bluuebunny/crossref_metadata_2025 huggingface dataset:
Fluid-Structure Interaction of a Steam Generator Tube in a Cross-Flow Using Large-Eddy Simulation
Authors: Kuehlert Karl , Webb Stephen , Joshi Mandar , Schowalter David
Published: September 2008

The behavior of a steam generator tube in a cross flow is examined by coupling the ABAQUS structural analysis code to Large-Eddy Simulation (LES) within the FLUENT Computational Fluid Dynamics solver. The moving tube is assumed to be within an infinite staggered array of other tubes, the centers being separated in the streamwise direction by 1.31 times the tube diameter and in the cross-stream direction by 2.73 times the tube diameter of 0.688 inches. The Reynolds number based upon total mass flow and one tube diameter is 1.0 × 105, indicating a transcritical regime, meaning that the boundary layer is expected to be transitional. The study begins with investigation of free two degree-of-freedom simulations of an individual tube, which are performed at low Reynolds number (3,800), where results are compared favorably to experimental results. High Reynolds number (3 × 106) two degree-of-freedom single tube simulations are also demonstrated. This is followed by a comparison between experiment and computation of flow past a stationary tube bank. Finally, in-phase free oscillation of the single row of tubes is simulated. System behavior in terms of oscillation frequency and amplitude, along with qualitative flow characteristics are discussed.
A Method to Quantify Plant Availability and Initiating Event Frequency Using a Large Event Tree, Small Fault Tree Model
Authors: Kee Ernest J. , Sun Alice , Rodgers Shawn , Popova Elmira , Nelson Paul , Moiseytseva Vera , Wang Eric
Published: September 2008

South Texas Project uses a large fault tree to produce scenarios (minimal cut sets) used in quantification of plant availability and event frequency predictions. On the other hand, the South Texas Project probabilistic risk assessment model uses a large event tree, small fault tree for quantifying core damage and radioactive release frequency predictions. The South Texas Project is converting its availability and event frequency model to use a large event tree, small fault in an effort to streamline application support and to provide additional detail in results. The availability and event frequency model as well as the applications it supports (maintenance and operational risk management, system engineering health assessment, preventive maintenance optimization, and RIAM) are briefly described. A methodology to perform availability modeling in a large event tree, small fault tree framework is described in detail. How the methodology can be used to support South Texas Project maintenance and operations risk management is described in detail. Differences with other fault tree methods and other recently proposed methods are discussed in detail. While the methods described are novel to the South Texas Project Risk Management program and to large event tree, small fault tree models, concepts in the area of application support and availability modeling have wider applicability to the industry.
Impact of Support Plate Flow Jets on Rod Vibration for Inlet Flow Region
Authors: Joffre Paul F. , Karoutas Zeses E.
Published: September 2008

An experimental investigation was performed to compare the axial velocity profiles occurring downstream of the inlet nozzle region of nuclear PWR fuel assemblies for two bottom nozzle designs. Axial velocities were measured in a 3.763:1 over-scale air test section simulating a 6×6 rod array of the inlet nozzle region and first grid span. The measurements were taken in multiple planes at 23 positions to map the velocity in both rod gaps and subchannels. Pressure drop measurements were also taken at selected axial elevations. The two bottom nozzle designs were tested to study how each nozzle impacts the dissipation of the lower support flow hole jet velocity downstream of the bottom nozzle. The nozzle designs included a standard round-hole flow plate and a slotted flow plate. The velocity profile results with these bottom nozzle flow plate designs are of particular interest since these bottom nozzle designs had previously been used in full scale out-of-pile water flow tests with contrasting results on wear at the first grid elevation. There were no rods with fretting wear in the test with the standard round hole flow plate and ten rods with fretting wear with varying depth in the test with the slotted flow plate. The full scale flow tests were performed at reactor operating temperatures (620°F), pressures (2250 psia) and at the same test flow rate. The grid cell settings were unchanged for the two tests. Therefore, the only test variable that differed for the full scale flow tests was the bottom nozzle design. The axial velocity profile investigation provided evidence that the standard round-hole flow plate dissipated the lower support jet velocity much better than the slotted plate design. In addition, based on the air test results, the slotted plate design had the largest axial velocity gradients at the same location as the observed rod wear. Based upon this investigation, it is concluded that excessive rod vibration in this region can occur due to high axial jet velocities and steep axial velocity gradients generated from the flow holes in the lower support plate. The excessive rod vibration can lead to fuel rod wear and fuel failure.
Two-Phase Flow Patterns in a Four by Four Rod Bundle
Authors: Mizutani Yoshitaka , Hosokawa Shigeo , Tomiyama Akio
Published: September 2008

Air-water two-phase flow patterns in a four by four square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12 mm in diameter were observed by utilizing a high speed video camera, FEP (fluorinated ethylene propylene) tubes for rods, and a fiberscope inserted in a rod. The FEP possesses the same refractive index as water, and thereby, whole flow patterns in the bundle and local flow patterns in subchannels were successfully visualized with little optical distortion. The ranges of liquid and gas volume fluxes,  and , in the present experiments were 0.1 <  < 2.0 m/s and 0.04 <  < 8.85 m/s, which covered typical two-phase flow patterns appearing in a fuel bundle of a boiling water nuclear reactor. As a result, the following conclusions were obtained: (1) the region of slug flow in the  –  flow pattern diagram is so narrow that it can be regarded as a boundary between bubbly and churn flows, (2) the boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows of the Mishima & Ishii’s flow pattern transition model, and (3) the boundary between churn and annular flows is well predicted by the Mishima & Ishii’s model.
System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink
Authors: Lin Meng , Hou Dong , Xu Zhihong , Yang Yanhua , Zhang Ronghua
Published: September 2008

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I&C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I&C system. Matlab/Simulink, a scientific computation software, just can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I&C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I&C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results.
Behavior of Full-Scale Model Cask Under 9m Drop Test and Simulation
Authors: Hojo Kiminobu , Kimura Tadashi , Kitada Akio , Tamaki Hiroshi , Kishimoto Junich , Saito Yuich
Published: September 2008

The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility. Leak tightness and subcriticality on transportation required from IAEA TS-R1 [1] have to be assured by a 9m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask. The test was conducted by German Federal Institute for Materials Research and Testing (BAM) at their test facility in Horstwalde, Germany and comparison of the test result with the “MH1 (Mitsubishi Heavy Industries, Ltd.)” numerical simulation using LS-DYNA code. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: • The leak rate of He gas after the drop tests satisfied the IAEA’s criteria. • The numerical simulation which modeled the cask body enabled dynamic response such as acceleration and strain of the cask body. This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior.
The Specifics of RBMK Core Cooling at Overheated Conditions
Authors: Uspuras Eugenijus , Urbonavicius Egidijus , Kaliatka Algirdas
Published: September 2008

Reactor RBMK-1500 of Ignalina NPP is a boiling light water reactor with graphite moderator. Several important design features of RBMK-1500 are unique and extremely complex in respect to western reactors: the fuel assemblies are loaded into individual channels rather than a single pressure vessel; the plant can be refueled on-line; the neutron spectrum is thermalized by a massive graphite moderator. The reactor coolant system consists of two loops, each having flow length of more than 200 m. There are 1661 of vertical parallel fuel channels and numerous components, such as headers, pumps, valves, etc. The fuel channels and fuel claddings are made of Zirconium-Niobium alloy. From the point of view of safety barriers, each fuel channel in RBMK-type reactor corresponds to a pressure vessel of vessel-type reactors. Thus, the fuel channels are the most important element in reactor cooling system. However, in case of beyond design basis accidents with loss heat removal from the core the integrity of fuel channels could be challenged as they are not so strong as the pressure vessel. The paper presents the analysis of different possibilities to cooldown the core of RBMK-type reactors. Injection of water to RCS is considered as main strategy. Such “bleed and feed” procedure is used for vessel type reactors, but at present is not considered at RBMK-1500.
Performance Evaluation of Control and Safety Rod and Its Drive Mechanism of Fast Breeder Reactor During Seismic Event
Authors: Chellapandi P. , Rajan Babu V. , Chetal S. C. , Baldev Raj
Published: September 2008

Control and safety rods & their drive mechanisms (CSRDM) and diverse safety rods & their drive mechanisms (DSRDM) are the main constituents of the two independent shutdown systems in Prototype Fast Breeder Reactor (PFBR) of 500 MWe. There are nine CSR and three DSR placed within the hexagonal sheaths, which in turn are located in two radial banks of reactor core. This paper deals with the analysis carried out to predict the performance of CSRDM along with CSR. Analysis is carried out for the static and seismic loading under the fuel handling as well as normal operating conditions with the objective of ensuring structural integrity as well as to estimate the drop time during seismic event. The effects of bowing of sub-assemblies corresponding to the end of life situation have been considered. From the analysis, it is found that the drop time of CSR is 0.82 s, which is less than 1 s, and hence there is no concern of reactor shutdown. Further, it is ensured that there is no mechanical interaction of concern between various parts. The induced stresses are found to be very much less than RCC-MR allowable stress intensity. Thus the performance of CSRDM and CSR is demonstrated to be sound during normal as well as seismic events.
Suppression Pool Mixing and Condensation Tests in PUMA Facility
Authors: Cheng Ling , Woo Kyoung Suk , Ishii Mamoru , Lim Jaehyok , Han James
Published: September 2008

Condensation of steam with noncondensables in the form of jet flow or bubbly flow inside the suppression pool is an important phenomenon on determining the containment pressure of a passively safe boiling water reactor. 32 cases of pool mixing and condensation test have been performed in Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility under the sponsor of the U.S. Nuclear Regulatory Commission to investigate thermal stratification and pool mixing inside the suppression pool during the reactor blowdown period. The test boundary conditions, such as the steam flow rate, the noncondensable gas flow rate, the initial water temperature, the pool initial pressure and the vent opening submergence depth, which covers a wide range of prototype (SBWR-600) conditions during Loss of Coolant Accident (LOCA) were obtained from the RELAP5 calculation. The test results show that steam is quickly condensed at the exit of the vent opening. For pure steam injection or low noncondensable injection cases, only the portion above the vent opening in the suppression pool is heated up by buoyant plumes. The water below the vent opening can be heated up slowly through conduction. The test results also show that the degree of thermal stratification in suppression pool is affected by the vent opening submergence depth, the pool initial pressure and the steam injection rate. And it is slightly affected by the initial water temperature. From these tests it is concluded that the pool mixing is strongly affected by the noncondensable gas flow rate.
“Long-Cell Action” Corrosion: A Basic Mechanism Hidden Behind Components Degradation Issues in Nuclear Power Plants
Authors: Saji Genn
Published: September 2008

In spite of industries’ effort over the last 40 years, corrosion-related issues continue to be one of the largest unresolved problems for nuclear power plants worldwide. There are several types of strange corrosion phenomena from the point of view of our current understanding of corrosion science established in other fields. Some of these are IGSCC, PWSCC, AOA, and FAC (Erosion-Corrosion). Through studying and coping with diverse corrosion phenomena, the author believes that they share a common basis with respect to the assumed corrosion mechanism (e.g., ‘local cell action’ hypothesis). In general, local cell action is rarely severe since it produces a fairly uniform corrosion. The ‘long cell action’ that transports electrons through structures far beyond the region of local cell corrosion activities has been identified as a basic mechanism in soil corrosion. If this mechanism is assumed in nuclear power plants, the structure becomes anodic in the area where the potential is less positive and cathodic where this potential is more positive. Metallic ions generated at anodic corrosion sites are transported to remote cathodic sites through the circulation of water and deposits as corrosion products. The SCC, FAC (E-C) and PWSCC occur in the anodic sites as the structure itself acts as a short-circuiting conductor between the two sites, the action is similar to a galvanic cell but in a very large scale. This situation is the same as a battery that has been short-circuited at the terminals. No apparent external potential difference exists between the two electrodes, but an electrochemical reaction is still taking place inside the battery cell with a large internal short current. In this example what is important is the potential difference between the local coolant and the surface of the structural material. Long cell action corrosion is likely enhancing the local cell action’s anodic corrosion activities, such as SCC, FAC/E-C, and PWSCC. It tends to be more hazardous because of its localized nature compared with the local cell action corrosion. There exist various mechanisms (electrochemical cell configurations) that induce such potential differences, including: ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author will discuss these potential differences and their relevance to the un-resolved corrosion issues in nuclear power plants. Due to the importance of this potential mechanism the author is calling for further verification experiments as a joint international project.
Management of Ultimate Risk of Nuclear Power Plants by Source Terms: Lessons Learned from the Chernobyl Accident
Authors: Saji Genn
Published: September 2008

The term ‘ultimate risk’ is used here to describe the probabilities and radiological consequences that should be incorporated in siting, containment design and accident management of nuclear power plants for hypothetical accidents. It is closely related with the source terms specified in siting criteria which assures an adequate separation of radioactive inventories of the plants from the public, in the event of a hypothetical and severe accident situation. The author would like to point out that current source terms which are based on the information from the Windscale accident (1957) through TID-14844 are very outdated and do not incorporate lessons learned from either the Three Miles Island (TMI, 1979) nor Chernobyl accident (1986), two of the most severe accidents ever experienced. As a result of the observations of benign radionuclides released at TMI, the technical community in the US felt that a more realistic evaluation of severe reactor accident source terms was necessary. In this background, the “source term research project” was organized in 1984 to respond to these challenges. Unfortunately, soon after the time of the final report from this project was released, the Chernobyl accident occurred. Due to the enormous consequences induced by then accident, the one time optimistic perspectives in establishing a more realistic source term were completely shattered. The Chernobyl accident, with its human death toll and dispersion of a large part of the fission fragments inventories into the environment, created a significant degradation in the public’s acceptance of nuclear energy throughout the world. In spite of this, nuclear communities have been prudent in responding to the public’s anxiety towards the ultimate safety of nuclear plants, since there still remained many unknown points revolving around the mechanism of the Chernobyl accident. In order to resolve some of these mysteries, the author has performed a scoping study of the dispersion and deposition mechanisms of fuel particles and fission fragments during the initial phase of the Chernobyl accident. Through this study, it is now possible to generally reconstruct the radiological consequences by using a dispersion calculation technique, combined with the meteorological data at the time of the accident and land contamination densities of 137Cs measured and reported around the Chernobyl area. Although it is challenging to incorporate lessons learned from the Chernobyl accident into the source term issues, the author has already developed an example of safety goals by incorporating the radiological consequences of the accident. The example provides safety goals by specifying source term releases in a graded approach in combination with probabilities, i.e. risks. The author believes that the future source term specification should be directly linked with safety goals.
Fatigue Evaluation of RCS Branch Line for License Renewal Application
Authors: Nam Il-Kwun , Youm Hagki , Jin Tae Eun , Kim Byung Sup
Published: September 2008

The fatigue effects have to be revaluated in preparing the license renewal application for the continued operation of an old vintage nuclear power plant. This paper presents a complete fatigue analysis for a branch piping with the effect of thermal stratification, induced by turbulent penetration, and environmental factors on fatigue. Three-dimensional computational fluid dynamics and finite element analyses were performed for the branch line to evaluate the thermal stratification loading. Proposed is a supplementary methodology of considering the effect of environmental factor on the combined conventional peak stress intensity range, based on the NB-3600 of ASME Section III Code, with thermal stratification loading. It can be used for safety enhancement of old vintage nuclear power plants.
Operation and Control of the PBMR Demonstration Power Plant
Authors: Kemp Petrus D. , Nieuwoudt Chris
Published: September 2008

A large interest in High Temperature Gas-cooled Reactors (HTGR) has been shown in recent years. HTGR power plants show a number of advantages over existing technology including improved safety, modular design and high temperatures for process heat applications. HTGR plants with closed loop direct cycle power conversion units have unique transient responses which is different from existing nuclear plants as well as conventional non-nuclear power plants. The operation and control for a HTGR power plant therefore poses new and different challenges. This paper describes the modes of operation for the Pebble Bed Modular Reactor (PBMR) demonstration plant. The PBMR demonstration plant is an advanced helium-cooled, graphite-moderated HTGR consisting of a closed loop direct cycle power conversion unit. The use of transient analysis simulation makes it possible to develop effective control strategies and design controllers for use in the power conversion unit as well as the reactor. In addition to plant controllers the operator tasks and operational technical specifications can be developed and evaluated making use of transient analysis simulation of the plant together with the control system. The main challenges in the operation and control of the reactor and power conversion unit are highlighted with simulation results. Control strategies in different operating regions are shown and results for the power conversion unit start-up transition and the loss of the grid connection during power operation are presented.
Experimental Evaluation of Ability of Relap5, Drako®, Flowmaster2™ and Program Using Unsteady Wall Friction Model to Calculate Water Hammer Loadings on Pipelines
Authors: Marcinkiewicz Jerzy , Adamkowski Adam , Lewandowski Mariusz
Published: September 2008

Mechanical loadings on pipe systems caused by water hammer (hydraulic transients) belong to the most important and most difficult to calculate design loadings in nuclear power plants. The most common procedure in Sweden is to calculate the water hammer loadings on pipe segments, according to the classical 1D theory of liquid transient flow in a pipeline, and then transfer the results to strength analyses of pipeline structure. This procedure assumes that there is quasi-steady respond of the pipeline structure to pressure surges — no dynamic interaction between the fluid and the pipeline construction. The hydraulic loadings are calculated with 1-D so-called “network” programs. Commonly used in Sweden are Relap5, Drako and Flowmaster2 — all using quasi-steady wall friction model. As a third party accredited inspection body INSPECTA NUCLEAR AB reviews calculations of water hammer loadings. The presented work shall be seen as an attempt to illustrate ability of Relap5, Flowmaster2 and Drako programs to calculate the water hammer loadings. A special attention was paid to using of Relap5 for calculation of water hammer pressure surges and forces (including some aspects of influence of Courant number on the calculation results) and also the importance of considering the dynamic (or unsteady) friction models. The calculations are compared with experimental results. The experiments have been conducted at a test rig designed and constructed at the Szewalski Institute of Fluid–Flow Machinery of the Polish Academy of Sciences (IMP PAN) in Gdansk, Poland. The analyses show quite small differences between pressures and forces calculated with Relap5, Flowmaster2 and Drako (the differences regard mainly damping of pressure waves). The comparison of calculated and measured pressures and also a force acting on a pre-defined pipe segment show significant differences. It is shown that the differences can be reduced by using unsteady friction models in calculations. Recently, such models have been subjects of works of several researches in the world.
Study on Characteristics of Thermal-Hydraulic Phenomena in Steam Injector
Authors: Takahashi Yuhki , Koizumi Yasuo , Ohtake Hiroyasu , Mori Michitsugu
Published: September 2008

Characteristics of thermal-hydraulic phenomena in the steam injector were examined. Two-types of a test section were used: a straight condensing section type and a taper condensing section type. In the straight condensing section type experiments, a water jet from a nozzle of 5 mm diameter flowed in to the test section concentrically. The inner diameter of the condensing section was 7, 10, or 20 mm and the length was 105 mm. Steam flowed into the peripheral space between the water jet and the inner wall of the condensing section. Experiments of the taper condensing section type were similar to those of the straight condensing section type. The inner diameter of the condensing section at the outlet of the water nozzle was 13.25 mm and the condensing section length was 52.9 mm. The inner diameter of the throat was 4 mm and the throat length was 5 mm. The vapor condensation rate, i.e. the condensation heat transfer rate, did not depend on the subcooling of the water jet. It was proved that the overall condensation heat transfer is controlled by the radial heat transport in the water jet since the condensation heat transfer resistance is much smaller than that of the radial heat transport in the water jet. The radial heat transport only depends on the jet flow rate. The criterion of the jet disruption was correlated by the Kelvin-Helmholtz instability wave length. The prediction provided conservative results; approximately one-tenth. When the condensing section was tapered such as the steam injector, the water jet became more stable. It was suggested that the accelerated water jet becomes more stable.
Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 13—Study on Heat Transfer of Direct Condensation of Steam on Subcooled Water Jet
Authors: Takahashi Yuhki , Koizumi Yasuo , Ohtake Hiroyasu , Miyashita Tohru , Mori Michitsugu
Published: September 2008

Characteristics of thermal-hydraulic phenomena in the steam injector were examined. In experiments, a water jet from a nozzle of 5 mm diameter flowed into the condensing test section pipe concentrically. The inner diameter of the condensing section was 7, 10, or 20 mm and the length was 105 mm. Steam flowed into the peripheral space between the water jet and the inner wall of the test section and condensed on the ware jet surface. The radial and the axial distributions of velocity and temperature of the water jet were measured. Analyses by using the STAR-CD code were also performed. The temperature measured in the central portion of the water jet was higher than the predicted assuming the ordinary turbulent flow in a pipe. The temperature measured in the peripheral region was lower than the predicted. The radial temperature distribution measured was flatter than the predicted. When the steam condensation rate was large, the measured radial velocity distribution in the water jet was flatter than the predicted. In the case that the steam velocity was quite high, the velocity measured in the peripheral region was higher than that in the center portion. These results implied that the steam condensing on the water jet brought momentum in the water jet to result in more effective radial transport of heat and momentum. The STAR-CD code analyses to allow the interface between the wall that simulated the steam flow part and the water flow that stood for the water jet to move, i.e. creating momentum in-flux at the water jet interface, provided better results to support the experimental results. To increase the interfacial friction had a minor effect on the radial velocity distribution in the tested range.
Modeling and Analysis of Thermal Damping in Heat Exchanger Tube Bundles
Authors: Khushnood Shahab , Malik Muhammad Afzaal , Khan Zaffar Muhammad , Khan Ayaz , Iqbal Qamar , Khalil Muhammad Shahid , Rashid Badar , Hussain Syed Zahid
Published: September 2008

Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles etc., are subjected to Flow Induced Vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as “thermal damping”. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.
Development of Outer Surface Irradiated Laser Stress Improvement Process (L-SIP)
Authors: Sugimoto Noriaki , Onitsuka Hironori , Okimura Koji , Ohta Takahiro , Kamo Kazuhiko
Published: September 2008

Improvement of residual stress is effective in a countermeasure to deal with the stress corrosion cracks in pipe welds. A irradiated laser stress improvement process (L-SIP) will be introduced as a method to improve residual stress inside steel pipes. This work method is to improve inner surface residual stress from tensile stress to compressive stress by irradiating laser beam around the welds of steel pipe and utilizing the temperature differences between inner and outer surface.
ATWS Frequency Evaluation of FBR MONJU
Authors: Sotsu Masutake , Kurisaka Kenichi
Published: September 2008

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor which can supply 280MW of electricity. The frequency of anticipated transient without scram (ATWS) of MONJU that used to be based on conservative assumption was re-evaluated by appropriately providing the event sequence from the result of plant response analysis. As a result, detailed evaluation of the ATWS frequency according to the result of the plant response analysis and reviewing event sequence scenarios have enabled realistic ATWS frequency estimation.
Experimental Study on Structural Integrity of a Core Shroud Support with a Crack Under Seismic Load
Authors: Okita Shunsuke , Saito Koichi , Dozaki Koji , Shima Akihiro
Published: September 2008

This study was performed to experimentally confirm the conservatism of the integrity assessment procedure for a cracked core shroud support of the boiling water reactor (BWR) under seismic load. From the comparison of the experimental and analytical results, it is shown that finite element method (FEM) is accurate and collapse load estimated by twice-elastic slope method is conservative.
Margin Analysis of MONJU Thermal Transient Based on Measured Plant Performance
Authors: Mori Takero , Yamada Fumiaki
Published: September 2008

This paper reports a part of realistic safety margin analysis of Japanese prototype fast breeder reactor MONJU, based on actually measured plant characteristics during a commissioning operation under up to 40% electric power generation mode. In order to evaluate realistic integrity margins against design-phase evaluations and against the limit values, the thermal stresses of critical component parts after a typical thermal transient event, i.e., a manual reactor trip, were derived through serially combined thermal transient and structural analyses. The results of thermal transient and structure analysis confirmed that the evaluation values of the structural stresses at the reactor vessel outlet nozzle, at the upper tube sheet of the intermediate heat exchanger, and at the evaporator feed water inlet tube sheet are well within the design evaluation values.
Applicability Examination and Evaluation of Reactor Dismantlement Technology in the FUGEN: Examination of Double Tubes Cutting by Abrasive Water Jet
Authors: Nakamura Yasuyuki , Kikuchi Kouichi , Morishita Yoshitugu , Usui Tatsuo , Ogane Daisuke
Published: September 2008

As a peculiar issue of the decommissioning of FUGEN, it is necessary to establish a dismantlement method for the reactor having a 224 double-tubes structure arranged with pressure and calandria tubes concentrically. Mechanical cutting method will be desirable considering the influence on the atmospheres because the double-tubes consist of highly activated zirconium alloy and zircalloy material. Therefore, the abrasive water jet method was tested and examined as a mechanical double-tube cutting method that needs the standoff comparatively longer. We confirmed the applicability of the abrasive water jet method to the dismantlement of FUGEN’s reactor.
Analytical Study on Micro-Indentation Method to Integrity Evaluation for Graphite Components in HTGR
Authors: Sumtia Junya , Hanawa Satoshi , Shibata Taiju , Tada Tatsuya , Iyoku Tatsuo , Sawa Kazuhiro
Published: September 2008

An analytical study on micro-indentation method to integrity evaluation for graphite components was carried out. The indentation method is used as simplicity test to measure mechanical properties of materials. This method is thought to be applicable to evaluate the residual stress from the relationship between indentation load and indentation depth. In this study, in order to confirm the applicability of the micro-indentation method for lifetime evaluation of the graphite component, indentation load-depth behavior under stress/strain condition was evaluated taking account of the specified minimum ultimate strength of IG-110 graphite. Moreover, analytical investigations of indentation load-depth behavior for oxidized graphite and oxidized graphite with residual strain were also carried out. As a result, it can be said that the indentation method is potentially applicable to evaluate the integrity of graphite components.
Pressure and Thermal Stress Analyses of a Generation IV Reactor System, Mini-Channel Heat Exchanger
Authors: Yoo Bong , Kulak Ronald F. , Kim Yun-Jae
Published: September 2008

An advanced power conversion system incorporating a Supercritical Carbon Dioxide (S-CO2) Brayton Cycle for Liquid Metal-Cooled Fast Reactors, can provide much improved cycle efficiency relative to a traditional Rankine cycle. Because of this, current plans for Generation IV reactor systems—like the Liquid Metal-Cooled Fast Reactor—include the use of the S-CO2 Brayton Cycle in the development of a power conversion system. However, a structural issue is raised with the use of a mini-channel heat exchanger because large temperature and pressure differences occur along the passages of the mini-channels during normal and transient operating conditions. The design parameters of the heat exchanger during normal operating conditions (i.e., steady state) are pressures and temperatures at the inlet and outlet of the hot and cold channels, and the average heat transfer coefficients within the mini-channels. In this paper, results are presented from preliminary uncoupled thermal and stress analyses of the heat exchanger based on very simple finite element models and the heat exchanger design parameters. Temperature distributions along the passage ways of the mini-channels are calculated. The stresses resulting from both the pressure load and the thermal load are compared with the ASME Section VIII design requirement. The structural integrity of the simplified heat exchanger model—during normal operating conditions of the S-CO2 Brayton Cycle—is evaluated.
Verification of the Plant Dynamics Analytical Code CERES Using the Results of the Plant Trip Test of the Prototype Fast Breeder Reactor MONJU
Authors: Nishi Yoshihisa , Ueda Nobuyuki , Kinoshita Izumi , Miyakawa Akira , Kato Mitsuya
Published: September 2008

CERES is plant system analysis code for LMRs (liquid metal cooled reactors) developed by the Central Research Institute of Electric Power Industry (CRIEPI). To verify the CERES code, analyses were performed by using the result of the plant trip test of the prototype FBR (fast breeder reactor) “MONJU” at 40% rated power. The verification work was performed as a joint research of CRIEPI and JAEA (Japan Atomic Energy Agency). Following three verification analyses were performed mainly. (I) Analysis concerning the primary/ secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) was modeled in R-Z 2-dimension). (II) Analysis concerning the thermal-hydraulic characteristics in the plenum of R/V (the plenum was modeled in 3-dimension). (III) Analysis concerning the flow characteristics inside the intermediate heat exchanger (IHX) (the plenum in the IHX was modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the “MONJU”. Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of “MONJU” became clear by these analyses.
The Development of Loss of Flow Analysis Method for OPR1000 Using RETRAN
Authors: Lee Dong Hyuk , Kim Yo Han , Sung Chang Kyung
Published: September 2008

A new loss of flow transient analysis method for OPR1000 (Optimized Power Reactor 1000, previously called KSNP: Korean Standard Nuclear Power Plant) based on RETRAN code were developed. The reference plant for the analysis is Ulchin Unit 3 and the transient analyzed is 4 pump coastdown. The current analysis for loss of RCS flow transient of OPR1000 uses COAST and CESEC codes. The new method uses RETRAN code to replace COAST and CESEC codes. Since the ability of RETRAN to replace CESEC has been studied in other non-LOCA transients, this paper will focus on COAST code and RCP coastdown flow rates. The results from simplified RETRAN nodalization corresponding to COAST show good agreement with RCS flow results from COAST code. The results are also compared with RETRAN basedeck for safety analysis which is more complex and show similar trends. Therefore, previous analysis method for loss of flow of OPR1000 using COAST code can be replaced with the new analysis method based on RETRAN.
Analytical Evaluation of Unsteady Thermal Hydraulic Characteristics in Shell and Tube Heat Exchangers
Authors: Takahashi Shiro , Narumi Yuichi , Ishihama Kiyoshi , Yokoyama Akihito , Tsuge Toyohiko , Hayashi Masaaki , Takehara Hidetoshi , Yamashita Norimichi
Published: September 2008

Many shell & tube heat exchangers are used in nuclear power plants. Unsteady thermal hydraulic phenomena have been studied in shell & tube heat exchangers to improve their safety and reliability and to extend their lifetime based on experience obtained from long periods of plant operation. We investigated unsteady flow in shell & tube heat exchangers by using computational fluid dynamics (CFD) analyses. The inlet flow on the shell side was separated and flow in several directions. A large part of the flow crossed over the tube bundle, and some parts of the flow took two circuitous roots (up and down) along the inner surface of the shell. Separated circuitous flows collided again where a baffle plate had been cut off. A pair of symmetric vortexes could be seen in that location. Some parts of the circuitous flow moved backwards into the tube bundle due to vortexes. These vortexes were unstable and changed their size and location. A pair of vortexes changed from symmetric to asymmetric. As a result, the direction of flow in the tube bundle near the vortexes changed continuously. Variations in vortexes simulated through CFD analyses could also be seen in tests on the actual size. Fluid temperature fluctuations around tubes were also evaluated through CFD analyses. Unsteady phenomena with changes from symmetric to asymmetric vortexes could be observed in the shell & tube heat exchanger and were simulated through CFD analyses with a detached eddy simulation (DES) turbulence model.
Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 9—System Outline and Endurance Test of Low-Pressure Steam Injectors
Authors: Ohmori Shuichi , Mori Michitsugu , Goto Shoji , Narabayashi Tadashi , Iwaki Chikako , Asanuma Yutaka
Published: September 2008

A Steam Injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to serve also as a direct-contact feedwater heater that heats up feedwater by using extracted steam from the turbine. We are developing technology for “Innovative Simplified Nuclear Power Plants” in order to further improve the economy and safety of nuclear power plants. Our technology development aims to significantly simplify equipment and reduce physical quantities by applying “High-Efficiency SI”, which are applicable to a wide range of operation regimes beyond the performance and applicable range of existing SIs and enables unprecedented multistage and parallel operation, to the low-pressure feedwater heaters and Emergency Core Cooling Systems (ECCS) of nuclear power plants, as well as achieve high inherent safety to prevent severe accidents by keeping the core covered with water (a Severe Accident-Free Concept). The innovative-simplified nuclear power plant consists of a simplified feedwater heating system, a passive core injection system and a passive containment cooling system. This report describes the results of the endurance and performance tests of low-pressure SIs for feedwater heaters with Jet-deaerator and core injection system. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba, and 7 Universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.
Reactor Analysis for Counter-Flowing Moderator and Coolant in a Supercritical Water Reactor
Authors: Wilson Paul P. H. , Hu Po
Published: September 2008

New thermal-hydraulic models have been implemented in the PARCS v2.5 core simulator to enable the simulation of heat transfer between the counter-flowing moderator and coolant in supercritical water reactors. The impact of this heat transfer on the performance of an individual assembly has been demonstrated. A quarter-core model has been simulated to demonstrate how this new capability can be used to identify hot channels in the core and to design different core configurations that may reduce the hot channel effects. Future developments will couple this core simulator capability to advanced thermal-hydraulic simulation capability for full core design and transient simulation.
Development of System Based Code: Methodologies for Life-Cycle Margin Evaluation
Authors: Morishita Masaki , Asayama Tai , Tashimo Masanori
Published: September 2008

The late Professor Emeritus Yasuhide Asada proposed the System Based Code concept, which intends the optimization of design of nuclear plants through margin exchange among a variety of technical options which are not allowed by current codes and standards. The key technology of the System Based Code is margin exchange evaluation methodology. This paper describes recent progress with regards to margin exchange methodologies in Japan.
Transient Pool Boiling Critical Heat Flux of FC-72 Under Saturated Conditions
Authors: Fitri Sutopo P. , Fukuda Katsuya , Liu Qiusheng , Park Jongdoc
Published: September 2008

In this study, the steady-state and transient critical heat fluxes (CHFs) in pool boiling were measured on 1.0 mm diameter horizontal cylinders of gold and platinum heaters under saturated conditions due to transient heat inputs, Q0exp(t/τ), in a pool of Fluorinert FC-72. Heaters were heated by electric current with the periods, τ, ranged from 10 ms to 20 s, and the pressures ranged from atmospheric up to around 1.2 MPa. The steady-state CHFs measured are dependent on pressure and almost agree with the values obtained by Kutateladze’s correlation based on hydrodynamic instability (HI) model. It was considered that the boiling inception and the direct transition during the steady-state period occur by the pre-pressure of ∼1.2 MPa. The trend of typical transient CHFs were clearly divided into the first, second, and third groups for long, short, and intermediate periods, respectively. The direct transition processes to film boiling without nucleate boiling for the short periods obtained from both heaters were confirmed due to the heterogeneous spontaneous nucleation (HSN) in flooded cavities on the cylinder surface. The empirical correlations to express each of corresponding CHFs measured on both heaters for the short periods are presented in this paper.
Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States
Authors: Permana Sidik , Takaki Naoyuki , Sekimoto Hiroshi
Published: September 2008

Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233U which shows superior characteristics in the thermal region. The Shippingport reactor used 233U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233U enrichment is about 2%–9% as charge fuel. The lower MFR and the higher enrichment of 233U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient.