bluuebunny/crossref_metadata_2025 huggingface dataset:
Impact Analysis of a Dipper-Type and Multi Spring-Type Fuel Rod Support Grid Assemblies in PWR
Authors: Song K. N. , Park K. J. , Yoon K. H. , Park G. J. , Kang B. S.
Published: March 2009

A spacer grid is one of the main structural components in a fuel assembly of a Pressurized light Water Reactor (PWR). It supports fuel rods, guides cooling water, and maintains geometry from external impact loads. A simulation is performed for the strength of a spacer grid under impact load. The critical impact load that leads to plastic deformation is identified by a free-fall test. A finite element model is established for the nonlinear simulation of the test. The simulation model is tuned based on the free-fall test. The model considers the aspects of welding and the contacts between components. Nonlinear finite element analysis is carried out by a software system called LS/DYNA3D. The results are discussed from a design viewpoint.
Importance Analysis of In-Service Testing Components for Ulchin Unit 3
Authors: Kang Dae-Il , Kim Kil-Yoo , Ha Jae-Joo
Published: March 2009

We performed an importance analysis of In-Service Testing (IST) components for Ulchin Unit 3 using the integrated evaluation method for categorizing component safety significance developed in this study. The importance analysis using the developed method is initiated by ranking the component importance using quantitative PSA information. The importance analysis of the IST components not modeled in the PSA is performed through the engineering judgment, based on the expertise of PSA, and the quantitative and qualitative information for the IST components. The PSA scope for importance analysis includes not only Level 1 & 2 internal PSA but also Level 1 external and shutdown/low power operation PSA. The importance analysis results of valves show that 167 (26.55%) of the 629 IST valves are HSSCs and 462 (73.45%) are LSSCs. Those of pumps also show that 28 (70%) of the 40 IST pumps are HSSCs and 12 (30%) are LSSCs.
Mechanical Properties of 20% Cold-Worked 316 Stainless Steel Irradiated at Low Dose Rate
Authors: Allen Todd R. , Tsai Hanchung , Cole James I. , Ohta Joji , Dohi Kenji , Kusanagi Hideo
Published: March 2009

To assess the effects of long-term, low-dose-rate neutron exposure on mechanical strength and ductility, tensile properties were measured on irradiated 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1–47 dpa at temperatures from 371–385°C and dose rates from 0.8–2.8 × 10−7 dpa/s. These dose rates are about one order of magnitude lower than those of typical EBR-II in-core experiments. Irradiation cuased hardening, with the yield strength (YS) following approximately the same trend as the ultimate tensile strength (UTS). At higher dose, the difference between the UTS and YS decreases, suggesting the work-hardening capability of the material is decreasing with increasing dose. Both the uniform elongation and total elongation decrease up to the largest dose. Unlike the strength data, the ductility reduction showed no signs of saturating at 20 dpa. While the material retained respectable ductility at 20 dpa, the uniform and total elongation decreased to <1 and <3%, respectively, at 47 dpa. Fracture in the 30 dpa specimen is mainly ductile but with local regions of mixed-mode failure consisting of dimples and microvoids. The fracture surface of the higher-exposure 47 dpa specimen displays significantly more brittle features. The fracture consists of maily small facets and slip bands that suggest channel fracture. The hardening in these low-dose-rate components differs from that measured in test samples irradiated in EBR-II at higher-dose-rate. The material irradiated at higher dose rate loses work hardening capactiy faster than the lower dose rate material, although this effect could be due to compositional differences.
In-Situ Pressure Testing of Steam Generator Tubes
Authors: Chandler Cordelia K.
Published: March 2009

The anatomist John Hunter said, “But why think? Why not try the experiment?” His words are applicable to in-situ pressure testing as a means of performing condition monitoring of steam generator tubes. Other methods of performing condition monitoring include the application of theoretical burst and leakage equations (deterministic methods), probabilistic analyses, and tube pull. The deterministic and probabilistic methods must address uncertainties in the material properties and flow measurements. Destructive examination of a pulled tube provides the verification of the damage mechanism and possible causes of the degradation; however, the axial loads required to pull the tube may affect the degradation. The results of a tube pull are generally not available until after the plant has returned to service. In-situ pressure testing provides structural and leakage integrity of SG tubing conditions. Candidates are selected for in-situ testing based on te EPRI Guidelines and include an evaluation of crak depth, length and/or voltage response from the eddy-current inspection results. Appropriate selection of tubes for in-situ testing provides confidence that the results adequately address the structural and leakage integrity of the SG tube population. In-situ pressure testing, combined with conservative operational assessments for the next cycle, assures that the integrity of the SG tubing is maintained.
Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids
Authors: Ratnayake Ruwan K. , Hochreiter L. E. , Ivanov K. N. , Cimbala J. M.
Published: March 2009

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained from operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids.
Generation and Control of Electricity in HTGR Helium Turbine Plant
Authors: Pradeep Kumar K. N. , Tourlidakis A. , Pilidis P.
Published: March 2009

This paper analyses the operational aspects of a temperature Gas Reactor using direct Helium Cycle for power conversion. As an engineering project, the complexity in the operational aspects of HTGR’s is more severe than that of its design and installation. Its transient stability is achieved in a different manner compared to conventional power plants. An efficient and steady operation is a must for its successful realization. The paper looks into the start-up, shutdown and part load performance of helium power conversion system. The study is based on an ongoing developmental project in South Africa called Pebble Bed Modular Reactor (PBMR).
ENHS: The Encapsulated Nuclear Heat Source — A Nuclear Energy Concept for Emerging Worldwide Energy Markets
Authors: Wade D. C. , Feldman E. , Sienicki J. , Sofu T. , Barak A. , Greenspan E. , Saphier D. , Brown N. W. , Hossain Q. , Carelli M. D. , Conway L. , Dzodzo M.
Published: March 2009

A market analysis is presented which delineates client needs and potential market size for small turnkey nuclear power plants with full fuel cycle services. The features of Encapsulated Nuclear Heat Source (ENHS) which is targeted for this market are listed, and the status of evaluation technological viability is summarized.
Experimental Study on Heat Transfer of Single-Phase Flow and Boiling Two-Phase Flow in Vertical Narrow Annuli
Authors: Qiu Suizheng , Takahashi Minoru , Su Guanghui , Jia Dounan
Published: March 2009

Water single-phase and nucleate boiling heat transfer were experimentally investigated in vertical annuli with narrow gaps. The experimental data about water single-phase flow and boiling two-phase flow heat transfer in narrow annular channel were accumulated by two test sections with the narrow gaps of 1.0mm and 1.5mm. Empirical correlations to predict the heat transfer of the single-phase flow and boiling two-phase flow in the narrow annular channel were obtained, which were arranged in the forms of the Dittus-Boelter for heat transfer coefficients in a single-phase flow and the Jens-Lottes formula for a boiling two-phase flow in normal tubes, respectively. The mechanism of the difference between the normal channel and narrow annular channel were also explored. From experimental results, it was found that the turbulent heat transfer coefficients in narrow gaps are nearly the same to the normal channel in the experimental range, and the transition Reynolds number from a laminar flow to a turbulent flow in narrow annuli was much lower than that in normal channel, whereas the boiling heat transfer in narrow annular gap was greatly enhanced compared with the normal channel.
HLMC Fast Reactor With Complete Natural Circulation
Authors: Enuma Yasuhiro , Mizuno Tomoyasu , Mihara Takatsugu , Konomura Mamoru , Mito Makoto , Tanji Mikio
Published: March 2009

To seek for a promising concept of a heavy liquid metal coolant (HLMC) fast reactor plant, Japan Nuclear Cycle Development Institute (JNC) and the electric utilities conducted conceptual design study on various types of plant concepts and compared these concepts based on technical feasibility and economical perspective. The Pb-Bi cooled complete natural circulation reactor concept may attain high safety level and construction cost goal (¥200,000/kWe).
Direct ECC Bypass Phenomena During LBLOCA Reflood Phase Observed in the MIDAS Test Facility
Authors: Yun B. J. , Kwon T. S. , Euh D. J. , Chu I. C. , Song C.-H. , Park J. K.
Published: March 2009

One of the advanced design features of the APR-1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is in progress. In this paper, test results of a direct ECC bypass performed in the steam-water test facility called MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) is presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’ with the 1/4.93 length scale. From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region is obtained.
Experimental Study on Flow Technology and Steel Corrosion of Lead-Bismuth
Authors: Takahashi Minoru , Sekimoto Hiroshi , Ishikawa Kotaro , Sawada Naoki , Suzuki Tadashi , Hata Koji , Yoshida Susumu , Qiu Suizheng , Yano Toyohiko , Imai Masamitsu
Published: March 2009

For the feasibility study of Pb-Bi-cooled fast reactors (FR) and the Pb-Bi target of accelerator-driven nuclear transmutation systems , Pb-Bi flow technologies were developed and steel corrosion behavior in a Pb-Bi flow was investigated using a Pb-Bi circulation loop. The performance of an electro-magnetic flow meter with electrically insulated electrodes plated with Rh was better than those of conventional and tubular types. Oxygen concentration was controlled by continuous injection of Ar, H2 and H2O mixture gas into the Pb-Bi flow. In order to have desired oxygen potential, the partial pressure ratio of PH2/PH2O was chosen in the range from 0.12 to 2.2 by bubbling the mixture of Ar and H2 in water columns at the room temperature. By injecting the mixture gas into the loop for sufficient time, the oxygen potentials measured by the oxygen sensor made of solid electrolyte ZrO2-Y2O3 agreed well with those in the injected gas mixture. In the first corrosion test, steels were exposed to a Pb-Bi flow at the temperature of 550 °C, the velocity of 2 m/s and the oxygen concentration of ∼5.0×10−7 wt.% for 959 hours. It was found that the weight loss was larger in the order of SS316, low Cr steel (SCM420) and high Cr steels (STBA26, SUS405, SUS430). Corrosion was suppressed by a Cr oxide layer for high Cr steels. A porous layer was formed on SS316 surface due to high solubility of Ni in Pb-Bi. In the second corrosion test, the oxygen concentration was kept at 3.6×10−7 wt.% by injecting Ar, H2 and H2O mixture gas into a Pb-Bi flow, and steels were exposed to a Pb-Bi flow at the temperature of 550 °C, the velocity of 2 m/s for 1000 hours. Serious erosion damage was observed in SCM420 at the entrance, and some erosion damages appeared in low Cr steels: SCM420, F82H, STBA26 and HCM12 downstream. Crack type damage was observed on the surface of HCM12, and pitting-type damage was observed on the surface of 2 1/4Cr-1Mo steel. Some penetration of Pb-Bi into the materials appeared in some of the erosion-damaged steels.
Lead-Bismuth Activities at the Karlsruhe Lead Laboratory KALLA
Authors: Knebel Joachim U. , Mu¨ller Georg , Konys Ju¨rgen
Published: March 2009

At Forschungszentrum Karlsruhe (FZK) the characteristics of an accelerator-driven subcritical system (ADS) are evaluated, mainly with respect to the potential of transmutation of minor actinides and long-lived fission products, to the feasibility and to safety aspects. All experimental activities, which are related to lead-bismuth as cooling fluid and spallation material, are performed in the KArlsruhe Lead LAboratory KALLA. This article gives an overview on KALLA, which has three stagnant experiments and three loop experiments. The stagnant experiments are concentrating on corrosion mechanisms, surface treatment, oxygen sensor development, and oxygen control system (OCS), the loop experiments are concentrating on thermalhydraulic measurement techniques, ADS-relevant component testing, and corrosion investigations in flowing lead-bismuth. A fourth loop experiment is planned to investigate the integral heat removal from a 4 MW spallation target for normal and decay heat removal conditions. Among others, latest results are presented of: characteristics of oxygen sensors in flowing liquid Pb-Bi, the oxygen control system (OCS) operating on a loop system, an ultrasonic flow meter applied to lead-bismuth at 400°C. In addition, results are given on the improvement of the corrosion resistivity of steels in flowing lead-bismuth, using a special temperature treatment method (electron beam facility GESA) and alloying aluminium in the surface layer.
Dynamic Loads Due to Blowdown Through a Sparger
Authors: Bae Yoon-Yeong , Park Choon-Kyung , Cho Seok , Chun Se-Young , Song Chul-Hwa , Park Jong-Kyun
Published: March 2009

An experiment has been performed using a facility, which simulates the safety depressurization system (SDS) and in-containment refueling water storage tank (IRWST) of APR1400, an advanced PWR being developed in Korea, to investigate the dynamic load resulting from the blowdown of steam from a steam generator through a sparger. The influence of the key parameters, such as air mass, steam pressure, submergence, valve opening time, and pool temperature, on frequency and peak loads was investigated. The blowdown phenomenon was analyzed to find out the real cause of the initiation of bubble oscillation and discrepancy in frequencies between the experiment and calculation by conventional equation for bubble oscillation. The cause of significant damping was discussed and is presumed to be the highly tortuous flow path around bubble. The Rayleigh-Plesset equation modified by the introduction of method of image reasonably reproduce the bubble oscillation in a confined tank. Right after the completion of air discharge the steam discharge immediately follows and it condenses abruptly to provide low-pressure pocket. It may contribute to the negative maximum being greater than positive maximum. The subsequently discharging steam does not play as the driving force anymore. The effect of various parameters on the peak pressure and frequency was analyzed. The air mass, valve opening time, and submergence give significant effect on both of frequency and peak pressure while steam mass flux gives questionable effect and pool temperature gives almost none.
Analysis and Insights About FE-Calculations of the EC-Forever-Experiments
Authors: Willschuetz H.-G. , Altstadt E. , Weiss F.-P. , Sehgal B. R.
Published: March 2009

To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments are currently underway at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behaviour of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. Due to the multi axial creep deformation of the vessel with a non-uniform temperature field these experiments are on the one hand an excellent source of data to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used to optimize the experimental procedure and by supporting decision making during the experiment. For that, a Finite Element model is developed based on a multi-purpose code. After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Additionally metallographic post test investigations of the experiments are conducted to improve the numerical damage model and to adjust the correlation between the metallographic observations and the calculated damage. Taking into account both — experimental and numerical results — gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analysing the calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.
Steam Generators Replacement and Power Uprate at Tihange 2 Power Plant
Authors: Allard R. , Delhaye X. , Gre´er J. L. , Thierens D. , Wilmart J. P. , Ryckaert R.
Published: March 2009

The Belgian Tihange 2 nuclear power plant went into commercial operation in 1983 producing a thermal power of 2785 MW. In 1995, the thermal power output was increased up to 2905 MW and the fuel cycle extended to 15 months. Since the commissioning of the plant, the steam generators U-tubes have been affected by primary stress corrosion cracking. In order to avoid further degradation of the performance and an increase in repair costs, Electrabel, the owner of the plant, decided in 1997 to replace the 3 steam generators. This decision was supported by the feasibility study performed by Tractebel Energy Engineering which demonstrated that an increase of 10% of the initial power was achievable together with a fuel cycle length of 18 months. Tractebel Energy Engineering was entrusted by Electrabel to manage the project. A multi-contract strategy was adopted. The new steam generators, designed by Mitsubishi, allow raising the thermal power to 3064MW which is 110% of the initial power by an increase of the primary to secondary heat transfer area. The safety analyses necessary to justify the new operation point and the fuel cycle extension to 18 months were performed by Framatome in association with Tractebel Energy Engineering. The work on site took place during the summer of 2001 and was managed by Tractebel Energy Engineering. The SG replacement itself was performed in 17.5 days by Westinghouse PCI and the plant was reconnected to the grid on August 11, after an outage of 63 days (grid to grid). This paper presents various aspects related to the steam generators replacement project, such as the safety analysis program together with the various works on site, project management, organization, and scheduling.
Corium Spreading Over Concrete: The Vulcano VE-U7 and VE-U8 Tests
Authors: Journeau Christophe , Boccaccio Eric , Fouquart Pascal , Je´gou Claude , Piluso Pascal
Published: March 2009

Two experiments have been performed in the VULCANO facility in which prototypic corium has been spread over concrete. In the VE-U7 test, a mixture representative of what can be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. In the VE-U8 test, a UO2-ZrO2 mixture, prototypic of in-vessel corium, has been spread over a lime-siliceous concrete. Although residual power was not simulated in this experiment, up to 2 cm of concrete have been eroded during the test. Results in terms of spreading behaviour, effects of gases, concrete erosion and thermal attack are presented and discussed.
Modeling of the Fluid Flow and Heat Transfer an a Pebble Bed Modular Reactor Core With a Computational Fluid Dynamics Code
Authors: Taylor J. Bryce , Yavuzkurt Savas , Baratta Anthony J.
Published: March 2009

The Pebble Bed Modular Reactor (PBMR), a promising Generation IV nuclear reactor design, raises many novel technological issues for which new experience and techniques must be developed. This brief study explores a few of these issues, utilizes a computational fluid dynamics code to model some simple phenomena, and points out deficiencies in current knowledge that should be addressed by future research and experimentation. A highly simplified representation of the PBMR core is analyzed with FLUENT, a commercial computational fluid dynamics code. The applied models examine laminar and turbulent flow in the vicinity of a single spherical fuel pebble near the center of the core, accounting for the effects of the immediately adjacent fuel pebbles. Several important fluid flow and heat transfer parameters are examined, including heat transfer coefficient, Nusselt number, and pressure drop, as well as the temperature, pressure, and velocity profiles near the fuel pebble. The results of these “unit cell” calculations are also compared to empirical correlations available in the literature. As FLUENT is especially sensitive to geometry during the generation of a computational mesh, the sensitivity of code results to pebble spacing is also examined. The results of this study show that while a PBMR presents a novel and complex geometry, a code such as FLUENT is suitable for calculation of both local and global flow characteristics, and can be a valuable tool for the thermal-hydraulic study of this new reactor design. FLUENT results for pressure drop deviate from the Darcy correlation by several orders of magnitude in all cases. When determining the heat transfer coefficient, FLUENT is again much lower than Robinson’s correlation. Results for Nusselt number show better agreement, with FLUENT predicting results that are 10 or 20 times as large as those from the Robinson and Lancashire correlations. These differences may arise because the empirical correlations concern mainly integral parameters, while the FLUENT model focuses on local flow behaviors. Local phenomena are significant in the case of local heat transfer characteristics, fine temperature distribution calculations to identify hot spots, and fission product transport phenomena. All of these are important to a safety analysis of the PBMR reactor during normal operation, as well as during transient circumstances, and should be the focus of future research efforts.
Power Optimization in the STAR-LM Modular Natural Convection Reactor System
Authors: Sienicki James J. , Spencer Bruce W.
Published: March 2009

STAR-LM is a modular, factory-fabricated, overland transportable, proliferation resistant, autonomous load following, and passively safe liquid metal-cooled fast reactor system that utilizes inert lead-bismuth eutectic coolant combined with 100+% natural circulation heat transport to achieve radial design simplification, enhanced reliability, and cost savings. STAR-LM has the potential to meet all of the U.S.DOE Generation IV goals of sustainable energy development, safety and reliability, and economics. Previous development of STAR-LM resulted in a 300 MWt modular pool type reactor in which modular steam generators are immersed directly inside the primary coolant and the core is part of a reactor module/flow-thru fuel cartridge that provides no access to fuel during the core lifetime. Recent concept development has focused on raising the power achievable in a small module size based on preserving key criteria for: i) full spectrum of modes of module transport from factory to site (including rail transport); ii) ultralong core cartridge lifetime; iii) 100% natural circulation heat transport; and iv) coolant and cladding peak temperatures well within the existing (mainly Russian) database for lead-bismuth eutectic coolant and ferritic steel core materials. For example, natural circulation is found to be capable of transporting 400 MWt in a fully transportable module size with 15 year core life (at 100% capacity factor;100,000 MWd/tonne average burnup; 1.5 peaking factor) with a core outlet temperature of 489 C, peak cladding inner surface temperature of 580 C, and steam outlet superheat of 82 C at 10 MPa.
Simulation of Axial Flow in a Bare Rod Bundle Using a Non-Linear Turbulence Model With High and Low Reynolds Approximations
Authors: de Lemos Marcelo J. S. , Assato Marcelo
Published: March 2009

This work presents a numerical investigation of fully developed turbulent flow in a triangular sub-channel of a bare rod bundle using a Non-Linear Eddy Viscosity Model (NLEVM). The numerical technique employed for discretizing the governing equations is the control-volume method with a boundary-fitted non-orthogonal coordinate system. The SIMPLE algorithm was used to correct the pressure field. The classical wall function and a low Reynolds model were used in order to handle flow calculations near the wall. In this work, the influence of constants of calibration existing in the non-linear terms of the model is analyzed.
Severe Accident Improvements for CAREM-25 to Arrest Reactor Vessel Meltdown Sequences
Authors: Pomier Ba´ez L. E. , Nun˜ez Mac Leod J. E. , Baro´n J. H.
Published: March 2009

Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The paper comprises the analyses, via numerical simulation, for the conceptual design of such a container type. Simulation model characteristics helping to establish geometrical dimensions, materials and container compatibility with power plant engineering features is addressed. The paper also presents the first model developed to analyze the complex relocation phenomena in the core of CAREM during a severe accident sequence caused by a loss of coolant. The PC version of MELCOR 1.8.4 code has been used to predict the transient behavior of core parameters. The finite element analysis (FEA) system ALGOR has been used to evaluate the thermal regime of the reactor pressure vessel wall, when the in-vessel metallic core catcher is present and when it is not present. Two different scenarios have been considered for heat transfer outside the reactor vessel, a pessimistic (dry) and optimistic (wet) conditions in the reactor cavity. This paper presents reactor variables behavior during the first hours of the event being studied, giving preliminary conclusions about the use and capability of a metallic in-vessel core catcher.
Thoria-Based Cermet Nuclear Fuel: Neutronics Fuel Design and Fuel Cycle Analysis
Authors: Downar Thomas J. , McDeavitt Sean M. , Revankar S. T. , Solomon A. A. , Kim T. K.
Published: March 2009

Cermet nuclear fuels have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle provide intriguing options for using thoria based cermet nuclear fuel in advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. The following paper will first describe the fuel thermal performance model developed for the analysis of dispersion metal matrix fuels. The model will then be applied to the design and analysis of thorium/uranium/zirconium metal matrix fuel pins for light water reactors using neutronic simulation methods.
Study on Ex-Vessel Cooling of RPV: Model Analysis of Critical Heat Flux on Inclined Plate Facing Downward and Development to Hemispherical Surface
Authors: Ohtake Hiroyasu , Koizumi Yasuo
Published: March 2009

The cooling of the Ex-vessel in Light-Water-Reactors has been proposed to maintain reactor vessel integrity during severe accident. The critical heat flux — CHF — from the underside of down-facing convex surfaces, like hemispheres, is important to the assessment of the cooling. The authors examined CHFs on inclined plates under saturated boiling experimentally, focusing on the effect of the inclination angle on the CHF and characteristic length and velocity of coalesced bubbles near the heater at the CHF. In this study, the critical heat fluxes on the inclined plates in saturated boiling were investigated analytically by using the macrolayer model and the Kelvin-Helmholtz instability, based on our previous experimental report. Furthermore, the present model was developed for the CHF on a hemispheric surface. In the present model, the most dangerous wavelength and propagative velocity of the wave in the Kelvin-Helmholtz instability for ideal fluid with vapor flow of finite thickness and surface tension on liquid-vapor interface were calculated to determine the length and the velocity of the coalesced bubble on the heating surface at CHF. The time covered over the heater with the bubble was estimated as the calculated value with the length divided by the velocity. The predictions of the present CHF model by using the macrolayer model for CHF and the Kelvin-Helmholtz instability for the characteristic values of the coalesced bubble agree well with previous experimental data for CHF on inclined plates with 30 to 180 degree in orientation. Furthermore, the present model given as a function of the inclination angle relative to the horizontal downward plate was extended to CHF on a hemispheric surface. The CHFs obtained by the present model are in qualitative agreement with experimental data on hemispheres reported by some investigators.
Thoria-Based Cermet Nuclear Fuel: Cermet Fabrication and Behavior Estimates
Authors: McDeavitt Sean M. , Downar T. J. , Solomon A. A. , Revankar S. T. , Hash M. C. , Hebden A. S.
Published: March 2009

Cermet nuclear fuels have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle provide intriguing options for using thoria based cermet nuclear fuel in advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. Critical elements in the demonstration of this new fuel form include developing low-cost fabrication methods and characterizing the cermet properties and important performance parameters. A powder-in-tube drawing and heat treatment process is being evaluated as an alternative to hot extrusion. In this method, zirconium metal and ceramic microspheres are mixed, poured into a Zircaloy shell, and compacted into simulated fuel pins. Important processing variables being evaluated include the amount of compaction required to achieve a desired matrix density and the inter-drawing thermal treatment temperature required to achieve adequate matrix fusion and grain growth.
A Study on the Embedment Effect in the Soil-Structure Interaction Analysis of the APR1400
Authors: Jang Young-Sun , Joo Kwang-Ho , Kim Chong-Hak
Published: March 2009

The SSI (Soil-Structure Interaction) analyses are being performed for the APR1400 (Advanced Power Reactor 1400MWe, Old name - KNGR ; Korean Next Generation Reactor) design, because the APR1400 is developed as a Standard Nuclear Power Plant concept enveloping suitable soil conditions. For the SSI analyses, SASSI program which adopts the Flexible Volume Method is used. In the SSI analyses, there can be uncertainties by Bond and De-bond problem between the structure and lateral soil elements. According to ASCE Standard 4, one method to address this concern is to assume no connectivity between structure and lateral soil over the upper half of the embedment of 20ft (6m), whichever is less. This study is performed as a part of the parametric analyses for the APR1400 seismic analyses to address the concern of the potential embedment effect on the in-structure response spectra due to connectivity between structure and lateral soil. In this study, 4 model cases are analyzed to check the potential embedment effect — Full connection, 20ft no connectivity which is defined as a minimum De-bond depth of the soil in ASCE Standard 4 and 26.5ft no connectivity between structure and lateral soil over the upper half of the embedment. Last one is full no connection for only reference. The in-structure response spectra are compared with the response spectra without considering the embedment effect.
Numerical Prediction of Secondary Flows in Complex Areas Using Concept of Local Turbulent Reynolds Number
Authors: Kriventsev Vladimir , Ohshima Hiroyuki , Yamaguchi Akira , Ninokata Hisashi
Published: March 2009

A new model of turbulence is proposed for the estimation of Reynolds stresses in turbulent fully-developed flow in a wall-bounded straight channel of an arbitrary shape. Ensemble-averaged Navier-Stokes, or Reynolds, equations are considered to be sufficient and practical enough to describe the turbulent flow in complex geometry of rod bundle array. We suggest the turbulence is a process of developing of external perturbations due to wall roughness, inlet conditions and other factors. We also assume that real flows are always affected by perturbations of any possible scale lower than the size of the channel. Thus, turbulence can be modeled in the form of internal or “turbulent” viscosity. The main idea of a Multi-Scale Viscosity (MSV) model can be expressed in the following phenomenological rule: A local deformation of axial velocity can generate the turbulence with the intensity that keeps the value of the local turbulent Reynolds number below some critical one. Therefore, in MSV, the only empirical parameter is the critical Reynolds number. From analysis of dimensions, some physical explanations of Reynolds number are possible. We can define the local turbulent Reynolds number in two ways: i) simply as Re = ul/v, where u is a local velocity deformation within the local scale l and v is total accumulated molecular and turbulent viscosity of all scales lower then 1. ii) Re = K/W, where K is kinetic energy and W is work of friction/dissipation forces. Both definitions above have been implemented in the calculation of samples of basic fully-developed turbulent flows in straight channels such as a circular tube and annular channel. MSV has been also applied to prediction of turbulence-driven secondary flow in elementary cell of the infinitive hexagonal rod array. It is known that the nature of these turbulence-driven motions is originated in anisotropy of turbulence structure. Due to the lack of experimental data up to date, numerical analysis seems to be the only way to estimate intensity of the secondary flows in hexagonal fuel assemblies of fast breeder reactors (FBR). Since MSV can naturally predict turbulent viscosity anisotropy in directions normal and parallel to the wall, it is capable to calculate secondary flows in the cross-section of the rod bundle. Calculations have shown that maximal intensity of secondary flow is about 1% of the mean axial velocity for the low-Re flows (Re = 8170), while for higher Reynolds number (Re = 160,100) the intensity of secondary flow is as negligible as 0.2%.
Development of an Inertia-Increased ABWR Internal Pump
Authors: Takahashi Shirou , Umemori Kousei , Shiina Kooji , Totani Tetsuya , Sakashita Akihiro , Yamashita Norimichi , Kondo Takahisa
Published: March 2009

It is possible to simplify the reactor internal pump power supply system in the ABWR without affecting the core flow supply when a trip of all RIPs event occurs by eliminating the motor-generator sets and increasing the rotating inertia of the RIPs. This inertia increase due to an additional flywheel, which leads to a gain in weight and length, requires a larger diameter nozzle with a thicker sleeve. However, a thicker sleeve nozzle and a longer and heavier motor casing may change the RIP performance. In the present study, the inertia-increased RIP was verified through full-scale tests. The rotating inertia time constant for coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP was doubled compared with the current RIP. The inertia-increased RIP with the thicker sleeve nozzle maintained good performance and its power supply system without motor-generator sets was judged appropriate for the ABWR.
Current Design Status of Sodium Cooled Super-Safe, Small and Simple Reactor
Authors: Ueda N. , Kinoshita I. , Nishi Y. , Minato A. , Yokoyama T. , Nishiguchi Y.
Published: March 2009

CRIEPI has been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. And a conceptual design of 4S (Super-Safe, Small and Simple) reactor was proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void coefficient are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core life time is more than 10 years; (5) Its construction, maintenance and operation are expected to be very simple by eliminating active components from inside of a reactor vessel. The 4S reactor is a sodium cooled fast reactor and its reactivity is not controlled by neutron absorber rods but by neutron reflectors. An electrical output is 50 MW. This paper describes a design modification to enhance the feasibility from the previous 4S design. A core active height can be shortened to 1.5 m from 4.0m to keep the reactivity characteristics. An averaged fuel burn-up is up to 70 GWD/ton and a pressure drop at the core region is less than 0.1 MPa. A reactivity control system is modified according with the core design change. As for the steam generator design, sodium-water reaction accidents must be taken into account as a design basis event for the utilization of the secondary sodium coolant. Therefore, a modified plate type heat exchanger is proposed as a steam generator. It may be possible to develop a compact steam generator, which is free from sodium-water reaction accidents and to eliminate the secondary sodium systems. The 4S reactor without secondary system has been proposed as a candidate design.
Effect of Thermal Stratification of Coolant in a Vertical Heated Channel
Authors: Zare Shahneh Abolghasem
Published: March 2009

In a vessel type low power research reactor having vertical fuel plates, while circulating pump is switched off, coolant (light water) would flow by natural convection. By using conservation equations, taking into account simplifying assumptions, coolant mass flow rate through the channel can be obtained. Due to the thermal stratification effect, coolant mass flow rate through the channel is shown to decrease. The present study shows that, assuming a linear thermal stratification, the variations of coolant mass flow rate versus stratification parameter behave in a non-linear manner. The aforementioned variations decrease down to 41%.
Simulation of the HDR E11.2 Hydrogen Mixing Experiment With the CONTAIN Code
Authors: Skerlavaj Aljaz , Kljenak Ivo
Published: March 2009

One of the most well-known experiments on atmosphere stratification in a nuclear power plant containment at severe accident conditions is the test E11.2 “Hydrogen distribution in loop flow geometry”, which was performed in the Heissdampf Reaktor containment test facility in Germany. In the present work, the simulation of the test E11.2 with the CONTAIN computer code is presented. An input model consisting of 72 cells and 263 flowpaths was developed. The predicted pressure history and thermal stratification agree relatively well with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was somewhat underestimated.
Design Characteristics of Ulchin Nuclear Power Units 5&6
Authors: Lee Deucksoo , Kim Dong-Su , Lee Young-Taik , Kwon O-Keol , Kim Jung-Cha
Published: March 2009

Ulchin nuclear power plant units 5&6 (UCN 5&6), which started excavation on January 1999, are two loop pressurized water reactors (PWR) with the capacity of 1000 MWe, and planned to start commercial operation on June, 2004 and June, 2005, respectively. The reactor coolant system of the UCN 5&6 consist of a reactor vessel, internals, and two steam generators, four reactor coolant pumps, a pressurizer and primary piping. Based on the system design of the first Korean Standard Nuclear Power Plant (KSNP), UCN 5&6 is designed to provide improvements in safety, reliability and cost by applying both advanced proven technology and experiences gained from the construction and operation of the previous KSNP plants. The result of the preliminary probabilistic safety assessment study for UCN 5&6 shows that the core damage frequency is lowered significantly. Several design improvement items have been adopted to the system design and contributed to lower the core damage frequency value. Among the design improvements, digital PPS and digital ESFAS are the key to the UCN 5&6 design. Furthermore, digitization of the Plant Protection System (PPS) and Engineered Safety Feature Actuation System (ESFAS) for the PWR is the first case in the PWR construction history. The Korean regulatory body reviewed the design concept of the digital PPS and digital ESFAS, and evaluated to be acceptable for the plant safety. Also, in-depth review on the detail design of the digital PPS/ESFAS and the special evaluation/audit for the software design process are underway to secure the software quality. The safety of the UCN 5&6 design has been evaluated through a two-year review on the preliminary safety analysis report. As a result, the construction permit was issued on May 17, 1999 by the government. In this paper, design characteristics of UCN 5&6 are discussed focussed on design improvements comparing with KSNR. And, some of the safety analysis results are presented as well as licensing status.