Annular Channels With any Ratio of Constant Heat Rate
Authors: Malon Alexandre , Muller Thierry
Published: April 2011

An analytic investigation of the thermal exchanges in channels is conducted with the prospect of building a simple method to determine the Nusselt number in steady, laminar or turbulent and monodimensional flow through rectangular and annular spaces with any ratio of constant and uniform heat rate. The study of the laminar case leads to explicit laws for the Nusselt number while the turbulent case is solved using a Reichardt turbulent viscosity model resulting in easy to solve one-dimensional ordinary differential equation system. This differential equation system is solved using a Matlab based boundary value problems solver (bvp4c). A wide range of Reynolds, Prandtl and radius ratio is explored with the prospect of building correlation laws allowing the computing of Nusselt numbers for any radius ratio. Those correlations are in good agreement with the results obtained by W.M. Kays and E.Y. Leung in 1963 [1]. They conduced a similar analysis but with an experimental basis, they explored a greater range of Prandtl but only a few discreet radius ratio. The correlations are also compared with a CFD analysis made on a case extracted from the Re´acteur Jules Horowitz.
Study on Nuclear Power Plant: A Recognition Category Emergency Action Level
Authors: Liu Tao
Published: April 2011

Emergency Action Level (EAL) is an effective basis and criteria for nuclear power plant emergency classification which is a pre-determined, site specific, observable threshold for a plant initiating condition that places the plant in a given emergency classification level. Systematic approaches have built a set of generic EAL guidelines, together with the basis for each, such that they could be used and adapted by each utility on a consistent basis. EALs information is presented by Recognition Categories (RC), in which “A” Recognition Category refers to abnormal Rad levels/Radiological effluent. The methodology of EALs development is reviewed first and the technical bases for “A” Recognition Category according to classic EAL methodology, i.e. NEI 99-01 series and IAEA-TECDOC-955 are introduced. The statue of Chinese nuclear power plants’ “A” Recognition Category EALs development is summarized after that, which intends to raise some challenging questions for classic “A” Recognition Category EALs development. One opinion is that some “A” Recognition Category EALs in NEI 99-01 series are not conservative. To explain the reason, the discussion mainly focuses on the interpretation and comparison of NEI 99-01 and IAEA-TECDOC-955 technical bases. Finally, it is recommended to select one suitable guideline as reference when developing plant specific EALs.
Life Beyond 60 Years: The Importance of Sustaining the Current Fleet of Light Water Reactors in the U.S. and the R&D Needs
Authors: Szilard Ronaldo , Zhang Hongbin
Published: April 2011

The current fleet of 104 nuclear power plants in the U.S. began their operation with 40 years operating licenses. About half of these plants have their licenses renewed to 60 years and most of the remaining plants are anticipated to pursue license extension to 60 years. With the superior performance of the current fleet and formidable costs of building new nuclear power plants, there has been significant interest to extend the lifetime of the current fleet even further from 60 years to 80 years. This paper addresses some of the key long term technical challenges and identifies R&D needs related to the long term safe and economic operation of the current fleet.
Reactivity Accident Due to Control Rod Inadvertent Withdrawal of the 250MW High Temperature Gas-Cooled Reactor
Authors: Zheng Yanhua , Shi Lei
Published: April 2011

Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM) with single module power of 250MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper, e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not, and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature and Xenon concentration are studied and compared in detail between these different cases. The calculating results show that, the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project.
On the Uncertainty and Sensitivity Analysis of Experiments With Supercritical Water With TRACE and SUSA
Authors: Jaeger Wadim , Sa´nchez-Espinoza Victor H. , Macia´n-Juan Rafael
Published: April 2011

This study was performed at the Institute for Neutron Physics and Reactor Technology at Karlsruhe Institute of Technology and is addressed to establish the combined usage of the best-estimate code TRACE and the uncertainty and sensitivity analysis tool SUSA, for safety related investigations of current and future nuclear energy systems. In the frame of this paper the applicability to supercritical water related investigations is covered. Several Nusselt correlation for the heat transfer to supercritical water are available and have been evaluated in previous investigations but not one of them gave satisfying results. Hence, the consideration of uncertainty and sensitivity measures applied to the topic of heat transfer seems to be an appropriate way. In a first step a post-test analysis of an experiment was conducted. Results showed that with the help of uncertainty and sensitivity methods parameters which affect the results most could be identified. The most important parameter was of course the Nusselt correlation. In addition to the identification of important parameters, the experimental results were enveloped by the calculated results. That means, in the sense of safety related evaluation of designs for reactors operated with supercritical water, that key parameters (cladding temperature) can be calculated with a certain confidence.
The Design of a Neutron Dosimeter
Authors: Xia Wenming , Jia Mingchun , Guo Zhirong
Published: April 2011

At present, most of the developed neutron dosimeters that have a moderator with a single counter, applied in neutron radiation fields within large range energies from thermal to MeV neutrons, are not a satisfaction to energy response. The purpose of the article is designing a suitable neutron dosimeter for the radiation protection purpose. In order to overcome the disadvantage of the energy response of the neutron dosimeters combined a single sphere with a single counter, three spheres and three 3He counters were combined for the detector design. The response function of moderators with different thicknesses combined with SP9 3He counters were calculated with MCNP program MCNP4C [1]. The selection of three different thicknesses of the moderating polyethylene sphere was done with a Matlab program [2]. A suitable combination of three different thicknesses was confirmed for the detector design. The electronic system of the neutron dosimeter was introduced. The fluence to ambient dose-equivalent conversion coefficient were calculated, analyzed and compared with the values recommended in the ICRP 74 Publication [3]. The calculated result explain that it is very significance to this design of neutron dosimeter, it may be applied to the monitor of the ambient dose in the neutron radiation fields, improving at present the status of the energy response of neutron dosimeters.
Study of the Influence of Axial Power Distribution on Dryout
Authors: Anglart Henryk
Published: April 2011

Axial power distribution is one of the parameters that influence the occurrence of the dryout in nuclear fuel assemblies. Experimental data indicate that this influence is quite substantial, ranging from few to above ten percent of the total power. Thus accurate prediction of the dryout power for various power distributions has important implications on the economy and safety of nuclear power plants. The difficulty with capturing the influence of that parameter stems from the fact that during reactor operation practically unlimited number of power shapes can occur. This fact makes it very difficult to investigate the effect experimentally, and an analytical approach is needed. Various methods have been proposed in the past to capture the effect of non-uniform power distribution on dryout. These approaches can be divided into several categories, where the two main ones are as follows: (a) methods based on introduction of a shape factor, which is calculated from the known shape of the power distribution; (b) methods using certain integral parameters, such as the boiling length and the annular flow length, which are expressed as functions of axial power distribution. In the present approach a simplified annular flow model is used, in which the dryout occurrence is based on the prediction of the disappearance of the liquid film. The dependence of the dryout power on the axial power shape is obtained in a general analytical form. Based on this analytical solution, a new set of terms that govern the dryout power in channels with various axial power distributions is proposed.
Initial Events Assessment of Auxiliary Building for Internal Flooding PSA
Authors: Chen Renzong
Published: April 2011

This documentation summarized initial events assessment of auxiliary building as a result of internal flooding based on Chinese 1000MWe PWR. Detailed screening procedure of plant areas and quantitative frequency of each identified initial event are presented. In order to implement the analysis, the whole auxiliary building is divided into several areas, depending on elevation, isolation and function. Areas that are essential to plant operation and that house safe shutdown equipment are screened through a two-step process to determine the susceptibility to an internal flood. After this process, areas that are not eliminated are where flooding scenarios may occur. All possible flooding scenarios are reviewed to determine if they could result in any of the initiating events identified in the internal initiating events analysis which belongs to Level 1 PSA. For those flooding scenarios that do not result in an existing initiating event, a new initiating event is developed. Eventually, the frequency of each initiating event is achieved by identifying each flooding source and its failure mode, and correlating this to failure data. It must be paid attention to that the identification of internal flood hazards is plant specific since the routing of piping systems varies substantially from plant to plant. Therefore, the conclusion obtained from this issue may be not applicable for other nuclear power plant.
Researches on the In-Core Reload Design for TianWan Nuclear Power Station
Authors: Wang Hongxia
Published: April 2011

The two Units of TianWan Nuclear Power Station (phase 1) are in their third fuel cycle operation now. China Nuclear Power Engineering Co., Ltd. (CNPE) has finished the reload design of Unit 1 and the check on Russia’s computation of Unit 2 in the third cycle. From the forth cycle on, the in-core reload design will be given by Chinese part. It is summarized in this paper the problems met and the way to solve them during our design process. We discussed how to retain the related parameters, such as maximum burn-up and power peak factor, in reasonable ranges in the first part. In the last part, four design cases which can be used in cycle 5 for Unit 1 are recommended and evaluated briefly. The first two cases are partly low-leakage schemas; they satisfy the needs of the related restrictions very well. Case 3 is completely a low-leakage schema and case 4 is a long life schema. For case 3 and 4, a few new kind of assemblies, which are different from the present assemblies in uranium enrichment or the numbers of Oxide Gadolinium fuel rods, may be adapt to for better parameters.
Simulation and Analysis: The Dose Distribution of KBS-3 Spent Nuclear Fuel Canister by MCNP
Authors: Yang Bo , Wu He-xi , Liu Yi-bao
Published: April 2011

With the sustained and rapid development of the nuclear power plants, the spent fuel which is produced by the nuclear power plants will be rapidly rising. Spent fuel is High-level radioactive waste and should be disposed safely, which is important for the environment of land, public safety and health of the nuclear industry, the major issues of sustainable development and it is also necessary part for the nuclear industry activities. It is important to study and resolve the high-level radioactive waste repository problem. Spent nuclear fuel is an important component in the radioactive waste, The KBS-3 canister for geological disposal of spent nuclear fuel in Sweden consists of a ductile cast iron insert and a copper shielding. The ductile cast iron insert provides the mechanical strength whereas the copper protects the canister from corrosion. The canister inserts material were referred to as I24, I25 and I26, Spent nuclear fuel make the repository in high radiant intensity. The radiation analysis of canister insert is important in canister transport, the dose analysis of repository and groundwater radiolysis. Groundwater radiolysis, which produces oxidants (H2O2 and O2), will break the deep repository for spent nuclear fuel. The dose distribution of canister surface with different kinds of canister inserts (I24, I25 and I26) is calculated by MCNP (Ref. 1). Analysing the calculation results, we offer a reference for selecting canister inserts material.
Acid Stimulation Used in In-Situ Leaching Uranium
Authors: Liao Wen-sheng , Wang Li-min , Yao Yi-xuan , Jiang Guo-ping , Zhao Hai-jun , Huang Xin
Published: April 2011

Acidization was studied on a uranium sandstone deposit in Inner Mongolia with low–permeability and heavy calcium cementation. Acid dissolving test indicates that hydrochloric acid, formic acid and mud acid can easily dissolve formation minerals. With proper volumes and concentrations of acids used, the risk of precipitation of reaction products could be minimized. Core flow acidizing trial shows that the acidic fluid systems of hydrochloride acid, formic acid or acetic acid can significantly improve the core permeability. The average permeability has increased by 763 percent for the above three systems. But mud acid didn’t increase the core permeability; on the contrary, it caused formation damage, and led to lowering permeability. In the pilot test, the injection rate has improved by 118 percent for 6 wells. The acid treatment results indicate that a significant production enhancement of wellfields can be achieved by acid stimulation.
Application of Neutron Radiography to Observe Water Absorbtion of Concrete
Authors: Zhang P. , Zhao T. J. , Zhang L. S. , Wittmann F. H. , Lehmann E. , Vontobel P.
Published: April 2011

It has been experienced that service life of reinforced concrete structures is often limited due to lack of durability of cement-based materials. One major reason for this durability problem is the penetration of water and compounds dissolved in water into concrete. Therefore, there is an urgent need to study water penetration into concrete in order to better understand deterioration mechanisms. Neutron radiography provides an advanced non-destructive technique with high spatial resolution. In this contribution, neutron radiography was successfully utilized to study the process of water absorption of two types of concrete with different water-cement ratios namely 0.4 and 0.6. It is shown that it is possible to visualize migration of water into concrete and to quantify the time-dependent moisture distribution with accurately and with high spatial resolution by means of neutron radiography. In concrete with high water-cement ratio, water penetrates much quicker than in concrete with lower water cement ratio. Water penetration depth obtained from neutron radiography is in good agreement with corresponding values obtained from capillary suction tests. Experimental results obtained by means of neutron radiography on water penetration into concrete will be presented and discussed in this contribution. Results will provide us with a solid basis for a better understanding of deteriorating processes in concrete and other cement-based materials. These results may be considered to be a first step to improve durability of concrete.
Concept Design of Computer Monitoring System for ITER Transfer Cask System
Authors: Wang Haitian , Li Ge , Cao Lei
Published: April 2011

The ITER (International Thermonuclear Experimental Reactor) is an international collaborative project, and its object is aiming at demonstrating the scientific and technological feasibility of fusion energy for peaceful purposes. In the ITER, the four important engineering challenges are the first wall of the blank model, the remote handling (RH), the heating of the plasma, and the superconducting technology. The RH control system is very complexity, and the total number of the control nodes is 15378, which consists of the control nodes in the Tokamak equipment, the hot cell equipment and the control room equipment. Only the number of the control nodes of the transfer cask system (TCS) is up to 2580. The TCS is one sub system of ITER RH system, which provides the means for the remote transfer of (clean/activated/contaminated) in-vessel components and remote handling equipment between hot cell facility and vacuum vessel through dedicated galleries and lift in the ITER buildings. Due to the experimental facility with a very long timeline, the control software on the ITER is better suited using an open source solution as compared to a commercial solution. According to market share and proven record, command control and data acquisition and communication (CODAC) group chooses the EPICS (Experimental Physics and Industrial Control System) as the preferred solution. China takes part in the ITER, and wishes to grasp the RH technology, which is one of the four key technologies related to the future fusion reactors for electric power generation. According to the CODAC system configuration with EPICS component, the concept control architecture of the TCS is presented in this paper. The control system consists of the file network, the interlock network, the safety network, real-time network, RH control network and visual supervising network etc. Then, the paper analyses each part corresponding requirement function. As a result, a development service platform concept is set up. Finally, the experiment system of the leg of the TCS has been constructed.
Dynamic Motion of RC Beams and One-Way Slabs Impacted by a Missile at Medium Velocity: Models for Spreadsheet Application
Authors: Rambach Jean-Mathieu
Published: April 2011

This paper deals with the simulation of the motion of a one-span slender rectilinear RC beam or a thin one-way RC rectangular slab when impacted by a missile, within and beyond the elastic domain, up to a maximal displacement equal to the height of the section. The loading is variable in time and in space and the supporting conditions at each extremity are either of simply resting type or of clamping type. The equation of motion for each beam segment is expressed, through finite differences method, by relations between velocity and flexural moment that are decomposed on the basis of the first N modes of beam deformation. The plastic deformations are taken into account by constraining the flexural moment to follow elastic/perfectly plastic flexural moment–curvature law. The algorithms are well adapted to a spreadsheet application that allows easily: i) presizing such a structure against this type of loading, ii) the sensitivity studies regarding the mechanical parameters, iii) checking the order of magnitude of the results of big FEM models by using highly sophisticated codes. Simulations by this method show good agreement with experimental results and give access to some mechanical properties of the structure when damaged.
Numerical Calculations for Air-Water Tests on CCFL in Different-Scale Models of a PWR Hot Leg
Authors: Murase Michio , Kinoshita Ikuo , Utanohara Yoichi , Lucas Dirk , Valle´e Christophe , Tomiyama Akio
Published: April 2011

Air-water CCFL (countercurrent flow limitation) tests using the 1/5th scale rectangular channel and 1/15th scale circular tube simulating a PWR hot leg and both air-water and steam-water CCFL tests using the 1/3rd scale rectangular channel were previously carried out at Kobe University and Forschungszentrum Dresden-Rossendorf (FZD), respectively. In this paper, numerical calculations for the air-water CCFL tests at FZD using FLUENT6.3.26 are presented and compared with the experimental data at Kobe University and FZD. In the calculations, the VOF (volume of fluid) model or two-fluid (2F) model was used. Major results were as follows: (1) the calculated CCFL characteristics using the 2F model for the FZD tests agreed well with the 1/15th scale circular tube data obtained at Kobe University and the calculated results for full-scale PWR conditions, which supported the validity of the 1/3rd scale rectangular channel to simulate CCFL in circular tubes; and (2) comparison with the FZD data showed that the calculations using the 2F and VOF models overestimated the falling water flow rates.
The Operation Characteristics of the Fuel Handling System of HTR-10
Authors: Huang Peng , Liang Xihua , Chen Xiaoming
Published: April 2011

10 MW High Temperature Gas-cooled Reactor (HTR-10) is a pebble bed reactor, helium serves as the coolant, the fuel element is spherical, and its diameter is 60 mm. In power operation, the reactivity of HTR-10 is almost kept in a constant, because the fuel elements could be loaded, discharged and reloaded online, all these processes referred above are carried out automatically by the fuel handling system. In this paper, the operation characteristics of the fuel handling system, including the features of some special equipments, three automatic processes of fuel elements handling, the configuration and operation features of the control system and the operation conclusion in these years since 2000 are introduced.
Measurement of Neutron and Gamma Spatial Distribution in Radial Channel at VR1 Reactor
Authors: Miletic´ Marija
Published: April 2011

This work gives the theoretical description with a practical measurement of neutron and gamma spatial distributions in a radial experimental channel (RC) at VR1 research reactor. For providing the measurements an “off-line” and “on-line” methods were used. The “on-line” measurement of neutron spatial distribution was performed with a 3He gas filled detector. From the “off-line” measurements Neutron Activation method (for neutron detection) and Thermoluminescent method (for gamma radiation detection) were chosen. Results obtained from experiments were compared with the results from MCNP code.
Design, Operation and Maintenance of Thermal Hydraulic Instrumentation System of HTR-10
Authors: Wei Liqiang , Zhong Shuoping
Published: April 2011

The 10 MW high temperature gas-cooled reactor–test module (HTR-10) reached the first critical at the end of year 2000, and has been running for over 9 years safely and stably till now. Comparing with the Pressure Water Reactor (PWR), HTR-10 has many different characteristics, such as core construction, special fuel elements, helium coolant and so on. Thus the thermal hydraulic parameter measurement has special requirement and it is indispensable to select or develop some new class 1E instrumentation and devices. This paper describes measurement requirements, measurement method and measurement instrumentations for measuring coolant temperature, primary loop pressure, primary loop mass flow rate, primary loop humidity, main steam pressure, feedwater mass flow rate, in-core components temperature, pressure vessel surface temperature. The class 1E sheathed thermocouples and thermocouple penetration assembly, the class 1E orifice plate throttle device, and the data acquisition and supervision system that were developed by the Institute of Nuclear and New Energy Technology (INET) according to the criteria IEEE 323 and IEEE 344 are introduced in detail. HTR-10 has been operated successfully for over 9 years up to the present. The operation and maintenance experience of above-mentioned instrumentations shows they are safe and reliable at normal and abnormal conditions. The experience described in this paper is valuable for the latter 2 × 250 MW modular high temperature gas-cooled reactor.
Condensate Polishing System for PWR Nuclear Power Plant
Authors: Wang Ailing
Published: April 2011

This paper provides summary of domestic in service nuclear power plant condensate polishing system (CPS) configuration, CPS is supplied for all domestic in service nuclear power plant and almost all above sub–critical fossil power plant, so condensate polishing system is necessary for PWR nuclear reactor. Based on USA industrial practice, the raw condensate for startup and normal operation is presented in this paper. Base on EPRI PWR nuclear reactor secondary water chemistry guideline for steam generator blowdown water quality, ion and water balance in secondary system, the effluent water quality for CPS is calculated. The condensate flow is proportional with electric power, during the plant startup operation, the electric power level from 0 to 30% to 50% to 100%, is longer time, i. e. condensate flow from 0 to 30% to 50% to 100%, is longer time. Before the plant electric power level 50% (or 30%), the impurity in condensate is purified, the makeup water for secondary system is demineralied water, the impurity level is lower than requirement, so 50% (or 30%) flow condensate polishing system is enough for startup operation. During the plant normal operation, condensate polishing system is in hot standby, if the condensate tube leak, condensate is deteriorated, the condensate polishing system need to be operated immediately, but CPS put in service from hot standby at least 5 minute. During this period, the units which is once–though cooled by sea water will be shutdown immediately with 2.71/h continuous condensate tube leakage, so full flow CPS is necessary for the units which is cooled by sea water, and it is better that one or two series put in service during normal operation other than all of them in hot standby. The units which is cooled by fresh water will be shutdown immediately with 2561/h continuous condensate tube leakage, for this level leakage, whichever 50%(30%) and 100% CPS, action level 2 will be preformed for the units, so 50%(30%) is enough for fresh water cooled units.
Material Surveillance Program Regulatory Guidance for Steam Generator Tubes Extracted From Canadian CANDU Reactors
Authors: Liu Sue , Carroll Blair , Riznic Jovica , Christodoulou Nicholas , Awad Raoul
Published: April 2011

The Canadian Nuclear Standard, CSA N285.4, requires periodic extraction of steam generator tubes for metallurgical examination; however, the Standard does not specify detailed program requirements. Therefore, Canadian Nuclear Safety Commission (CNSC) staff has recently brought forth a set of guidelines to clarify the regulatory expectations for steam generator tube material surveillance programs. This paper introduces the material surveillance program required by CSA N285.4, and provides a discussion of the CNSC staff proposed guidelines, to explain the technical basis behind the expectations for tube removal and examination. The objectives of the steam generator tube material surveillance program include evaluating inspection tool capabilities, confirming tube material performance over the life of the plant and assessing the operating environment’s potential contribution to tube degradation. The knowledge acquired from the program can reduce uncertainties associated with the non-destructive inspection practices, and can provide data for steam generator aging management programs. Over the life of a plant, CSA N285.4 requires that only a few tubes be extracted for examination. However, each tube removal is costly and represents significant personnel dose. The tube removal and examination guidelines will aid in maximizing the value of each tube extracted for metallurgical examination.
Effect of Temperature on the Ni-63 Betavoltaic Cell
Authors: Wang Guanquan , Hu Rui , Wei Hongyuan , Zhang Huaming , Yang Yuqing , Xiong Xiaoling , Luo Shunzhong
Published: April 2011

There are significant relations between temperature and the electrical performances of betavoltaic cell. Two silicon diodes as the energy conversion device of betavoltaic cell were irradiated by Ni-63, checked the relations between temperature and the electrical performances such as Voc, Isc, Pmax etc. Isc increased very little as the temperature increasing but Voc decreased significantly. The changing values of Voc were −3.1 mV/k and −3.0 mV/k respectively within the temperature range of 233.15 K∼333.15 K. As a result of this, the Pmax, η etc. decreased markedly too.
Verification and Validation of AEGIS/SCOPE2: The State-of-the-Art In-Core Fuel Management System for PWRs
Authors: Tatsumi Masahiro , Ohoka Yasunori , Endo Tomohiro , Sugimura Naoki , Tabuchi Masato , Hyoudou Hideaki , Ushio Tadashi , Mori Masaaki , Yamamoto Akio
Published: April 2011

Verification and validation of AEGIS/SCOPE2, a state-of-the-art in-core fuel management system for PWRs, was conducted. Verification of implementations for processes of tabulation and reconstruction of cross sections data in the system was confirmed through benchmark problems to confirm consistency between AEGIS and SCOPE2 in terms of reactivity and pin power distribution with identical computational condition in single- and multi-assembly geometry. In validation study numerical performance of the system was demonstrated in analyses of the B&W’s critical experiment and tracking calculations for commercial PWRs in combination of ENDF/B-VII.0 library. It is found the present system gives stability in prediction of critical boron concentration and radial power distribution.
Investigations and Countermeasures for Deactivation of the Hydrogen Recombination Catalyst at Hamaoka Unit 4 and 5
Authors: Kawasaki Toru , Aizawa Motohiro , Iizuka Hidehiro , Yamada Koji , Kugimoto Mitsuo
Published: April 2011

At Hamaoka Unit 4 and 5, the hydrogen concentration in the outlet of off-gas recombiner had increased, and the reactors could not continue start-up operation. Therefore, we investigated the causes of the deactivating the recombination reaction and selected appropriate countermeasures to the plants. From our investigation, two types of deactivation mechanism are found. One of the causes was decreasing the active surface area of alumina as support material by the dehydrative condensation. The other cause was poisoning of the catalyst by organic silicon compound. The organic silicon was introduced from organosilicon sealant used at the junctions of the low-pressure turbine. We also found that the boehmite rich catalyst was deactivated more easily by the organic silicon than gamma alumina because boehmite had a lot of hydroxyl groups. Finally, we estimated that the deactivation of the hydrogen recombination catalysts was caused by combined two factors, which are characteristics of boehmite catalyst support and the poisoning by the organic silicon on the catalyst surface. As the countermeasures, the boehmite was changed into more stable gamma alumina by adding the heat treatment in hydrogen atmosphere at 500°C for 1 hour, and the source of organic silicon, organosilicon sealant, was removed. At Hamaoka Unit 4 and 5 improved catalysts were applied. Moreover, linseed oil that used to be used at the plants was applied again as sealant of the low-pressure turbine casing instead of the organosilicon sealant. As a result of application of these countermeasures, the reactors could be started without increase of the hydrogen concentration at these plants.
Development of Large Size Moisture Separator Reheater for Up to 1700 MW PWR Class
Authors: Fujita Issaku , Sakata Teruaki , Kojima Toshiki
Published: April 2011

The recent trend in increasing the power generation capacity of pressurized water reactors (PWRs) up to the 1700MW Class for greater economy is met with inherent challenges. Higher generation capacity necessitates turbines with higher efficiency. Operating the high pressure (HP) turbine at higher efficiency requires development of large size moisture separator reheater (MSR) to accommodate the higher specific steam volume due to the reduced HP exhaust pressure. Higher specific volume of steam also results in higher velocities and increased pressure drops. At higher steam velocity, the flow accelerated corrosion (FAC) is enhanced and the moisture separator performance will be deteriorated due to the increased mist carry over across the Chevron type vanes. The development of MSR for up to 1700MW Class PWR involved optimizing the heat balance around the MSR and selecting the optimized size of the MSR for performance and cost. This also included selection of an optimized terminal temperature difference (TTD) for the MSR. Then the newly developed MSR was verified by computational (using CFD) and experimental techniques. Using CFD, the pressure, flow, and velocity distribution, pressure drop, and velocity distribution on the shell surface were analyzed. An experimental set up using actual size sliced model of the MSR was used to predict the separator performance and drainage characteristics. The testing was carried out with two phase air-water flow at atmospheric conditions. This paper summarizes the large size MSRs for up to 1700 MW Class PWR, its development, and design verification.
3D Simulation of Molten Metal Freezing Behaviour Using Finite Volume Particle Method
Authors: Mahmudah Rida S. N. , Kumabe Masahiro , Suzuki Takahito , Guo LianCheng , Morita Koji , Fukuda Kenji
Published: April 2011

Understanding the freezing behavior of molten metal in flow channels is of importance for severe accident analysis of liquid metal reactors. In order to simulate its fundamental behavior, a 3D fluid dynamics code was developed using Finite Volume Particle (FVP) method, which is one of the moving particle methods. This method, which is fully Lagrangian particle method, assumes that each moving particle occupies certain volume. The governing equations that determine the phase change process are solved by discretizing its gradient and Laplacian terms with the moving particles. The motions of each particle and heat transfer between particles are calculated through interaction with its neighboring particles. A series of experiments for fundamental freezing behavior of molten metal during penetration on to a metal structure was also performed to provide data for the validation of the developed code. The comparison between simulation and experimental results indicates that the present 3D code using the FVP method can successfully reproduce the observed freezing process such as molten metal temperature profile, frozen molten metal shape and its penetration length on the metal structure.
Investigation of Subcooled Flow Boiling Model Under Low Pressure
Authors: Pu Fan
Published: April 2011

Two fluid model integrating a set of closure relationships (such as inter-phase heat transfer model, inter-phase mass transfer model, inter-phase momentum transfer model, mean bubble diameter model, bubble departure diameter model, bubble departure frequency model, onset of nucleate boiling model, wall heat flux partition model) are applied to solve the local flow and heat transfer of subcooled flow boiling under low pressure, and local flow parameters such as volume fraction, liquid velocity, vapor phase etc. are obtained using developed code. The subcooled flow boiling results predicted in this paper are compared to the subcooled flow boiling experimental results of TH Lee etc. in annular channel with inner tube heated uniformly and adiabatic outer tube, and they agree well. The code can also be used to predict the multiphase flow field in power reactor core when the subcooled flow boiling take place and the critical heat flux of Departure Nucleate Boiling (DNB) in reactor core and it is meaningful for reactor core design.
The Swiss Master in Nuclear Engineering: A Collaboration Between Universities, Research Centre and Industry
Authors: Chawla R. , Ansermet J. P. , Cavedon J. M. , Hirt P. , Kro¨ger W. , Prasser H. M. , Tran M. Q.
Published: April 2011

The two national technical universities in Switzerland, viz. the Swiss Federal Institutes of Technology at Lausanne (EPFL) and at Zurich (ETHZ) have a rich and long tradition in nuclear education. Student research in nuclear engineering, particularly at the doctoral level, has usually been conducted in collaboration with the Paul Scherrer Institute (PSI) at Villigen, the national research centre where most of the country’s fission-related R&D is carried out. A significant part of this R&D is carried out in close collaboration with the Swiss Nuclear Utilities (swissnuclear). The four above, key national players in nuclear teaching and research in Switzerland — EPFL, ETHZ, PSI and swissnuclear — have recently pooled resources in implementing a new Master of Science degree in Nuclear Engineering (NE). The present paper describes the main features and experience acquired to date in the running of this, first-ever, common degree offered jointly by the two Swiss Federal Institutes of Technology. The program, although naturally addressing Switzerland’s needs, is clearly to be viewed in an international context, e.g. that of the Bologna Agreement. This is reflected in the composition of the first two batches, with about 70% of the students having obtained their Bachelor degrees from universities outside Switzerland. Starting September 2010, the curriculum of the EPFL-ETHZ NE Master will be upgraded, from its current 90 ECTS credit points (3 semesters) to a 120 ECTS (4 semesters) program. An overview is provided of the current 90-ECTS curriculum, as also a sketch of the changes foreseen in going to 120 ECTS.
Upwinding Meshfree Point Collocation Method for Steady MHD Flow
Authors: Cai Xinghui , Su Guanghui , Qiu Suizheng
Published: April 2011

In this paper, a meshfree point collocation method, with a upwinding scheme, is presented to obtain the numerical solution of the coupled equations in velocity and magnetic field for the fully developed magnetohydrodynamic (MHD) flow through a straight pipe of rectangular section with insulated walls. The moving least-square (MLS) approximation is employed to construct shape functions in conjunction with the framework of point collocation method. Computations have been carried out for different applied magnetic field orientations and different Hartmann numbers from 5 to 1,000,000. As the adaptive upwinding local support domain is introduced in the meshless collocation method, numerical results show that the method can compute MHD problems not only at low and moderate values but also at high values of the Hartmann number with high accuracy and good convergence.
Fortum Participation in IAEA Benchmark for KKNPP7 RHR Piping Response NCO 2007 Event
Authors: Varpasuo Pentti , Ka¨hko¨nen Jukka
Published: April 2011

In this report Fortum participation in a benchmark related to the residual heat removal (RHR) piping system response of the Kashiwazaki-Kariwa Nuclear Power Plant unit 7 (KKNPP7) of Tokyo Electric Power Company (TEPCO) in Japan, during the 16 July 2007 Niigataken-Chuetsu-Oki earthquake (NCO) is described. The goal of this benchmark is to conduct a comparison between different analytical techniques, as used in the usual engineering practice. Equipment behavior during the NCO earthquake constitutes an extensive database. There are no direct acceleration or displacement measurements on equipment. For this reason quantitative simulation comparison is not possible. Qualitative observations of the fuel tank sloshing and tank buckling are available and these can be used in numerical simulation benchmark. The benchmark is divided in three phases and it will be carried out during the years of 2009–2011. The number of participants in the benchmark is 25 organizations and institutions. Detailed goals of the benchmark as a whole can be summarized as follows: 1) Understanding of soil, structures and mechanical equipment response during the Niigataken-Chuetsu-Oki earthquake. 2) Simulation of equipment response for residual heat removal equipment. 3) Simulation of liquid sloshing in the fuel pools in reactor building. 4) Simulation of buckling phenomena of vertical, cylindrical yard tanks. In order to carry out this simulation task a joint model of the reactor building and the residual heat removal-system is developed. The main results of the numerical simulation will be the maximum values of the stresses in the most critical locations of the residual heat removal piping system. These values will be compared to known material properties of the piping such as yield and the tensile failure strain.
An Open-Walled Ionization Chamber Appropriate to Tritium Monitoring for Glovebox
Authors: Chen Zhilin , Wang Heyi
Published: April 2011

An open-walled ionization chamber is developed to monitor the tritium concentration in glovebox in the tritium processing systems. Two open walls are used to replace the sealed wall in common ionization chambers, through which the tritium gas can diffuse into the chamber without the aid of pumps and pipelines. Some basic properties of the chamber are examined to evaluate its performance. Results turn out that an open-walled chamber of 1 L in volume shows considerably flat plateaus over 700 V. The chamber also gives a good linear response to the gamma field over five degrees under condition of 1 atm. The pressure dependence characters show that the ionization current is only sensitive to low pressure. The pressure influence becomes weaker as pressure increases mainly due to the decrease of the mean free path of β ray decayed by tritium. The minimum detection limit of the chamber is 3.7×105 Bq/m3.
Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor
Authors: Zhang Aimin , Kang Yalun
Published: April 2011

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.
Development and Preliminary Verification of a Thermal-Hydraulic Multi-Scale Coupled Code
Authors: Liu Yu , Zhang Hong , Jia Baoshan
Published: April 2011

On the basis of best estimate thermal-hydraulic system code RELAP5, sub-channel code COBRA-1V, and commercial Computational Fluid Dynamics (CFD) code CFX, a thermalhydraulic multi-scale coupled code RECOX has been developed. The coupling strategy was designed to keep the integral structure of each code and minimize modifications of code source. Under the Parallel Virtual Machine (PVM) environment, an external control code has been developed to perform codes spawn, data exchange and mapping, time step coordination, etc. Two test cases including single phase blowdown and temperature fluctuation transient have been carried out to evaluate the coupling between codes. Compared with stand-alone simulations very good agreement was achieved. Then in order to demonstrate the coupled analysis capability of RECOX, an asymmetry transient in a simple two loops system which is similar to the nuclear power plant was simulated. The result is correct and reliable, although further verification of coupled code with related experiment is needed. Finally, some potential improvements of coupling and future work were presented.
Irradiation-Embrittlement of Reactor Pressure Vessel Steel
Authors: Li Guoyun , Wu Yukun , Jiang Guofu , Huang Juan , Zhang Haisheng
Published: April 2011

Irradiated for 4 cycles inside the reactor, the irradiation surveillance capsule (ISC) was withdrawn and transported to Nuclear Power Institute of China (NPIC) for dismantling in the hot cell. This paper presented tensile tests and Charpy-V impact tests on non-irradiated and irradiated specimens. Tensile tests were performed at room temperature and 300 °C respectively, and Charpy-V impact tests were performed at series temperatures. Instrumented impact curves and tensile curves were obtained and analyzed. Impact absorbed energy, lateral expansion and crystallinity transition curves were plotted by a hyperbolic tangent function. The irradiation-induced shift of transition temperature, upper shelf energy, tensile strength and yield strength were determined. And irradiation-embrittlement effect on RPV was evaluated by prediction and measurement of irradiation-induced transition temperature shift. Results showed that the parameters including neutron fluence and transition temperature were in normal range, and the next capsule could be withdrawn according to original irradiation monitoring program.
Investigation on the Accident of Loss of Offsite Power of CARR
Authors: Lu Qing , Qiu Suizheng , Tian Wenxi , Su Guanghui
Published: April 2011

A code has been developed with proper models for the thermal-hydraulic simulation of research reactors, and the accident of loss of offsite power of CARR (the China Advanced Research Reactor) has been analyzed. It is found that the transient can be departed into four parts with two peaks, and at last terminated to a natural circulation state. Furthermore, in order to investigate the effect of the height of the in core structures during this accident, several structure changes have been proposed and analyzed. It is demonstrated that the increase of the vertical height of the in core structures not always do help to decreasing the peak temperatures and increasing the natural circulation flow in the transient.