bluuebunny/crossref_metadata_2025 huggingface dataset:
Experimental Investigation of Coolant Mixing in the RPV of PWR in the Late Phase of a SBLOCA Event
Authors: Kliem So¨ren , Prasser Horst-Michael , Su¨hnel Tobias , Weiss Frank-Peter , Hansen Asmus
Published: September 2008

Partial depletion of the primary circuit of a pressurized water reactor during a postulated small break loss of coolant accident can lead to interruption of one-phase flow natural circulation. In this case, the decay heat is removed from the core in the reflux-condenser mode. In this operation mode, slugs of lower borated water can accumulate in the cold legs. After refilling of the primary circuit, the natural circulation in the two loops not receiving emergency core cooling injection (ECC) re-establishes and the lower borated slugs are shifted towards the reactor pressure vessel (RPV). Entering the core, the lower borated water causes a reactivity insertion. Mixing inside the RPV is an important phenomenon limiting the reactivity insertion and preventing a re-criticality. The mixing of these lower borated slugs with the ambient coolant in the RPV was investigated at the 1:5 scaled coolant mixing test facility ROCOM. Wire mesh sensors based on electrical conductivity measurement are used in ROCOM to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a sensor which spans a measuring grid of 64 azimuthal and 32 positions over the height. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly. The boundary conditions for the mixing experiment were taken from an experiment at the thermal-hydraulic test facility PKL operated by FANP Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after shifting into the RPV. The ECC-water injected into the RPV falls almost straight down through the lower borated water and accelerates. On the outer sides of the ECC-streak, lower borated coolant admixes and flows together with the ECC-water downwards. This is the only mechanism of transporting the lower borated water into the lower plenum. All these effects could be visualized and quantified by the downcomer sensor. On the way to the core, the lower borated water is effectively mixing with the ambient, high borated water. Therefore, in the core inlet plane, lower borated water is detected only in the outer part. The minimum boron concentration, measured at one fuel element inlet position at one certain time point, was 71% of the initial 2500 ppm. There is no change of the initial boron concentration in the inner part of the core inlet plane during the whole transient at all.
PBMR as an Ideal Heat Source for High-Temperature Process Heat Applications
Authors: Correia Michael , Greyvenstein Renee´ , Silady Fred , Penfield Scott
Published: September 2008

The Pebble Bed Modular Reactor (PBMR) is an advanced helium-cooled, graphite-moderated High Temperature Gas-cooled Reactor (HTGR). A 400 MWt PBMR Demonstration Power Plant (DPP) for the production of electricity is being developed in South Africa. This PBMR technology is also an ideal heat source for process heat applications, including Steam Methane Reforming, steam for Oil Sands bitumen recovery, Hydrogen Production and co-generation (process heat and/or electricity and/or process steam) for petrochemical industries. The cycle configuration used to transport the heat of the reactor to the process plant or to convert the reactor’s heat into electricity or steam directly influences the cycle efficiency and plant economics. The choice of cycle configuration depends on the process requirements and is influenced by practical considerations, component and material limitations, maintenance, controllability, safety, performance, risk and cost. This paper provides an overview of the use of a PBMR reactor for process applications and possible cycle configurations are presented for applications which require high temperature process heat and/or electricity.
The Human Bathtub: Safety and Risk Predictions Including the Dynamic Probability of Operator Errors
Authors: Duffey Romney B. , Saull John W.
Published: September 2008

Reactor safety and risk are dominated by the potential and major contribution for human error in the design, operation, control, management, regulation and maintenance of the plant, and hence to all accidents. Given the possibility of accidents and errors, now we need to determine the outcome (error) probability, or the chance of failure. Conventionally, reliability engineering is associated with the failure rate of components, or systems, or mechanisms, not of human beings in and interacting with a technological system. The probability of failure requires a prior knowledge of the total number of outcomes, which for any predictive purposes we do not know or have. Analysis of failure rates due to human error and the rate of learning allow a new determination of the dynamic human error rate in technological systems, consistent with and derived from the available world data. The basis for the analysis is the “learning hypothesis” that humans learn from experience, and consequently the accumulated experience defines the failure rate. A new “best” equation has been derived for the human error, outcome or failure rate, which allows for calculation and prediction of the probability of human error. We also provide comparisons to the empirical Weibull parameter fitting used in and by conventional reliability engineering and probabilistic safety analysis methods. These new analyses show that arbitrary Weibull fitting parameters and typical empirical hazard function techniques cannot be used to predict the dynamics of human errors and outcomes in the presence of learning. Comparisons of these new insights show agreement with human error data from the world’s commercial airlines, the two shuttle failures, and from nuclear plant operator actions and transient control behavior observed in transients in both plants and simulators. The results demonstrate that the human error probability (HEP) is dynamic, and that it may be predicted using the learning hypothesis and the minimum failure rate, and can be utilized for probabilistic risk analysis purposes.
Ignalina NPP Unit 2 Gas Gap Probabilistic Evaluation
Authors: Simaityte Volskiene Jurgita , Augutis Juozas , Uspuras Eugenijus
Published: September 2008

One of the factors limiting RMBK service life is the closure of the gas gap between the pressure tubes and the graphite masonry. The objective of the present work is to develop a mathematical model for Unit 2 of the Ignalina nuclear power plant (NPP) reactor, in Lithuania, which estimates the probability of the presence of a gas gap for different periods of operation. The calculations are based on collected data of graphite and pressure tube diameters and burn up measurements in period 1983–2005. There were no performed enough measurements for Unit 2, but big amount of statistical data was available for Unit 1. Therefore Unit 1 data properties were analyzed and constructed model was applied for INPP reactor Unit 2. The main analysis results are gas gap existence probabilities evaluation and the issued channel measuring strategies for the next planed preventive maintenance period for Ignalina NPP.
Parameterization of Buoyancy Effects in Generic PWR Boron Dilution Scenarios
Authors: Galindo-Garci´a Iva´n F. , Cotton Mark A. , Axcell Brian P.
Published: September 2008

A computational investigation is undertaken into the role of buoyancy in a PWR boron dilution transient following a postulated Small Break Loss of Coolant Accident (SB-LOCA). In the scenario envisaged there is flow of de-borated and relatively high temperature water from a single cold leg into the downcomer; flow rates are typical of natural circulation conditions. The study focuses upon the development of boron concentration distributions in the downcomer and adopts a 3D-unsteady formulation of the mean flow equations in combination with the standard high-Reynolds-number k-ε turbulence model. It is found that the Richardson number (Ri = Gr/Re2) is the most important group parameterizing the course of a concentration transient. At Ri values characterizing a ‘baseline’ scenario the results indicate that there is a stable, circumferentially-uniform, descent through the downcomer of a stratified region of low-borated fluid. Qualitatively the same behaviour is found at higher Richardson number, although at Ri values of approximately one-fifth the baseline level there is evidence of large-scale mixing and a consequent absence of concentration stratification.
Heat Transfer to Supercritical Water in Gaseous State or Affected by Mixed Convection in Vertical Tubes
Authors: Pis’menny E. N. , Razumovskiy V. G. , Maevskiy E. M. , Koloskov A. E. , Pioro I. L.
Published: September 2008

The results on heat transfer to supercritical water heated above the pseudocritical temperature or affected by mixed convection flowing upward and downward in vertical tubes of 6.28-mm and 9.50-mm inside diameter are presented. Supercritical water heat-transfer data were obtained at a pressure of 23.5 MPa, mass flux within the range from 250 to 2200 kg/(m2s), inlet temperature from 100 to 415°C and heat flux up to 3.2 MW/m2. Temperature regimes of the tubes cooled with supercritical water in a gaseous state (i.e., supercritical water at temperatures beyond the pseudocritical temperature) were stable and easily reproducible within a wide range of mass and heat fluxes. An analysis of the heat-transfer data for upward and downward flows enabled to determine a range of Gr/Re2 values corresponding to the maximum effect of free convection on the heat transfer. It was shown that: 1) the heat transfer coefficient at the downward flow of water can be higher by about 50% compared to that of the upward flow; and 2) the deteriorated heat-transfer regime is affected with the flow direction, i.e., at the same operating conditions, the deteriorated heat transfer may be delayed at the downward flow compared to that at the upward flow. These heat-transfer data are applicable as the reference dataset for future comparison with bundle data.
Corrosion Resistance of Active Coating Materials Exposed to Lead-Alloy
Authors: Cho Choonho , Song Tae-Yung , Cho Chungho
Published: September 2008

A key problem in the development of heavy liquid metal cooled reactors is a corrosion of the structural and fuel cladding materials by the coolants. Thus, the problem has been considered as an important design-factor that limits the operational temperature and flow velocity of the next generation nuclear reactors using lead-alloys. Corrosion data has been obtained on as-received and active coating materials of HT9 and 316L in a stagnant lead-alloy containing a reduced atmosphere of oxygen with an exposure time of 1500 hours at 600°C. After each test, the specimens were analyzed metallurgically by using a scanning electron microscopy (SEM) with a energy dispersive X-ray analysis (EDX) for the cross sections of the specimens. In addition, X-ray diffraction (XRD) was performed to evaluate the phase composition of the steels.
On the Factors Affecting the Fretting-Wear Risk of PWR Fuel Assemblies
Authors: Rubiolo Pablo R. , Young Michael Y.
Published: September 2008

An analysis of the factors affecting the fretting-wear risk of fuel assemblies operated in Pressurized Water Reactors (PWR) is presented. In this work the effect of rod-to-grid gap size, spring relaxation, assembly grids misalignments, rod structural damping and stiffness, and friction coefficients are investigated by performing Monte Carlo simulations with a non-linear vibration model of the fuel rod. The goal of the study is to identify key factors that have to be included in the assessment of the fretting-wear performance of core fuel assemblies.
CFD Analyses of Damaged Fuel Inside a Cleaning Vessel
Authors: Le´gra´di Ga´bor , Boros Ildiko´ , Aszo´di Attila
Published: September 2008

On 10–11th of April, 2003, a serious incident occurred in a special fuel assembly cleaning tank, which was installed into the service shaft of the 2nd unit of the Paks NPP in Hungary. During this incident, most of the 30 fuel assemblies put into the cleaning tank have seriously damaged. In the Institute of Nuclear Techniques of the Budapest University of Technology and Economics several CFD investigations were performed concerning the course of the incident, the post incidental conditions and the recovery work. The main reason of the incident can be originated from the defective design of the cleaning tank which resulted in the insufficient cooling of the system in a special operational mode. Our investigation performed with a complex 3D CFX model clearly showed how could as strong temperature stratification develop inside the cleaning tank that it was able to block the coolant flow through the fuel assemblies. After the blocking of the flow, the coolant turned into boiling and the assemblies became uncovered. The temperature of the surfaces of the fuel assemblies went above 1000 °C. With the aid of the radiative heat transfer model of the CFX-5.6 code, the surface temperatures were analyzed. When the cleaning instrument got opened the fuel assemblies suffered a serious thermal shock and the assemblies highly damaged. The post-incident thermo-hydraulic state inside the cleaning vessel was investigated with a rather complex CFX model. The uncertainties were decreased by a wide parameter study. The recovery work is planned to be started in the close future. The operators of the damaged fuel removing equipments will work standing on a platform which will be placed into the service shaft just above the surface of the coolant of decreased level. Protecting the workers against unnecessary personal doses is a very important task. In this situation, while the coolant is important part of the biological shielding, it is also a source of radiation due to the considerable amount of radioactive contamination dispersed into it. Therefore, the 3D distribution of the contamination in the service shaft was estimated for different operational and incidental scenarios with a wide parameter study made by a 3D CFX model. This comprehensive work performed with several models and calculations is tersely outlined according to the limited extent of the paper.
Advanced Neutronics Tools for BWR Design Calculations
Authors: Santamarina A. , Hfaiedh N. , Letellier R. , Marotte V. , Misu S. , Sargeni A. , Vaglio C. , Zmijarevic I.
Published: September 2008

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific “2005–2007 BWR Plan” settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy.
Steady State Film Boiling Heat Transfer Simulated With TRACE V4.160
Authors: Jasiulevicius Audrius , Macian-Juan Rafael
Published: September 2008

This paper presents the results of the assessment and analysis of TRACE v4.160 heat transfer predictions in the post-CHF (critical heat flux) region and discusses the possibilities to improve the TRACE v4.160 code predictions in the film boiling heat transfer when applying different film boiling correlations. For this purpose, the TRACE v4.160-calculated film boiling heat flux and the resulting maximum inner wall temperatures during film boiling in single tubes were compared with experimental data obtained at the Royal Institute of Technology (KTH) in Stockholm, Sweden. The experimental database included measurements for pressures ranging from 30 to 200 bar and coolant mass fluxes from 500 to 3000 kg/m2s. It was found that TRACE v4.160 does not produce correct predictions of the film boiling heat flux, and consequently of the maximum inner wall temperature in the test section, under the wide range of conditions documented in the KTH experiments. In particular, it was found that the standard TRACE v4.160 underpredicts the film boiling heat transfer coefficient at low pressure-low mass flux and high pressure-high mass flux conditions. For most of the rest of the investigated range of parameters, TRACE v4.160 overpredicts the film boiling heat transfer coefficient, which can lead to non-conservative predictions in applications to nuclear power plant analyses. Since no satisfactory agreement with the experimental database was obtained with the standard TRACE v4.160 film boiling heat transfer correlations, we have added seven film boiling correlations to TRACE v4.160 in order to investigate the possibility to improve the code predictions for the conditions similar to the KTH tests. The film boiling correlations were selected among the most commonly used film boiling correlations found in the open literature, namely Groeneveld 5.7, Bishop (2 correlations), Tong, Konkov, Miropolskii and Groeneveld-Delorme correlations. The only correlation among the investigated, which resulted in a significant improvement of TRACE predictions, was the Groeneveld 5.7. It was found, that replacing the current film boiling correlation (Dougall-Rohsenow) for the wall-togas heat transfer with Groeneveld 5.7 improves the code predictions for the film boiling heat transfer at high qualities in single tubes in the entire range of pressure and coolant mass flux considered.
Fragmentation of a Single and Continuous Molten Copper Droplets Penetrating a Sodium Pool
Authors: Zhang Zhi-Gang , Sugiyama Ken-Ichiro , Narabayashi Tadashi , Nishimura Satoshi , Kinosita Izumi
Published: September 2008

The progression of hypothetical core disruptive accidents in metal fuel cores is strongly affected by exclusion of molten metal fuel from the core region due to molten fuel-coolant interaction (FCI). As a basic study of FCI in metal fuel fast reactors, the present paper focuses on the fragmentation and the characteristics of the debris produced during a series of experiments of a single molten copper droplet from 1g to 5g and continuous droplets with a total mass from 16g to 26g, which penetrated a sodium pool at instantaneous contact interface temperatures from 1005 °C far below the freezing point: 1084°C to 1342°C. The results show that the mass median diameters (Dm) of different mass copper droplets both a single droplet and continuous ones penetrating a sodium pool differ very little, nearly the same when the instantaneous contact interface temperatures (Ti) are above the freezing point, and also the droplets are fragmented finely with their increasing Tis; but when Tis are below the freezing point, the Dms of different mass copper droplets scatter a little widely. These results basically show the fragmentation of molten fuel, which is important to assure the termination of accidents, is promising in the sodium space in the upper and lower plenums.
Application of CFD Model for Inlet Flow Region of 17×17 Fuel Assembly
Authors: Dzodzo Milorad B. , Liu Bin , Rubiolo Pablo R. , Karoutas Zeses E. , Young Michael Y.
Published: September 2008

A numerical investigation was performed to study the variation in axial and lateral velocity profiles occurring downstream of the inlet nozzle of a typical Westinghouse 17×17 PWR fuel assembly. A Computational Fluid Dynamic (CFD) model was developed with commercial CFD software. The model comprised the lower region of the fuel assembly, including: the Debris Filter Bottom Nozzle (DFBN), P-grid, Bottom Inconel grid, one and half grid span, as well as the lower core plate hole. The purpose of the study was to obtain insight into the flow redistribution resulting from the interaction of the jet arising from the lower core plate hole and the fuel assembly structure. In particular the axial and lateral velocities before and after the nozzle were studied. The results, axial and lateral velocity contours, streamlines and maximum axial and lateral velocity distributions at various elevations are presented and discussed in relation to the potential risk of high turbulent excitation over the rod and the resulting rod-to-grid fretting-wear damage. The CFD model results indicated that the large jet flows from the lower core plate are effectively dissipated by DFBN nozzle and the grids components of the fuel assembly. The breakup of the large jets in the DFBN and the lower grids helps to reduce the steep velocity gradients and thus the rod vibration and fretting-wear risk in the lower part of the fuel assembly. The presented CFD model is one step towards developing advanced tools that can be used to confirm and evaluate the effect of complex PWR structures on flow distribution. In the future the presented model could be integrated in a larger CFD model involving several fuel assemblies for evaluating the lateral velocities generated due to the non-uniform inlet conditions into the various fuel assemblies.
Neutron Transport Characteristics of a Nuclear Reactor Based Dynamic Neutron Imaging System
Authors: Khaial Anas M. , Harvel Glenn D. , Chang Jen-Shih
Published: September 2008

An advanced dynamic neutron imaging system has been constructed in the McMaster Nuclear Reactor (MNR) for nondestructive testing and multi-phase flow studies in energy and environmental applications. A high quality neutron beam is required with a thermal neutron flux greater than 5.0×106 n/cm2-s and a collimation ratio of 120 at image plane to promote high-speed neutron imaging up to 2000 frames per second. Neutron source strength and neutron transport have been experimentally and numerically investigated. Neutron source strength at the beam tube entrance was evaluated experimentally by measuring the thermal and fast neutron fluxes, and simple analytical neutron transport calculations were performed based upon these measured neutron fluxes to predict facility components in accordance with high-speed dynamic neutron imaging and operation safety requirements. Monte-Carlo simulations (using MCNP-4B code) with multiple neutron energy groups have also been used to validate neutron beam parameters and to ensure shielding capabilities of facility shutter and cave walls. Neutron flux distributions at the image plane and the neutron beam characteristics were experimentally measured by irradiating a two-dimensional array of Copper foils and using a real-time neutron radiography system. The neutron image characteristics — such as neutron flux, image size, beam quality — measured experimentally and predicted numerically for beam tube, beam shutter and radiography cave are compared and discussed in detail in this paper. The experimental results show that thermal neutron flux at image plane is nearly uniform over an imaging area of 20.0-cm diameter and its magnitude ranges from 8.0×106 – 1.0×107 n/cm2-sec while the neutron-to-gamma ratio is 6.0×105 n/cm2-μSv.
BWR Fuel Thermomechanical Evaluation for Preconditioning Procedures With FEMAXI-V
Authors: Herna´ndez-Lo´pez He´ctor , Lucatero Marco A. , Ortiz-Villafuerte Javier
Published: September 2008

Burnup limitations are normally set to limit stresses in the fuel assembly components. The defined limits provide guidance to the fuel designer to minimize fuel failure during steady sate operation, and also prevent against some thermal and mechanical phenomena that could occur during overpower transients. In particular, a LHGR limit value is set to take into account physical phenomena that could lead to pellet-cladding interaction. This limit value directly relates to a PCI limit, which may be set based on experimental ramp tests. Thus, to avoid violating the PCI limit, fuel conditioning procedures are still required for both barrier and non-barrier fuel. Simulation of the power ramp procedures to be performed by the reactor operator during startup or power increase maneuvers is advisable as a preventive measure of possible overpower consequences on the fuel thermomechanical behavior. In this paper, the thermomechanical behavior of two different kinds of BWR fuel rods is analyzed for fuel preconditioning procedures. Five different preconditioning computations were performed, each with three different ascending linear power rate ramps. The starting point of the ramps was taken from data of the Cycle 8 of the Unit 1 of the Laguna Verde Nuclear Power Plant, located in MEXICO. The top limit of the ramps was the threshold linear power at which failure by PCI could occur, as a function of burnup. The analysis was performed with the FEMAXI-V code.
Evaluation of the Sensitivity of a Two-Phase Flow Model for the Steam Separators Analysis
Authors: Chaki Masao , Murase Michio
Published: September 2008

Reducing of the pressure losses of steam separator systems of boiling water reactor (BWR) plants is useful to reduce the required pump head and enhance core stability design margin. The need to reduce the pressure losses of steam separator systems is especially important in BWR plants that have high power density cores and natural circulation systems. The core flow rate of a BWR plant with a natural circulation system is affected by the pressure losses of steam separator systems. In BWR plants with high power density cores, the core stability design margin is affected by these pressure losses. Generally, reducing the pressure losses of the steam separator systems leads to increased carry-under and carryover. Reducing the pressure losses while keeping the characteristics of both carry-under and carryover is desired, so many studies have been done. The steam separator of a BWR plant consists of a standpipe section, a swirl vane section and three-barrel sections. Two-phase flow of steam and water enters the steam separator through the standpipe section and reaches the swirl vane section. In the swirl vane section, the two-phase flow is given centrifugal force and is basically separated into steam and water. Therefore investigating the two-phase flow characteristics of the swirl vane section is very important. After the swirl vane section, the two-phase flow enters the barrel sections. Each barrel has a pick-off ring. The water in the barrel section is mainly removed by these pick-off rings because the water mainly flows upward as a liquid film in the barrel section due to the centrifugal force given in the swirl vane section. We researched the effect of using the drag force model of the swirling two-phase flow in analyzing a steam separator and we found that the drag force model greatly affects the results of the analysis.
Emergency Operating Procedures (EOPs) for the AP1000 Simulator
Authors: Hayashi Yuichi , Saiu Gianfranco , Wright Richard F.
Published: September 2008

The AP1000 is two-loop 1100 MWe advanced pressurized water reactor (PWR) that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 30 years of operating PWR experience. The AP1000 final design certification was approved by the NRC in December, 2005. A total of 34 Emergency Operating Procedures (EOPs) for operation of the AP1000 simulator have been prepared based on the AP1000 Emergency Response Guidelines (ERGs), background information documents and detailed plant information. These include 28 EOPs at power and 6 EOPs during shutdown. The AP1000 ERGs were developed by using the generic ERGs for the low pressure reference PWR plant as a basis. The AP1000 design differences from the reference plant were reviewed and reflected in the process of developing operational steps in each ERG. The provisions of the AP1000 PRA were also reviewed and incorporated into the ERGs. Although the AP1000 design does not require operator actions for the first 72 hours after accidents, the operator actions with both safety-related and nonsafety-related equipment have an important role to mitigate the consequence of accidents. In the event of a steam generator tube rupture (SGTR), although the AP1000 is designed so that no operator actions are required to recover from the event, there are actions that can be taken by the operator to limit the release of radioactive effluents from the ruptured SG. These actions include isolation of the ruptured SG and depressurization of the reactor coolant system (RCS) to terminate primary-to-secondary leakage, restoring reactor coolant inventory to ensure adequate core cooling and plant pressure control. It is expected that these operator actions should be incorporated into the ERG to reduce the fission product release. To support the development of the AP1000 ERGs, several transient and accident analyses were performed. These include analyses for LOCA, post-LOCA cooldown and depressurization, passive safety system termination, SGTR and faulted SG isolation. These analyses results were incorporated into the ERG background information documents. In the event of SGTR, several cases were analyzed, including consideration of operator recovery actions. These cases were modeled using the best-estimate state-of-art RELAP5 code. The analyses results show that operator recovery actions are effective for SGTR to be placed under operator control.
Expedition to the 30-km Chernobyl Exclusion Zone and the Utilization of Its Experience in Education and Communication
Authors: Aszo´di Attila , Yamaji Bogda´n , Silye Judit , Pa´zma´ndi Tama´s
Published: September 2008

Between May 28 – June 4, 2005, under the organization of the Hungarian Nuclear Society (HNS) and the Hungarian Young Generation Network (HYGN) — which operates within the framework of the HNS — a scientific expedition visited the Chernobyl Nuclear Power Plant and the surrounding exclusion zone. The participants were young Hungarian nuclear professionals supervised by more experienced experts. The main scientific goals of the expedition were the followings: • Get personal experiences in a direct way about the current status of the Chernobyl Power Plant and its surroundings, the contamination of the environment and about the doses. • Gather information about the state of the shut down power plant and the shelter built above the damaged 4th unit. • Training of young nuclear experts by performing on site measurements. The Hungarian expedition successfully achieved its objectives by performing wide-range of environmental and dosimetric measurements and collecting numerous biological and soil samples. Within the 30-km exclusion zone the influence of the accident occurred 20 years ago still could be measured clearly; however the level of the radioactivity is manageable in most places. The dosimetrical measurements showed that no considerable exposure occurred among the members of the expedition. The analysis of samples has been started at the International Chernobyl Center in Slavutich. During the expedition not only environmental sampling and in-situ measurements were carried out but it was also well documented with photos and video recordings for educational, training and PR purposes. A documentary TV film was recorded during the expedition. The first-hand knowledge acquired during the expedition helps the authentic communication of the accident and its present-day consequences, which is especially important in 2006, 20 years after the Chernobyl accident. Since Ukraine and Hungary are neighbor countries the media constantly discuss the accident, the consequences and the risks of using nuclear energy. In addition in November 2005 Hungary’s parliament approved plans to extend the lifetime of the country’s four-unit nuclear power plant. In order to have the crucial public support for nuclear energy it is very important to dispel unrealistic dismay and misbelieves regarding these questions. Thus it is extremely beneficial to have a film on this topic created by nuclear professionals especially for the public audience. In 2005 a book on the Chernobyl accident was published in Hungary that covers this expedition in a full chapter [2]. We plan to present the film to the audience of the conference.
CFD Prediction of Hydraulic Effects on External Ultrasonic Flowmeters: A Case Study
Authors: Augenstein Don , Estrada Herb , Hauser Ernie , Madera Ed , Keijers Steven , Bertels Frank , Hoefnagels Rudi , Gorleer Eric , Beirnaert Dirk , Goorden Peter
Published: September 2008

External ultrasonic flow meters installed in a nuclear power plant feedwater system were suspected of containing a conservative bias (high reading) despite having been calibrated in a scale model laboratory test. Several independent measures of plant power level indicated that the bias could be as high as 2.0%. The first step in resolving the discrepancy was the construction of a computational fluid dynamics (CFD) model of the hydraulic configurations in which the flow meters were tested in the lab. The CFD model closely matched the original calibration test results. The CFD model for real plant geometry including certain upstream hydraulic features predicted calibration changes of the same sign and order of magnitude as the suspected bias. These upstream elements were absent in the original hydraulic model test. As a consequence of this finding, these additional features were included in new hydraulic calibration tests. Some effects predicted by the CFD model were verified and traceably confirmed, while others were contradicted by the data from the new model. Final calibration test results confirmed the presence of a bias, although not of the magnitude originally suspected. The calibration results provide a traceable measurement basis for use of the flow meters within their specified accuracy after correction for the bias. By use of a chordal ultrasonic flowmeter, velocity profile (pointed-ness) and swirl rates were measured during the tests. The tests, together with the velocity profile information from the CFD simulations, provided insights into the errors and omissions in the original calibration model tests and the assumptions on which the model was based. Finally the calibration tests highlight the strengths and limitations of CFD models for work of this kind.
Scoping System Analysis of KALIMER-600 Design Concept
Authors: Kwon Young-Min , Jeong Hae-Yong , Ha Ki-Seok , Chang Won-Pyo , Lee Yong-Bum , Hahn Dohee
Published: September 2008

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced LIquid Metal Reactor), which is a sodium-cooled, metallic-fueled, pool-type reactor. The KALIMER-600 design concept (600 MWt) was selected as one of the reference GEN-IV sodium-cooled fast reactors (SFRs). The safety design philosophy of KALIMER-600 places maximum reliance on passive responses to abnormal and emergency conditions, and minimizes the need for active and engineered safety systems. KALIMER-600 utilizes the intrinsic negative reactivity feedback effect under design extended conditions where reactor scram failures are postulated. In order to assess the effectiveness of the inherent safety features, a scoping system analysis during unprotected overpower, loss of flow and under cooling events has been performed using the system-wide transient analysis code SSC-K. The present scoping analysis focuses on identification of enhanced safety design features that provide passive and self-regulating response to transient conditions and evaluation of safety margins. The results of the scoping analysis indicate an understanding of various inherent reactivity feedback mechanisms is very important in establishing design features. The analysis results show that the KALIMER-600 design concepts provide larger safety margins with respect to sodium boiling, fuel rod integrity, and structural integrity. The inherent safety can be enhanced by accounting for reactivity feedback mechanisms in the design process.
Structure-Property Relationships in Polymer Derived Amorphous/Nano-Crystalline Silicon Carbide for Nuclear Applications
Authors: Zunjarrao Suraj C. , Singh Abhishek K. , Singh Raman P.
Published: September 2008

Silicon carbide (SiC) is a promising candidate for several applications in nuclear reactors owing to its high thermal conductivity, high melting temperature, good chemical stability, and resistance to swelling under heavy ion bombardment. However, fabricating SiC by traditional powder processing route generally requires very high temperatures for pressureless sintering. Polymer derived ceramic materials offer unique advantages such as ability to fabricate net shaped components, incorporate reinforcements and relatively low processing temperatures. Furthermore, for SiC based ceramics fabricated using polymer infiltration process (PIP), the microstructure can be tailored by controlling the processing parameters, to get an amorphous, nanocrystalline or crystalline SiC. In this work, fabrication of polymer derived amorphous and nano-grained SiC is presented and its application as an in-core material is explored. Monolithic SiC samples are fabricated by controlled pyrolysis of allylhydridopolycarbosilane (AHPCS) under inert atmosphere. Chemical changes, phase transformations and microstructural changes occurring during the pyrolysis process are studied as a function of the processing temperature. Polymer cross-linking and polymer to ceramic conversion is studied using infrared spectroscopy (FTIR). Thermogravimetric analysis (TGA) and differential thermal analysis (DTA) are performed to monitor the mass loss and phase change as a function of temperature. X-ray diffraction studies are done to study the intermediate phases and microstructural changes. Variation in density is carefully monitored as a function of processing temperature. Owing to shrinkage and gas evolution during pyrolysis, precursor derived ceramics are inherently porous and composite fabrication typically involves repeated cycles of polymer reinfiltration and pyrolysis. However, there is a limit to the densification that can be achieved by this method and porosity in the final materials presents difficulties in interpreting “true” properties from bulk measurements. Hence, hardness and modulus measurements are carried out using instrumented nanoindentation to establish property–structure relationship for SiC derived from the polymer precursor. It is seen that the presence of nanocrystalline domains in amorphous SiC significantly influences the modulus and hardness.
Environmental Qualification of an Actuator Torque Switch
Authors: Chan A. M. , Barreca S. L. , Kostela T.
Published: September 2008

Environmental qualification testing was performed on a modified Limitorque torque switch for the torque switch safety functions in the Limitorque type SMB actuators located inside and outside containment in a nuclear power plant. The torque switch specimen was installed in a Limitorque SMB-1 electric actuator mounted on an 8” Velan gate valve and operated with a customized programmable logic controller to allow normal torque switch behaviour to be observed. The present paper describes the qualification testing performed. The modified torque switch was aged to a 30-year service life at the normal service conditions for both inside and outside containment. Aging included radiation, thermal and cycle aging. A seismic test and then a combined Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) steam accident simulation were followed. After each stage of aging, functional tests were done to confirm normal insulation resistance, normal contact resistance and normal operation.
Water Hammer Analysis of an Emergency Cooling System With the Presence of Air
Authors: Malanca Rodrigo , Delmastro Dari´o , Garci´a Juan Carlos
Published: September 2008

In this paper the effect of air presence in an emergency cooling system of a nuclear power plant during the pipelines pressurization stage is analyzed. The pressure waves propagation along the system pipelines after a fast pressurization of the water tanks is considered. The presence of different amounts of air in one pipe is analyzed. A code was developed to simulate the pressure waves propagation in the system. This code uses the classical Method of Characteristics (MOC) solving the mass and momentum equations. The two phase homogeneous model was used to represent the two-phase mixture in the pipe with presence of air. The maximal pressure in each pipe is calculated for the presence of different amounts of air in one pipe and a time ramp of 0.50 seconds for the operations of valves. The influence of the presence of air in the system is shown.
An Assisted Flow Circulation Option for Integral Pressure Water Reactors
Authors: Delmastro Dario F. , Patruno Luciano , Masson Viviana
Published: September 2008

In this paper the use of jet pumps for assissting the coolant circulation in a integral pressure water reactor is considered. The integral pressure water reactors are characterized by the presence of all high pressure and temperature components inside a single pressure vessel. In this way the core, steam generators, pressurizer and pumps are located inside the same pressure vessel. This kind of systems usually consider natural circulation for the low power modules and forced circulation for the high power ones. CAREM-25 is the prototype reactor of CAREM design. It is a 100 MW thermal power integral pressure water reactor that use natural circulation in the primary circuit. The possibility, advantages and constraints, of using jet pumps for assissting the primary flow circulation and increase the reactor power are analyzed.
AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis
Authors: Wright Richard F. , Schwall James R. , Taylor Creed , Karim Naeem U. , Thakkar Jivan G. , Schulz Terry
Published: September 2008

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.
PWR Fuel Assembly Damping Characteristics
Authors: Lu Roger Y. , Steel David D.
Published: September 2008

PWR fuel assembly damping is a key parameter in seismic/LOCA safety analysis. The damping coefficients of a fuel assembly in air, still water and flowing water are significantly different. Several researchers and engineers have published their results and methods in the past. With this paper, PWR fuel assembly damping was studied and tested in air, still water, and flowing water (including flowrate and temperature variation). The damping coefficients were obtained by the initial displacement and first response method. The coefficients are also compared with published data. Several conclusions are obtained. • The damping obtained from the tests in air gives the damping component of assembly structure damping. From the comparison of the damping in air with still water the amount of viscous damping can be determined. The viscous damping component is the effect of still water on damping. The amount of viscous damping is represented by the increase in the damping ratio from air to still water at room temperature. The results show that damping in still water is approximately two times the damping in air. • The temperature effect on damping in still water is minimal. In flowing water, the results show a very slight effect of temperature, as the damping slightly decreases with an increase in temperature. This temperature effect is much smaller than the data scatter observed in most damping measurement tests under the same test conditions. • The damping is significantly affected by flowing water. For relatively low flow velocities, compared to in-core conditions, the damping coefficient is around two times the damping in still water. For intermediate to high flow velocities, all damping coefficients are 2.5 times higher than that in still water. For high velocities and large displacement, the damping coefficient can be over 3 times higher than that in still water. The flow velocity appears to be acting on the system by suppressing the motion of the assembly. Additional damping due to flowing water is called hydraulic damping, which is generated by hydraulic force. When a fuel assembly vibrates in flowing water, the assembly is trying to change the flow direction and momentum, but the flow mass wants to retain its pure axial direction which suppresses the motion of the assembly.