bluuebunny/crossref_metadata_2025 huggingface dataset:
Subchannel Analysis with Mechanistic Methods for Thermo-Hydro Dynamics in BWR Fuel Bundles
Authors: Imura Kentaro , Yoshida Kenji , Kataoka Isao , Naitoh Masanori
Published: September 2008

In order to predict the critical power or void fraction in BWR fuel bundles and the DNB heat flux of PWR fuel assemblies, the boiling transition analysis code called “CAPE” with mechanistic models has been developed in the IMPACT project by NUPEC. The objective of the CAPE code development is to perform with good accuracy the safety evaluation for a new type or improved fuel bundle design of BWR and PWR without full-scale experiments or any tuning parameters in the analysis code. In this study, the CAPE for BWR was validated by the test analysis for 8 × 8 fuel bundles comparing with the void distribution data of the experimental data, which was carried out under several operational conditions in a BWR. The computations were carried out by changing the operational parameter such as the inlet subcooling, mass flow rate and the power output of the fuel bundles. Resultantly, the thermal equilibrium quality at the outlet ranges from 2% to 25%. From these results, though the predictive accuracy of the analytical results are in close agreement with the experimental data, it was noted that the errors were relatively outstanding in some subchannels, which was surrounded by the heated fuel rods and partially unheated walls, such as an unheated rod, a water rod and a separation wall of the channel box. The reason for this error is thought to be that the cross sectional void distribution was partially distributed in such subchannels surrounded partially by unheated wall, so the multidimensional void distribution structure might be formed in these subchannels. Under such conditions, it is very important to take into consideration the multidimensional structure of the two-phase flow in subchannel, and perhaps improve the estimation or correlations for the distribution parameter, as well as the amount of void exchange between neighboring subchannels.
Identification of CSSC Caused by Ageing and Degradation
Authors: Florescu Gheorghe , Cojan Mihail
Published: September 2008

PSA studies, that were developed for some NPPs, permit the using of the created models to perform many research tests, in order to optimize the structures, systems and components (SSCs) operation or to identify the NPP or systems weaknesses, due to specific or special factors. SSCs that influence decisively the NPP reliability are considered as critical. Also, for the accident conditions, the SSC, which have a major influence to the system availability or operability, are considered as critical. Many worldwide NPPs reached the life time or are very close to do that. Several SSCs have shorter life times than NPP’s life time. Ageing is one of the factors that decrease the SSC life time. Due to ageing, if are not replaced, some SSCs, or groups of redundant SSCs, become critical looking to safety. Some questions for what to do in the situation when a SSC must be replaced and the SSC specific manufacturer doesn’t exist, could also be put. The paper tried to solve the problem of SSC modeling by introducing of an ageing factor in SSC model. Fault tree (F/T) modeling approach is assumed. There are two possibilities for modeling: failure rates that are changed or specific MCS (minimal cut set) term modified by ageing. Risk analysis and PSA techniques are used as a basis for analysis. The paper includes: the steps to establish the systems or components that suffer ageing; methods to identify CSSC taking into account ageing; the events associated to ageing/degradation and presentation of method to determine the ageing related events, selection of the SSCs that are important for analyses; selection of the most significant ageing events; ranking of ageing events; association of events to these components in order to decide for the CSSC detailed analyses; ranking / ordering of the ageing related events; optimization of NPP systems design and operation considering ageing; impact of ageing to NPP operation/safety/safety margins and to manufacturer technical specifications. The paper presents a brief description of the most important aspects of the methods, used to analyze the ageing effects on appearing of CSSCs, taking into account the previous developed NPP PSA models and PSA modeling tools.
A 3D BWR Core Model Within the GRS Analysis Simulator ATLAS
Authors: Velkov K. , Langenbuch S. , Lerchl G. , Pointner W.
Published: September 2008

The paper describes the application of the ATLAS simulator environment and the coupled three dimensional (3D) neutron-kinetics and thermal-hydraulics system code QUABOX/CUBBOX-ATHLET for a boiling water reactor (BWR) plant transient. A turbine trip (TT) transient is simulated and analyzed once with the 3D core model and once with the point kinetics (PK) model using data generated on the basis of the 3D calculations by the kinetics data generation system SIGMAS developed in GRS. The comparison shows a very good agreement, which is an important precondition for performing transient analyses with an on-line switch from PK to 3D calculation for a BWR plant transient within the ATLAS simulator.
Development of Crystallizer for Advanced Aqueous Reprocessing Process
Authors: Washiya Tadahiro , Kikuchi Toshiaki , Shibata Atsuhiro , Chikazawa Takahiro , Homma Shunji
Published: September 2008

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. [1, 2, 3] In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described.
Development of Water Radiolysis Code for the JMTR IASCC Test Loop
Authors: Hanawa Satoshi , Sato Tomonori , Mori Yuichiro , Oogiyanagi Jin , Kaji Yoshiyuki , Uchida Shunsuke
Published: September 2008

In order to evaluate the water chemistry in the irradiation field during IASCC irradiation test, a water radiolysis code for IASCC irradiation loop system was developed. In the water radiolysis code, a multiple node model was introduced since the irradiation loop system has a wide rage temperature distribution as well as the dose distribution. To investigate the applicability of developed water radiolysis code, water chemistry at the water sampling point of the irradiation loop system was measured and compared with analytical results under several water chemistry conditions. Further, water chemistry distribution in the in-pile region as well as in the out-pile region was calculated by the developed water radiolysis code.
Development of Residual Stress Improvement for Nuclear Pressure Vessel Instrumentation Nozzle Weld Joint (P-43+P-8) by Means of Induction Heating
Authors: Terajima Takuro , Hirano Takashi
Published: September 2008

As a counter measurement of intergranular stress corrosion cracking (IGSCC) in boiling water reactors, the induction heating stress improvement (IHSI) has been developed as a method to improve the stress factor, especially residual stresses in affected areas of pipe joint welds. In this method, a pipe is heated from the outside by an induction coil and cooled from the inside with water simultaneously. By thermal stresses to produce a temperature differential between the inner and outer pipe surfaces, the residual stress inside the pipe is improved compression. IHSI had been applied to weld joints of austenitic stainless steel pipes (P-8+P-8). However IHSI had not been applied to weld joints of nickel-chromium-iron alloy (P-43) and austenitic stainless steel (P-8). This weld joint (P-43+P-8) is used for instrumentation nozzles in nuclear power plants’ reactor pressure vessels. Therefore for the purpose of applying IHSI to this one, we studied the following. i) Investigation of IHSI conditions (Essential Variables); ii) Residual stresses after IHSI; iii) Mechanical properties after IHSI. This paper explains that IHSI is sufficiently effective in improvement of the residual stresses for this weld joint (P-43+P-8), and that IHSI does not cause negative effects by results of mechanical properties, and IHSI is verified concerning applying it to this kind of weld joint.
Design Study of Small Lead-Cooled Fast Reactors Using SiC Cladding and Structure
Authors: Rivai Abu Khalid , Takahashi Minoru
Published: September 2008

Effects of SiC cladding and structure on neutronics of reactor core for small lead-cooled fast reactors have been investigated analytically. The fuel of this reactor was uranium nitride with 235U enrichment of 11% in inner core and 13% in outer core. The reactors were designed by optimizing the use of natural uranium blanket and nitride fuel to prolong the fuel cycle. The fuels can be used without reshuffling for 15 years. The coolant of this reactor was lead. A calculation was also conducted for steel cladding and structure type as comparison with SiC cladding and structure type. The results of calculation indicated that the neutron energy spectrum of the core using SiC was slightly softer than that using steel. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have critical condition with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The result of the design analysis showed that neutron flux distributions and power distribution was made flatter because the outer core enrichment was higher than inner core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 13% for 125 MWt reactor and 25% for 375 MWt reactor at EOC.
Numerical Simulation of Two-Phase Flow in Severely Damaged Core Geometries
Authors: Meekunnasombat Phongsan , Fichot Florian , Quintard Michel
Published: September 2008

In the event of a severe accident in a nuclear reactor, the oxidation, dissolution and collapse of fuel rods is likely to change dramatically the geometry of the core. A large part of the core would be damaged and would look like porous medium made of randomly distributed pellet fragments, broken claddings and relocated melts. Such a complex medium must be cooled in order to stop the accident progression. IRSN investigates the effectiveness of the water re-flooding mechanism in cooling this medium where complex two-phase flows are likely to exist. A macroscopic model for the prediction of the cooling sequence was developed for the ICARE/CATHARE code (IRSN mechanistic code for severe accidents). It still needs to be improved and assessed. It appears that a better understanding of the flow at the pore scale is necessary. As a result, a direct numerical simulation (DNS) code was developed to investigate the local features of a two-phase flow in complex geometries. In this paper, the Cahn-Hilliard model is used to simulate flows of two immiscible fluids in geometries representing a damaged core. These geometries are synthesized from experimental tomography images (PHEBUS-FP project) in order to study the effects of each degradation feature, such as displacement and fragmentation of the fuel rods and claddings, on the two-phase flow. For example, the presence of fragmented fuel claddings is likely to enhance the trapping of the residual phase (either steam or water) within the medium which leads to less flow fluctuations in the other phase. Such features are clearly shown by DNS calculations. From a series of calculations where the geometry of the porous medium is changed, conclusions are drawn for the impact of rods damage level on the characteristics of two-phase flow in the core.
Erosion Evaluations of a Slurry Mixer Tank With Computational Fluid Dynamics Methods
Authors: Lee Si Y. , Dimenna Richard A. , Taylor Glenn A.
Published: September 2008

This paper discusses the use of computational fluid dynamics (CFD) methods to understand and characterize erosion of the floor and internal structures in the slurry mixing vessels in the Defense Waste Processing Facility. An initial literature survey helped identify the principal drivers of erosion for a solids laden fluid: the solids content of the working fluid, the regions of recirculation and particle impact with the walls, and the regions of high wall shear. A series of CFD analyses was performed to characterize slurry-flow profiles, wall shear, and particle impingement distributions in key components such as coil restraints and the vessel floor. The calculations showed that the primary locations of high erosion resulting from abrasion were at the leading edge of the coil guide, the tank floor below the insert plate of the coil guide support, and the upstream lead-in plate. These modeling results based on the calculated high shear regions were in excellent agreement with the observed erosion sites in both location and the degree of erosion. Loss of the leading edge of the coil guide due to the erosion damage during the slurry mixing operation did not affect the erosion patterns on the tank floor. Calculations for a lower impeller speed showed similar erosion patterns but significantly reduced wall shear stresses.
Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors
Authors: Carlsson Johan , Tucek Kamil , Wider Hartmut
Published: September 2008

This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MWth power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MWth critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinde fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations.
Design Status and Applications of Small Reactors Without On-Site Refuelling
Authors: Kuznetsov Vladimir
Published: September 2008

Small reactors without on-site refuelling (SRWORs) are the reactors that can operate without reloading and shuffling of fuel for a reasonably long period with no refuelling equipment being present in the reactor and no fuel being stored at the site during reactor operation. By virtue of being small, transportable and requiring no operations with fuel from a customer, such reactors form an attractive domain for fuel or even NPP leasing. SRWORs could simplify the implementation of safeguards and provide certain guarantees of sovereignty to those countries that would agree to forego the development of the indigenous fuel cycle. About 30 concepts of such reactors are being analyzed or developed in 6 IAEA member states. Based on intermediate results of IAEA activities in support of the design and technology development for such reactors, the paper provides technical details on the design status, fuel cycle options and possible applications of SRWORs.
Experience With Damaged Spent Nuclear Fuel at U.S. DOE Facilities
Authors: Carlsen Brett , Fillmore Denzel , McCormack Roger L. , Sindelar Robert , Spieker Timothy , Woolstenhulme Eric
Published: September 2008

This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report.
Comparison of CFD Model Velocity Profiles to Test Data for Inlet Nozzle Region of Nuclear PWR Fuel Assemblies
Authors: Karoutas Zeses E. , Liu Bin , Dzodzo Milorad , Joffre Paul F.
Published: September 2008

A numerical investigation was performed to study the variation in axial velocity profiles occurring downstream of the inlet nozzle region of Nuclear PWR fuel assemblies. Computational Fluid Dynamic (CFD) models were prepared for the inlet nozzle region of a section of fuel assembly, simulating the lower support plate located under the fuel assembly, the inlet nozzle of the fuel and the downstream fuel region. Two different nozzle designs were modeled to study how each nozzle impacts the dissipation of the jet velocity profiles occurring downstream of the nozzle. The two different nozzle designs included a standard round chamfered hole flow plate and a chamfered slotted flow plate. The evaluation of the axial velocity profiles occurring downstream of the nozzle flow plate is critical in understanding the fuel rod vibration and rod fretting in the first grid span. Excessive rod vibration in this region can occur due to high axial jet velocities and steep axial velocity gradients generated from the holes in the lower support plate. The excessive rod vibration can lead to fuel rod wear and fuel failure. Axial velocity profiles were predicted for the different nozzle designs using the CFX code. These velocity profiles were compared to air test velocity measurements for the same nozzle designs. Velocity measurements were made in a 3.763/1 over-scale air test section simulating a 6×6 rod array of the inlet nozzle region and downstream fuel region. Reasonable agreement was observed between the velocity measurements and CFD model predictions. The results also indicate that nozzle flow plate geometry can have a significant affect on the dissipation of the jet axial velocity profiles and the steepness of the axial velocity gradients downstream on the inlet nozzle. The application of CFD tools can be used to optimize the inlet nozzle geometry to better dissipate jets and reduce axial velocity gradients downstream of the nozzle at a minimal increase in pressure drop. This will help reduce fuel rod vibration and rod fretting.
CFD Application to Flow-Accelerated Corrosion in Feeder Bends
Authors: Pietralik John M. , Smith Bruce A. W.
Published: September 2008

Feeder piping in CANDU® plants experiences a thinning degradation mechanism called Flow-Accelerated Corrosion (FAC). The piping is made of carbon steel and has high water flow speeds. Although the water chemistry is highly alkaline with room-temperature pH in a range of 10.0–10.5, the piping has FAC rates exceeding 0.1 mm/year in some locations, e.g., in bends. One of the most important parameters affecting the FAC rate is the mass transfer coefficient for convective mass transport of ferrous ions. The ions are created at the pipe wall as a result of corrosion, diffuse through the oxide layer, and are transported from the oxide-layer/water interface to the bulk water by mass transport. Consequently, the local flow characteristics contribute to the highly turbulent convective mass transfer. Plant data and laboratory experiments indicate that the mass transfer step dominates FAC under feeder conditions. In this study, the flow and mass transfer in a feeder bend under operating conditions were simulated using the Fluent™ computer code. Because the flow speed is very high, with the Reynolds numbers in a range of several millions, and because the geometry is complex, experiments in a 1:1 scale were conducted with the main objective to validate flow simulations. The experiments measured pressure at several key locations and visualized the flow. The flow and mass transfer models were validated using available friction-factor and mass transfer correlations and literature experiments on mass transfer in a bend. The validation showed that the turbulence model that best predicts the experiments is the realizable k-ε model. Other two-equation turbulence models, as well as one-equation models and Reynolds stress models were tried. The near-wall treatment used the non-equilibrium wall functions. The wall functions were modified for surface roughness when necessary. A comparison of the local mass transfer coefficient with measured FAC rate in plant specimens shows very good agreement. Visualization experiments indicate secondary flows in the bends. No boundary layer separation was observed in experiments or in simulations.
Advanced Pipe Replacement Procedure for a Defective CRDM Housing Nozzle Enables Continued Normal Operation of a Nuclear Power Plant
Authors: Gilmore Geoff , Becker Andrew
Published: September 2008

During the 2003 outage at the Ringhals Nuclear Plant in Sweden, a leak was found in the vicinity of a Control Rod Drive Mechanism (CRDM) housing nozzle at Unit 1. Based on the ALARA principle for radioactive contamination, a unique repair process was developed. The repair system includes utilization of custom, remotely controlled GTAW-robots, a CNC cutting and finishing machine, snake-arm robots and NDE equipment. The success of the repair solution was based on performing the machining and welding operations from the inside of the SCRAM pipe through the CRDM housing since accessibility from the outside was extremely limited. Before the actual pipe replacement procedure was performed, comprehensive training programs were conducted. Training was followed by certification of equipment, staff and procedures during qualification tests in a full scale mock-up of the housing nozzle. Due to the ingenuity of the overall repair solution and training programs, the actual pipe replacement procedure was completed in less than half the anticipated time. As a result of the successful pipe replacement, the nuclear power plant was returned to normal operation.
In-Pile SCC Growth Behavior of Type 304 Stainless Steel in High Temperature Water at JMTR
Authors: Kaji Yoshiyuki , Ugachi Hirokazu , Tsukada Takashi , Matsui Yoshinori , Ohmi Masao , Nagata Nobuaki , Dozaki Koji , Takiguchi Hideki
Published: September 2008

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing under a framework of cooperative research program between JAERI and the JAPC. In-pile IASCC growth tests have been successfully carried out using the compact tension (CT) type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1×1025n/m2 before the in-pile testing since 2004. The tests were carried out in pure water simulated boiling water reactor (BWR) coolant condition. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC.
The Production of Radioisotope (Zn-65) with Miniature Neutron Source Reactor
Authors: Rezvanifard Mehdi , Shahabi Iraj , Ebadati Javad
Published: September 2008

MNSR Reactor is a tank-pool type research reactor.
Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 11—Visualization Study on the Start-Up of the Steam Injector
Authors: Okamoto Koji , Narabayashi Tadashi , Iwaki Chikako , Ohmori Shuichi , Mori Michitsugu
Published: September 2008

The Steam Injector is the superior system to pump the fluid without rotating machine. Because the water column is surrounded by the saturated steam, very high heat transfer is also expected with direct condensation. The inside of the Steam Injector is very complicated. To improve the efficiency of the Steam Injector, the water column behavior inside the Injector is visualized using the Dynamic PIV system. Dynamic PIV system consists of the high-speed camera and lasers. In this study, 384×180 pixel resolution with 30,000fps camera is used to visualize the flow. For the illumination CW green laser with 300mW is applied. To view inside the Injector, relay lens system is set at the Injector wall. Very high speed water column during the starting up of Steam Injector had been clearly visualized with 30,000fps. The wave velocity on the water column had been analyzed using PIV technique. The instability of the water column is also detected.
Results of the QUENCH-L2, DISCO-L2, and COMET-L2 Experiments Performed Within the LACOMERA Project at the Forschungszentrum Karlsruhe
Authors: Miassoedov Alexei , Alsmeyer Hans , Meyer Leonhard , Steinbrueck Martin , Groudev Pavlin , Ivanov Ivan , Sdouz Gert
Published: September 2008

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: Fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.
SCC in Shroud Support Cylinder Vertical Weld Lines (Tokai-2)
Authors: Arita Yuichi , Dozaki Koji , Manabe Fumio , Kanno Satoshi
Published: September 2008

SCC was found outer surface of shroud support cylinder vertical weld lines (V8) made of Nickel based alloy, alloy 182, for Tokai-2 (BWR-5) operated since 1978. Three SCC among 4 weld lines of V8 were observed. The material of Alloy182 was known to have SCC potential in BWR environment. Based on the result of finite element method analysis, it was estimated that tensile circumferential stress generated at the upper corner of cylinder corresponding crack location when H7 welded. The integrity assessment against seismic load in design was performed using shell model in finite element method analysis. It assumed conservatively that four vertical through wall cracks along whole length of V8 weld lines and four horizontal through wall partial cracks along H7 weld line at intersections of V8 and H7. The collapse load was estimated by twice slope method, a kind of limit load analysis. As the result of the integrity assessment, a critical horizontal through wall crack length along H7 weld line was about 1400 mm per one crack (about 70% of all circumferences). SCC growth was evaluated to reach the critical length after about 20 years, where maximum crack growth rate, 63mm/year was assumed. It is not judged that the immediate repair of vertical SCC is necessary.
Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)
Authors: Tamura Masataka , Kawano Shohei , Kouno Wataru , Kanazawa Yasushi
Published: September 2008

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report.
Methodology for Determining the Radiological Status of a Process: Application to Decommissioning of a Fuel Reprocessing Facility
Authors: Girone`s Ph. , Ducros C. , Legoaller C. , Lamadie F. , Fulconis J.-M. , Thiebaut V. , Mahe´ C.
Published: September 2008

Decommissioning a nuclear facility is subject to various constraints including regulatory safety requirements, but also the obligation to limit the waste volume [1] and toxicity. To meet these requirements the activity level in each component must be known at each stage of decommissioning, from the preliminary studies to the final release of the premises. This document describes a set of methods used to determine the radiological state of a spent fuel reprocessing plant. This approach begins with a bibliographical survey covering the nature of the chemical processes, the operational phases, and the radiological assessments during the plant operating period. In this phase it is also very important to analyze incidents and waste management practices. All available media should be examined, including photos and videos which can provide valuable data and must not be disregarded. At the end of this phase, any items requiring verification or additional data are reviewed to define further investigations. Although it is not unusual at this point to carry out an additional bibliographical survey, the essential task is to carry out in situ measurements. The second phase thus consists in performing in situ measurement campaigns involving essentially components containing significant activity levels. The most routinely used methods combine the results of elementary measurements such as the dose rate [2] [3] or more sophisticated measurements such as gamma spectrometry using CdZnTe detectors [4] and gamma imaging [2] to estimate and localize the radioactivity. Each instrument provides part of the answer (location of a contamination hot spot, standard spectrum, activity). The results are combined and verified through the use of calculation codes: Mercure [5], Visiplan [6] and Microshield [7].
Modeling and Simulation of Cross-Flow Induced Vibrations in Tube Bundles Using Bondgraph Approach
Authors: Malik M. Afzaal , Rashid Badar , Khan M. Anwar , Bashir Khawaja Sajid , Khushnood Shahab
Published: September 2008

A considerable research has been carried out in the field of Cross-Flow Induced Vibrations (CFIV) in tube bundles of process exchangers and nuclear steam generators. Various excitation mechanisms such as vortex shedding, turbulent buffeting, fluid-elastic instability and acoustic resonance and other parameters like natural frequencies, damping, wear work rates at the loose tube supports and various geometric tube arrangements have been the focus in single and two-phase cross-flow. In the current research work, CFIV has been studied by using Bondgraph approach. The Bondgraph models have been subjected to simulation using the software (20-SIM). Results obtained have shown a strong usefulness of Bondgraph approach to complex CFIV systems.
Design Study of Mechanical Disassembly System for FBR Fuel Reprocessing
Authors: Tooya Yuuichi , Washiya Tadahiro , Koizumi Kenji , Morita Shinichi
Published: September 2008

Japan Atomic Energy Agency (JAEA) has been leading feasibility study on commercialized fast reactor cycle systems in Japan. In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. In the mechanical disassembly system, high durability mechanical tool grinds the wrapper tube (Slit-cut (S/C) operation in circle direction), and then the wrapper tube is pulled out and removed from the fuel assembly. Then the fuel pins are cut (Crop-cut (C/C) operation at entrance nozzle side) and the entrance nozzle is removed. The fuel pins are transported to the shearing device in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As results, we established an efficient disassembly procedure and the fundamental design of mechanical disassembly system.
The Design and Fabrication of the Lead Corrosion Test Loop
Authors: Cho Chungho , Lee Woonkyu , Cho Choonho , Kim Jongman , Song Tae Yung
Published: September 2008

Lead corrosion test loop named KPAL-II (KAERI Pb Alloy Loop II) has been designed and fabricated at the Korea Atomic Energy Research Institute (KAERI) and an initial operation was performed recently. The KPAL-II was designed to study the long-term corrosive effects of liquid lead on structural and fuel cladding materials at temperatures up to 600°C. The first run of KPAL-II took place on March 15th, 2006 and three hundred hours of shakedown testing were performed with an isothermal condition at about 450°C. This paper summarizes the general design concept and the results of the shakedown testing of KPAL-II.
Development of 3-D Simulation Analysis System for PWR Control Rod Drive Mechanism
Authors: Tanaka Akio , Futahashi Kensuke , Takanabe Kiyoshi , Kurimura Chikara , Kato Jungo , Hara Hidekiyo
Published: September 2008

3-D virtual analysis system to analyze the motion of Control Rod Drive Mechanism (CRDM) was developed. The analysis system consists of 3-D model established per the actual dimensions and interfaces of CRDM parts and a routine to calculate forces acting on the mechanism, and was verified by mock-up test using equipment same to actual product. The analysis system is useful for functional evaluation in maintenance or to factor out root causes in case of malfunction of CRDM.
Natural Circulation Characteristics of an Integral Pressurized Water Reactor
Authors: Gou Junli , Qiu Suizheng , Su Guanghui , Jia Dounan
Published: September 2008

Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation.
Development of an Advanced Flow Meter Applicable to Large Diameter Piping System Using Average Bi-Directional Flow Tube
Authors: Kang Kyoung-Ho , Yun Byong-Jo , Euh Dong-Jin , Baek Won-Pil
Published: September 2008

A new instrument, an average BDFT (birectional flow tube), is proposeded to measure the area-averaging single- and two-phase flow rates. This average BDFT is capable of being applicable to the low or bidirectional flow condition. In this study, an advanced flow meter applicable to large diameter piping system was developed using the average BDFT. This flow meter can be used to measure the flow rate of discharged gas having fine floating particles in the stack of a chemical reactor or a large-sized boiler. Also it can be applied to the large-sized industrial piping system having flow of oil without structural instability. As a first step of commercialization, we designed and manufactured the flow meters which are applicable to piping system having diameter of 200mm and 500mm. The calibration test results indicated that this flow meter has linaerity within 0.5% in the Re number range of above 10,000 where is meaningful in terms of application. The flow analyses results using CFX code demonstrate that the flow meter has a good linearity with various design features. Therefore, it is quite reasonable to select the typical design of the average BDFT for a convenience of the experimental conditions and pipe arrangements. The structural analyses results imply that this flow meter is applicable to the various flow rate conditions.
CFD Analysis of the Active Part of the HYPER Spallation Target
Authors: Tak Nam-il , Cho Chungho , Song Tae-Yung
Published: September 2008

KAERI (Korea Atomic Energy Research Institute) is developing an accelerator driven system (ADS) named HYPER (HYbrid Power Extraction Reactor) for a transmutation of long-lived nuclear wastes. One of the challenging tasks for the HYPER system is to design a large spallation target having a beam power of 15∼25 MW. The present paper focuses on the thermal-hydraulic performance of the active part of the HYPER target. Computational fluid dynamics (CFD) analysis was performed using a commercial code CFX 5.7.1. Several advanced turbulence models with different grid structures were applied. The CFX results show the significant impact of the turbulence model on the window temperature. It is concluded that experimental verifications are very important for the design of the HYPER target.
Rewetting of Vertical Hot Surface in a Centrally Heated Annulus and a 6×6 Rod Bundle Geometry During Reflood Phase
Authors: Cho Seok , Moon Sang-Ki , Choi Ki-Yong , Chun Se-Young , Chung Moon-Ki , Baek Won-Pil
Published: September 2008

A series of bottom reflood tests were carried out to investigate thermal-hydraulic characteristics during the reflood phase. This paper includes descriptions of three related groups of reflood tests categorized by the geometry of a flow channel and an electric power shape of heater rods. A centrally-heated annular geometry with an outer-visualizing tube was adopted for the first two groups of tests (group-A and -B), and a 6×6 rod bundle geometry for the other group of tests (group-C). The ranges of experimental parameters are 2∼8 cm/s of flooding velocity, 20∼80 °C of inlet subcooling temperature, and 500∼700 °C of initial wall temperature. In the single rod annular flow channel reflood test, quench front behavior can be easily observed through the visualizing window and a dominant flow regime near downstream of quench front is inverted annular film boiling regardless of the flooding velocity. For the case of the 6×6 rod bundle experiments, on the other hand, the dominant flow regime is dispersed flow film boiling (DFFB), and therefore the thermal hydraulic behavior becomes more complicated and chaotic due to the interaction between liquid phase such as droplet and liquid film and vapor phase generated from liquid-wall heat transfer.
Assessment of Thermal and Hydrodynamic Fragmentation in Molten Fuel Coolant Interaction With Simulant System
Authors: Narayanan K. S. , Das S. K. , Jasmin Sudha A. , Rao E. V. H. M. , Lydia G. , Murthy S. S. , Kumaresan M. , Harvey J. , Kasinathan N. , Rajan M.
Published: September 2008

In the Safety analysis of Fast Breeder Reactor, assessment of Molten Fuel Coolant Interaction (MFCI) assumes importance for two aspects, namely the characterization of the debris and severity of pressure pulses generation. An attempt has been made to investigate the debris generation characteristics with molten Woods Metal (Alloy of Bi 50% Pb 25% Sn 12.5% & Cd 12.5% & melting point of 346 K) - Water simulant system. Liquid Woods metal and liquid Uranium dioxide physical properties (Density, Surface tension & Kinematic viscosity) are similar. Experimental studies were conducted for various melt temperatures covering non-boiling, convective boiling and film boiling regimes of water, to assess the debris generation resulting from both hydrodynamic and thermal interaction. Woods metal was heated to the desired temperature and poured through a hot funnel having a nozzle of 8 mm release diameter into a water column of height up to 140 cm. Experiments were repeated for different coolant temperature and melt inventory up to 5 kg. The melt entry velocity was determined from video recordings. The debris is analyzed on the basis of interface temperature, Rayleigh-Taylor and Kelvin-Helmholtz instabilities. It is observed that Kelvin-Helmholtz instability is the dominant fragmentation phenomena. Contribution due to coolant boiling resulted in more debris generation in the size less than 4 mm.
Improvement of Predictive Accuracy on Subchannel Analysis Code (NASCA) for Tight-Lattice Rod Bundle Tests: Optimization of UEDA’S Entrainment Model Parameter and Cross Flow Model Parameters
Authors: Chitose Hiromasa , Hotta Akitoshi , Ohnuki Akira , Fujimura Ken
Published: September 2008

The Reduced-Moderation Water Reactor (RMWR) is being developed at Japan Atomic Energy Agency and demonstration of the core heat removal performance is one of the most important issues. However, operation of the full-scale bundle experiment is difficult technically because the fuel rod bundle size is larger, which consumes huge electricity. Hence, it is expected to develop an analysis code for simulating RMWR core thermal-hydraulic performance with high accuracy. Subchannel analysis is the most powerful technique to resolve the problem. A subchannel analysis code NASCA (Nuclear-reactor Advanced Sub-Channel Analysis code) has been developed to improve capabilities of analyzing transient two-phase flow phenomena, boiling transition (BT) and post BT, and the NASCA code is applicable on the thermal-hydraulic analysis for the current BWR fuel. In the present study, the prediction accuracy of the NASCA code has been investigated using the reduced-scale rod bundle test data, and its applicability on the RMWR has been improved by optimizing the mechanistic constitutive models.
An Analytical Method to Predict Fretting-Wear Damage in the Double 90° U-Bend Tubes
Authors: Choi Seog-Nam , Yoon Ki-Seok , Choi Taek-Sang
Published: September 2008

When the flow velocity exceeds a critical value, fluid-elastic instability is believed to be a factor of the large-amplitude vibration and results in the rapid wear of heat exchanger tubes. For sub-critical flow velocities, however, the random turbulence excitation is considered the main mechanism to cause the long-term wear on steam generator tubes. Since flow-induced interactions between the tube and tube supports in the sub-critical flow velocity can cause a localized tube wear, tube movement in the clearance between the tube and tube support, as well as the normal contact force on the tubes by fluid should be maintained as low as possible. An analytical method is proposed for predicting fretting-wear damage of the double 90° U-bend tubes. This approach is based on Archard’s equation for adhesive wear with both the straight single-span tube analytical model proposed by Connors and the linear structural dynamic theory of ASME Section III, Appendix N-1300. The results from the presented method show good trend compared with the field data. It is expected that this method can be utilized to predict the fretting-wear of the double 90° U-bend tubes in steam generators.
Passive Safety Small Reactor for Distributed Energy Supply: Heavy Water Mixing Core
Authors: Sawada Ken-Ichi , Odano Naoteru , Ishida Toshihisa
Published: September 2008

The purpose of this paper is to study the core performance of passive safety small reactor for distributed energy supply by changing the heavy water (D2O) concentration in the mixed coolant together with the fuel pitch. The long core life with conditions of the excessive reactivity of 2 %Δk/k, the reactivity shutdown margin of 1 %Δk/k and the negative coolant temperature reactivity coefficient is attained for the case of D2O concentration of 60 % with 10 % enrichment gadolinia (Gd2O3) doped fuel rods. This D2O core has a shorter core life 4.14 years than the original light water (H2O) core 4.76 years, while it needs a larger core size. However, changing the D2O concentration on the way during the burn-up shows a possibility of extending more the core life than that of the original H2O core.