bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of the Optimized Fatigue Analysis Procedure to Evaluate Design Life for APR1400
Authors: Lee Jaegon , Kim Taesoon , Moon Chankook , Lee Kwanghan
Published: July 2009

Fatigue is one of the most important failure mechanisms to assess integrity and design life of nuclear power plants. Fatigue analysis procedure and the standard fatigue design curve (S-N curve) for the class 1 components are given in ASME code section III NB. However, the existing ASME fatigue design curve does not address the effects of light water reactor coolant environment. The life time of ALWRs is designed for 60 years, and recently the plant life time of currently operating NPPs has been extended 20 years more. If we assess the integrity and design life of major components by fatigue analysis considering environmental factor and S-N curve, the estimated fatigue usage factor will not meet the criterion. In this study, detailed fatigue analysis using three dimensional models were performed to develop the optimized fatigue analysis procedure and their results were compared with other references. The locations considered are the pressurizer surge line, the CVCS charging inlet nozzle and the steam generator economizer nozzle of the advanced power reactor 1400 (APR1400).
CFD Modelling of Liquid Poison Injection in ACR-1000 Shutdown System
Authors: Song F. , Noghrehkar R. , Hau K. F.
Published: July 2009

The Liquid Injection Shutdown System (LISS) is a safety system in the CANDU™-type reactor to provide rapid reactor shutdown by automatically injecting a neutron absorbing liquid (“poison”), via injection nozzles, into the moderator in the calandria. The poison distribution in the moderator plays a significant role in the shutdown performance. The Advanced CANDU Reactor (ACR-1000™) is a Generation III+ type reactor as an evolutionary extension of the proven CANDU-6 reactor. The basic design concept of the CANDU-6 LISS is adopted for the ACR-1000. The injection nozzle design has been modified to suit the ACR-1000 reactor core configuration. In this study, a Computational Fluid Dynamics (CFD) model was developed using the ANSYS-CFX software to examine the poison injection characteristics in the ACR-1000 design. The effects of calandria tubes on the poison jet growth and poison distribution in the reactor core were discussed.
Development of an Evaluation Tool of Radiation Dose Rate in Primary Containment Vessel for Hamaoka BWR Plants
Authors: Inagaki Hiromitsu
Published: July 2009

We are developing a radiation dose rate evaluation tool for work areas in the primary containment vessel (PCV) of the Hamaoka BWR plant in order to reduce radiation exposure to workers. The evaluation tool is composed of two calculation codes; a water chemistry code and a radiation dose rate code. The radiation dose rate calculates radiation maps of the work areas in the PCV of Hamaoka plant. We compared measured values with calculated values in order to improve the accuracy of the tool.
Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor
Authors: Sta˚lek Mathias , Ba´na´ti Jo´zsef , Demazie`re Christophe
Published: July 2009

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.
MEFISTO: A Mechanistic Tool for the Prediction of Critical Power in BWR Fuel Assemblies
Authors: Adamsson Carl , Le Corre Jean-Marie
Published: July 2009

Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. To achieve this goal, the code resolves the multi-film mass balance equations at the sub-channel level within the annular flow region while relying on a simple two-field (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide detailed solution to the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to standard three-field subchannel analysis. Models for the numerical computation of the sub-channel multi-film flowrate distributions, including the treatment of cross-flows, part-length rods and spacers grids are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a two-field sub-channel driver code (based on a fast and robust four-equation model). The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle reaches the dryout criteria. Dryout predicted powers (including trends with flow, pressure, inlet subcooling and power distribution) and predicted dryout locations are compared to experimental results, using a large Westinghouse SVEA-96 dryout database, and are shown to yield excellent results.
Experimental Verification of the Behavior of Underclad Cracks Using Semi-Large Scale Specimens
Authors: Pisˇtora Vladislav , Brumovsky´ Milan , Lauerova´ Dana
Published: July 2009

During years 2005–2006 a series of semi-large scale experiments of specimens containing underclad (embedded) crack were performed in NRI Rez. The experiments were performed within EU PHARE project in cooperation with VTT, FNS, TVONS (Finland) and Tecnatom (Spain). The aim of the project was to investigate fracture mechanics properties of cladding in particular to establish the role of cladding in fracture failure process of the specimens, and to exploit the obtained knowledge in the procedure for evaluation of integrity of WWER reactor pressure vessels. The specimens were manufactured from an artificially aged material of decommissioned (not operated) RPV of WWER-440 type. Each of specimens contained through-thickness crack embedded in the base material and was loaded by four-point-bending (4PB) loading. During loading, majority of specimens exhibited pop-ins followed by ductile tearing of cladding and final failure. Evaluation of the experiments concentrated on both process of cleavage fracture in base material and process of ductile tearing in cladding. In the presented paper, the results of experiments are described in detail. New proposal for procedure of assessment of RPV integrity, based on the PHARE project, was prepared and will be presented in a separate paper.
The Effect of Inspection Uncertainties in Predicting the Probability of Leak of Steam Generator Tubing Due to Pitting
Authors: Pandey M. D. , Datta S. V. , Komljenovic D.
Published: July 2009

Pitting corrosion is a serious form of degradation in steam generator (SG) tubing of some nuclear generating stations. The nature and extent of the pitting process is assessed through inspection programs, typically using eddy current probes. Although the data can be used in a probabilistic framework to predict the probability of a tube leak in future time intervals, the consideration of uncertainties associated with inspection tools is extremely important. The paper shows that the defect size data collected through inspections are significantly contaminated by probability of detection and measurement error such that the data present an incomplete picture of the defect population. The use of inspection data without accounting for inspection uncertainties in the analysis results in highly biased estimates of the leak probability. The paper presents a statistically correct likelihood method to account for inspection uncertainties and demonstrates its effectiveness through examples.
High-Cycle Thermal Fatigue in Mixing Tees: Large-Eddy Simulations Compared to a New Validation Experiment
Authors: Westin Johan , Veber Pascal , Andersson Lars , ’t Mannetje Carsten , Andersson Urban , Eriksson Jan , Henriksson Mats E. , Alavyoon Farid , Andersson Claes
Published: July 2009

The present paper describes new experimental data of thermal mixing in a T-junction compared with results from Large-Eddy Simulations (LES) and Detached Eddy Simulations (DES). The experimental setup was designed in order to provide data suitable for validation of CFD-calculations. The data is obtained from temperature measurements with thermocouples located near the pipe wall, velocity measurements with Laser Doppler Velocimetry (LDV) as well as single-point concentration measurements with Laser Induced Fluorescence (LIF). The LES showed good agreement with the experimental data also when fairly coarse computational meshes were used. However, grid refinement studies revealed a fairly strong sensitivity to the grid resolution, and a simulation using a fine mesh with nearly 10 million cells significantly improved the results in the entire flow domain. The sensitivity to different unsteady inlet boundary conditions was however small, which shows that the strong large-scale instabilities that are present in the mixing region are triggered independent of the applied inlet perturbations. A shortcoming in the performed simulations is insufficient near-wall resolution, which resulted in poor predictions of the near-wall mean velocity profiles and the wall-shear stress. Simulations using DES improved the near-wall velocity predictions, but failed to predict the temperature fluctuations due to high levels of modeled turbulent viscosity that restrained the formation of small scale turbulence.
Prediction of Hydrogen Explosions and Their Effects on Vessels: A Design Approach
Authors: Hart Andrew , Hamaoui Amir
Published: July 2009

The use of the simulation code, FLACS, to determine the onset of detonation in a deflagration explosion is a significant step forward in determining the behavior of a deflagration in piping systems. Over time, it is envisaged that the software will be greatly enhanced to fully model a DDT. Generally, in blast analysis, deflagration can be the worst loading condition. This is due to the amount of energy that a deflagration imposes on a structure. In a detonation explosion, the pressures are high but durations are short, thus the amount of energy being applied is small when compared to deflagrations that have a low pressure and a long duration. This long duration has a detrimental effect on the structure. In the analysis performed, no plastic deformation is allowed and the dynamic load factors (DLF’s) are expected to be high since the material is not allowed to absorb any of the impulsive energy which in blast design is taken by the plastic deformation of the material. The analysis results show that the longer the duration of the detonation portion of the DDT, the higher the DLF’s with the DLF’s utlimitely leveling off at the theoretical level of 2 for the top head models. The results show that pipe sizes do not have a significant factor on the results with the worse effective area occurring locally around the nozzle.
Viability of Ceramic High Temperature Heat Exchangers in NGNP Applications
Authors: Wilson Merrill A. , Lewinsohn Charles , Cutts James , Chen Yitung , Ponyavin Valery
Published: July 2009

The recent developments in the energy industry have kindled renewed interest in producing energy more efficiently. This has motivated the development of higher temperature cycles and their associated equipment. In this paper we will discuss several design configurations coupled with the inherent properties of preferred ceramic materials to assess the viability and design reliability of ceramic heat exchangers for next generation high temperature heat exchangers. These analyses have been extended to conceptually compare the traditional shell and tube heat exchanger with shell and plate heat exchangers. These analyses include hydrodynamic, heat transfer, mechanical stress and reliability models applicable to an Intermediate Heat Exchanger (IHX) and Process Coupling Heat Exchangers. It was found that ceramic micro-channel heat exchanger designs proved to have the greatest reliability due to their inherent mechanical properties, minimal thermo-mechanical stresses while improving the performance efficiency in a compact footprint.
Analyze and Simulation of a Typical MEMS RPG Using MCNP Code
Authors: Mohamadian sM. , Feghhi S. A. H. , Afarideh H.
Published: July 2009

Microbatteries are essential for portable electronics, cellular phones and MEMs devices to be miniaturized. Use of radioisotopes to realize nuclear microbatteries have been extensively researched. Electrical energy of a nuclear battery is produced from radioactive materials decaying by a suitable energy conversion process. Our approach in this paper is study of a direct collected charge to motion conversion. In this manuscript, the performance of radioisotope powered piezoelectric generator has been analyzed and simulated. The generator employs direct charging to convert radiated beta particles kinetic energy into stored electromechanical energy in a piezoelectric unimorph piezoelectricity to stored mechanical energy into extractable electrical energy.
Particle Size Characteristics of Molten Corium Quenched in Water
Authors: Min B. T. , Kim H. D. , Kim J. H. , Hong S. W. , Park I. K.
Published: July 2009

During a hypothetical severe accident in a nuclear reactor, a steam explosion might occur when molten corium interacts with water. The strength of a steam explosion affects the integrity of the containment of a nuclear reactor and is highly dependant on the characteristics of the melt-water-steam mixture. Since a break-up and fragmentation process during a pre-mixing are important mechanisms for a steam explosion behavior and affect the debris size distribution, the particle size characteristics of quenched corium have been investigated. For several years, series of experiments have been performed using prototypical corium in the TROI test facility with a high frequency induction heating using cold crucible technology. The molten corium was discharged into the cold water and the quenched debris particles were collected, sieved and examined for the effect of a size distribution on a steam explosion. The small corium droplets do not seem to contribute to a steam explosion owing to solidification at an early stage before the explosion but the large droplets contribute to it owing to their liquid state. It was also shown that single oxides and binary oxides with an eutectic composition (UO2/ZrO2 = 70/30 at weight percentage) led to steam explosions, but a binary oxide with a non-eutectic one did not. The mass mean diameters of the debris of the steam explosive composition was less than that of the non-steam explosive composition. Zirconia was the most energetic steam-explosive material in these tests, and an eutectic composition of corium also lead to a steam explosion, but a non-eutectic composition corium hardly led to a steam explosion. The particle sizes of the molten corium participating in a steam explosion were shown to be mainly 3–6 mm depending on the material and composition.
Relational Design Method for New Reactor Licensing
Authors: August J. K. , Hunter Joe
Published: July 2009

Document-based design is slow, error-prone, and inherently unfriendly maintaining a traceable source design basis. An alternative approach — relational design, maintains critical top level regulatory criteria design relationships based upon the fundamental technology framework. Virtually no systematic, process-induced errors result. Documents cannot match relational-based design methods, which effectively cover entire plant operating lifetimes. Compared with documents, modern relational data technology structures are more accurate, and cost less to build, maintain, and use over time. Applying relational methods will improve new nuclear construction and subsequent plant operations quality, accuracy, and speed — overall lowering costs.
Post-Test Analysis of an Integral Effect Reflood Test for an APR1400 Large-Break LOCA Scenario
Authors: Park Hyun Sik , Euh Dong-Jin , Choi Ki-Yong , Kim Yeon-Sik
Published: July 2009

Post-test analysis was performed on an integral effect reflood test, the ATLAS Test No. 9, for an APR1400 Large-Break LOCA (LBLOCA) scenario by using the MARS code. Integral effect tests on the reflood period of a large break LOCA were performed by using the ATLAS facility to help understand the thermal-hydraulic phenomena during the reflood period of a large-break LOCA for APR1400 and for resolving the current safety issues for the APR1400 licensing on the downcomer boiling phenomenon. The present ATLAS Test No. 9 is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period to provide reliable data to help validate the LBLOCA analysis methodology for APR1400.
Analysis on the Rotary Inertial of the Primary Pump for China Advanced Research Reactor
Authors: Zhong Jia , Tian Wenxi , Jiang Shengyao
Published: July 2009

The rotary inertia of the reactor’s primary pump is an important factor related to the reactor’s safety under some accidental conditions. If the rotary inertia of the primary pump is big, that the pump’s inertia time will be long and the coolant mass flow rate will decrease slowly. So it is helpful to remove the residual heat and enhance the safety of the reactor. On the other hand, a bigger rotary inertia increases the cost, and causes some other inconvenience. In this paper, our research object is China Advanced Research Reactor (CARR) which is under construction. Systemic mathematical and physical models were set up and a program was developed using FORTRAN language with GEAR numerical method to analyze the transient of CARR. The primary thermal-hydraulic parameters of the reactor core were obtained in case of loss of power accident under different rotary inertia of the pump. These parameters include the temperature of the coolant the clad and the fuel, the maximal quality of the reactor core and MDNBR, etc. On the basis of the calculate result, it can be found that when the rotary inertia of the pump is larger than 150 Kg·m2, all of the thermal parameters in loss of power accident meet the safety requirement. And if the rotary inertia of the pump is 450 Kg·m2 which is the design value of CARR, the delay shutdown time in loss of power accident must not exceed 20 seconds.
Development of Atlas of Defects of Steam Generators Heat Exchanging Tubes of NPPs With WWER and Analysis of Testing Results
Authors: Bakirov Murat , Chubarov Sergei , Trunov Nikolay
Published: July 2009

The basic method of the operational inspection of metal of heat exchanging tubes (HET) of steam generators (SG) is eddy-current multifrequency method all over the world. The greatest distribution was received variant with eddy-current testing (ECT) by use of a bobbing probe on the Russian nuclear power plants (NPP). Tubes with the defects which a subject to plugging are defined by results of lead operational ECT over the certain algorithm. SG resource is settled and replacement is required at achievement of a limit on number of the plugged tubes [1].
A Method for Selecting Software for Dynamic Event Analysis: II — The Taylor Anvil and Dynamic Brazilian Tests
Authors: Lacy J. M. , Novascone S. R. , Richins W. D. , Larson T. K.
Published: July 2009

New nuclear power reactor designs will require resistance to a variety of possible malevolent attacks as well as traditional dynamic accident scenarios. The design/analysis team may be faced with a broad range of phenomena including air and ground blasts, high-velocity penetrators or shaped charges, and vehicle or aircraft impacts. With a host of software tools available to address these high-energy events, the analysis team must evaluate and select the software most appropriate for their particular set of problems. The accuracy of the selected software should then be validated with respect to the phenomena governing the interaction of the threat and structure. Several software codes are available for the study of blast, impact, and other shock phenomena. At the Idaho National Laboratory (INL), a study is underway to investigate the comparative characteristics of a group of shock and high-strain rate physics codes including ABAQUS, LS-DYNA, CTH, ALEGRA, ALE3D, AUTODYN, and RADIOSS. In part I of this report published in 2007, a series of five benchmark problems to exercise some important capabilities of the subject software was identified. The benchmark problems selected are a Taylor cylinder test, a split Hopkinson pressure bar test, a free air blast, the dynamic splitting tension (Brazilian) test, and projectile penetration of a concrete slab. Part II — this paper — reports the results of two of the benchmark problems: the Taylor cylinder and the dynamic Brazilian test. The Taylor cylinder test is a method to determine the dynamic yield properties of materials. The test specimen is a right circular cylinder which is impacted against a theoretically rigid target. The cylinder deforms upon impact, with the final shape depending upon the specimen density, the impact velocity, and the dynamic yield stress, in turn a function of strain and strain rate. The splitting tension test, or Brazilian test, is a method to measure the tensile strength of concrete using a cylindrical specimen. The specimen is loaded diametrically in compression, producing a fracture at the center of the specimen that propagates toward the loading points until the cylinder is split. To generate a dynamic load, different methods such as a drop-weight or a split Hopkinson pressure bar are employed. The Taylor anvil and dynamic Brazilian test analyses are presented, including discussion of the analysis approach for each of the five subject software packages and two vendor submittals; comparison of results both among the codes and to physical test results; and conclusions as to the applicability of the subject codes to these two problems. Studies of the remaining three benchmark problems and overall conclusions will be presented in future publications.
Hitachi-GE Advanced Construction Management Technologies
Authors: Onda Kimiharu , Takada Masatoshi , Okada Hisako , Tatehora Kazuto
Published: July 2009

Hitachi has so far developed the construction management system with information technology since more than twenty years ago. As the recent results, we developed the advanced new functions and applied them to the Nuclear Power Plant (NPP)’s under construction. RFID-applied construction management method with personal digital assistant (PDA) could realize more reliable and efficient construction control on a site. By enhancing an Integrated Schedule Planning function, the construction scheme could be more accurately and speedily developed, which made it possible to realize the standardization of schedule planning, the linkage with other systems, and the improvement of operability. By applying these new technologies to Hitachi construction management, Hitachi can supply the nuclear power plant with higher quality, safety, and reliability.
Effect of CANDU Bundle-Geometry Variation on Dryout Power
Authors: Leung Laurence K. H.
Published: July 2009

Dryout powers have been evaluated at selected inlet-flow conditions for two proposed designs of CANDU® bundles and compared to those of the 37-element and CANFLEX® bundles. These proposed designs consist of a large centre element (18 mm for one design and 20 mm for the other) and three rings of elements of 11.5 mm in outer diameter. The critical heat flux for each bundle design has been predicted using the correlation derived with Freon data obtained from the corresponding full-scale bundle test. An improvement in dryout power has been shown for the proposed design having a 20-mm centre element with a radial power profile corresponding to the natural-uranium fuel as compared to other bundles, particularly the natural-uranium 37-element bundle, with a symmetric cosine axial power profile. The dryout power improvement is further enhanced for the upstream-skewed axial power profile.
Use of Polyethylene (PE) Pipe in Safety-Related, Class 3, Service-Water Piping
Authors: Krishnaswamy Prabhat , Focht Eric M. , Shim Do-Jun , Zhang Tao
Published: July 2009

The ASME Boiler and Pressure Vessel Code Committee (BPVC) has recently published Code Case N-755 that describes the requirements for the use of Polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. The code case was developed by Special Working Group–Polyethylene Pipe (SWG-PP) within Section III (Design) of the BPVC. This paper provides a critical review of the design requirements described in CC N-755 from pressure boundary integrity considerations. The various technical issues that need to be addressed for safety-critical PE piping are discussed in this paper. Specifically, the premise of allowing defects in pipe that are 10% of the wall thickness has been reviewed especially for cases involving large diameter piping [> 304.8 mm (12 inches)] that is to be operated at elevated temperatures as high as 60°C (140°F). One of the common modes of failure in PE piping under sustained internal pressure is due to slow crack growth (SCG) from manufacturing or installation defects in the pipe wall. The effect of pipe diameter and stresses on the crack driving force for a 10% deep flaw is calculated for comparison with the material resistance to SCG at elevated temperatures.
Numerical Analysis of Thermal-Hydraulic Behavior of Supercritical Water in Vertical Upward/Downward Flow Channels
Authors: Gu Hanyang , Cheng Xu , Liu Xiaojing
Published: July 2009

Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical fluids. In this paper, the numerical analysis is carried out to study the thermal-hydraulic behaviour in vertical sub-channels cooled by supercritical water. Remarkable differences in characteristics of secondary flow are found, especially in square lattice, between the upward flow and downward flow. The turbulence mixing across sub-channel gap for downward flow is much stronger than that for upward flow in wide lattice when the bulk temperature is lower than pseudo-critical point temperature. For downward flow, heat transfer deterioration phenomenon is suppressed with respect to the case of upward flow at the same conditions.
Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig
Authors: Efanov A. D. , Kalyakin S. G. , Morozov A. V. , Remizov O. V. , Tsyganok A. A. , Generalov V. N. , Berkovich V. M. , Taranov G. S.
Published: July 2009

In new Russian NPP with VVER-1200 reactor (V-392M reactor plant) in the event of an accident being due to the rupture of the reactor primary circuit and accompanied by the loss of a.c. sources, provision is made for the use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. To experimental investigation of the condensation mode of operation of VVER steam generator, a large scale HA2M-SG test rig was constructed. The test rig incorporates: tank-accumulator, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger simulator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure – 1.6 MPa, temperature – 200 Celsius degrees. Experiments at the HA2M-SG test rig have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond basis accident. The report presents the test procedure and the basic obtained test results.
Numerical Simulation Using Realizable k-e Turbulence Model of Jet Flow Mixing Within Rod Bundles
Authors: Salpeter Nathaniel O. , Hassan Yassin A.
Published: July 2009

In this investigation, a steady state simulation of jet mixing in a rod bundle with varied jet Reynolds numbers was performed with CD-Adapco’s Star-CCM+ computational fluid dynamics (CFD) code utilizing the realizable k-ε turbulence model and two-layer y+ wall treatment. The goal of the work is to investigate the ability of the realizable k-ε turbulence model to predict phenomena expected in the exit plenum of the prismatic gas cooled reactor as well as provide a benchmark for future numerical investigation into the validity of certain turbulence models for rod bundle simulation. Numerical results were qualitatively verified against available experimental data [6]. Two injection sites were present in the rod bundle with impingement of the jets on the lower plane of the domain. The Reynolds numbers specified for the inlets varied between 6,300 and 12,700 for different cases. A mesh of over four million polyhedron cells was generated to capture critical flow characteristics within the domain. Polyhedral cells were chosen for this application because they provided a better quality mesh and reduced the total number of cells necessary to achieve accurate results [4]. The simulations carried out were defined as having reached a converged solution when all residuals reduced to less than 10−5. The simulation of the flow in the rod bundle was successful in providing insight into locations of some key recirculation zones and the dependence they have on inlet conditions. The comparison between numerical and experimental results showed similar key flow patterns as well as aided in possible points of focus for future investigation. The information obtained and conclusions drawn will be critical in future numerical benchmarks in this area of research.
The Potential Application of Metal Binding Hydro- and Silica-Gels in Low-Level Radioactive Waste Processing
Authors: Tokuhiro Tadashi , Carey Josh W. , Bertino Massimo , Tokuhiro Akira
Published: July 2009

Within the management of radioactive waste, we sought to consider a new approach radioactive hazardous waste processing in aqueous or similar (low-level waste; LLW) forms LLW and in fact, ‘contaminants of concern’ is often stored as diluted aqueous solutions of radioactive (or non) elements and contained in storage containers. One of the general problems associated with mixed liquid waste is the lack of an efficient, effective, and inexpensive means of processing (separating) its constituents. Two of the objectives in processing solid, radioactive laden liquid LLW are as follows: 1) to separate/extract the radioisotopes from the rest of the mixed constituents, and 2) to produce stable solidified forms encapsulating radioactive elements. Recent R&D in the physical chemistry of gel materials, have identified promising approach to simultaneously achieve the above objectives. That is, by utilizing and manipulating the physicochemical properties of various silica- and polymer-based gels at the nanoscale, we have demonstrated a process by which to specific chemical species are encapsulated.
Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts — Steam Cycle Options
Authors: Duffey R. B. , Pioro I. , Zhou X. , Zirn U. , Kuran S. , Khartabil H. , Naidin M.
Published: July 2009

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.
Bayesian Prediction for the Gumbel Distribution Applied to Feeder Pipe Thicknesses
Authors: Hoppe Fred M. , Fang Lin
Published: July 2009

This paper develops Bayesian prediction intervals for the minimum of any specified number of future measurements from a Gumbel distribution based on previous observations. The need for such intervals arises in the analysis of data from outlet side feeder pipes at Ontario nuclear power plants. The issue is how to best use these measurements in order to arrive at a statistically sound conclusion concerning the minimum thickness of all remaining uninspected pipes, in particular with what confidence can it be asserted that the remaining wall thicknesses are above an acceptable minimum to ensure a sufficiently high thickness up to the end of the next operating interval. The result gives a probability measure of the potential benefit of performing additional inspections when considered against the additional radiation exposure and the cost of performing additional inspections. Previously, this problem was approached by adapting a classical prediction interval that was originally derived for normal data. Here we examine both a hybrid Bayesian method that combines Bayesian ideas with maximum likelihood and also a full Bayesian approach using Markov Chain Monte Carlo. We show that the latter gives larger lower prediction limits and therefore more margin to fitness for service.
Fragmentation of a Single Molten Copper Droplet Compared With Molten Copper Jets Penetrating a Sodium Pool
Authors: Zhang Zhi-Gang , Itagaki Wataru , Sugiyama Ken-Ichiro , Narabayashi Tadashi , Nishimura Satoshi , Kinoshita Izumi
Published: July 2009

The progression of hypothetical core disruptive accidents in metallic fuel fast breeder reactors is strongly affected by the exclusion of molten metallic fuels from the core region due to the molten fuel-coolant interaction (FCI). As a basic simulation study of FCI, the present paper focuses on the fragmentation of a single molten copper droplet from 1g to 5g, which penetrated a sodium pool at instantaneous contact interface temperatures Ti from 995°C to 1342°C. Intensive fragmentation of single molten copper droplet was clearly observed in all runs even if Tis are below the melting point (1083°C) of copper except low Tis. In the low Tis condition the mass median diameters Dm of droplet fragments both the same and different mass scatter widely. When Tis are somewhat below and above the melting point, the Dms of copper droplet with the different mass of 5 times or the same mass differ very little, nearly the same. The present Dms show somewhat larger distribution than the Dms of molten copper jets with large mass in the low superheating condition, which were previously reported by the authors. The correlation between molten metal droplets and jets found in the present study is very useful to conservatively predict the fragment size in the FCI of molten jets.
Experiments and Analytical Simulation on Steam Injector Driven Passive Core Injection System for Innovative-Simplified Nuclear Power Plant
Authors: Ohmori Shuichi , Narabayashi Tadashi , Mori Michitsugu , Watanabe Fumitoshi
Published: July 2009

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.
Development of a Next Generation Canister Sipping System
Authors: Longenecker Richard E. , Wood John D.
Published: July 2009

In order to meet the rising demand for fuel leak detection and Dry Cask Storage Services, Westinghouse Electric Company chose to re-design and market the industry-proven failed fuel detection technique of canister sipping. Based on a custom fast-plastic scintillator gamma detector, Westinghouse chose simply not to re-build its legacy system. Instead, the system was redesigned from the ground up, considering materials, limitation of overall size and weight, system control, ease of use, and decontamination. The result was a brand new design and unique improvements utilizing new, yet proven technologies. The following paper will illustrate the considerations that engineering undertook while in development of the new system. These include industry lessons learned and the fundamental improvements considered. New aspects of the design will be highlighted, including steps to be taken to improve ease of decontamination, a unique sliding lid, Nuclear Instrument Module (NIM-Bin) replacement electronics, and an up-ending system. Steps taken to decrease weight and overall footprint size will be illustrated. Lastly, improvements in control structure, and a highlight of WeSiP, Westinghouse’s new object-oriented Global Sipping Platform Software, used to log data and power the system will be examined. With these unique improvements, the new Westinghouse Canister Sipping System meets the industry’s demands for a reliable system to identify failed fuel.
Insights Into Operating a Low Pressure Feed Pump for Loss of All AC and DC Power
Authors: Baron Barbara R. , Lutz Robert J.
Published: July 2009

Generic guidance for Pressurized Water Reactors (PWRs) has been developed to address the beyond design basis event of coincident loss of all Alternating Current (AC) and Direct Current (DC) power. The generic guidance included a strategy to use a low pressure feed pump to provide adequate secondary side heat removal via the Steam Generators (SGs) to delay or prevent core uncovery following loss of all AC power with battery depletion, loss of all DC power, seismic initiated events, and/or terrorist initiated events. The purpose of the project was to use thermal hydraulic analyses, operating experience, and other engineering analyses to identify and evaluate technical issues associated with the implementation of the low pressure feed pump strategy at Westinghouse and Combustion Engineering (CE) designed plants. The technical issues that were evaluated are those issues typically addressed in the development of a plant’s Emergency Operating Procedures (EOPs) and Off-Normal Operating Procedures (ONOPs). The thermal hydraulic analyses were performed using the computer code MAAP 4.0.5 and a plant model of a 4-loop Westinghouse designed PWR. The results of the analyses are also applicable to 2-Loop and 3-Loop Westinghouse and CE designed PWRs. The results of the evaluation indicated that the key technical issue potentially impacting the prevention of core uncovery for the implementation of the low pressure feed pump strategy is the potential and consequences of injecting nitrogen into the Reactor Coolant System (RCS) from the cold leg accumulators/Safety Injection Tanks (SITs). The results of the evaluation were used to develop sample instructions for implementing the low pressure feed pump strategy for Westinghouse and CE designed PWRs. The sample instructions were developed for two categories of low pressure feed pumps: (1) low pressure feed pumps with shutoff heads greater than the pressure that nitrogen injects into the RCS and (2) low pressure feed pumps with shutoff heads less than the pressure that nitrogen injects into the RCS. The usefulness of the sample instructions is maximized when the low pressure feed pump has local flow indication, throttling capabilities, deadhead protection, and local SG pressure indication is available. The results demonstrated that the design characteristics of the low pressure feed pump are important to prolonging/preventing the time of core uncovery.
Numerical Prediction and Optimization of Depressurized Sodium-Water Reaction Experiment With Counter-Flow Diffusion Flame
Authors: Yamaguchi Akira , Takata Takashi , Ohshima Hiroyuki , Kohara Yoshitaka
Published: July 2009

Sodium-water reaction (SWR) is a design basis accident of a Sodium Fast Reactor (SFR). A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. Typical phenomenon is that the pressurized water blows off, vaporizes and mixes with the liquid sodium. In this paper, a new computer program has been developed and the SWR in a counter-flow diffusion flame is studied by a numerical simulations and an experiment. The experiment is designed with the numerical simulation so that the stable reaction flame is maintained for long time and physical and chemical quantities are measured. From the comparison of analysis and experiment, there exist discrepancies that may be caused by the assumptions of chemical reaction. Hence, a new experiment is proposed to enhance the measurement accuracy and to investigate the reason of the disagreement. The authors propose a depressurized experiment. With the depressurization, it is expected the flame location can be controlled and the reaction region becomes thicker because of decrease in reactant gas density.