bluuebunny/crossref_metadata_2025 huggingface dataset:
First Results of the Phebus FPT3 Test
Authors: March Philippe , Biard Bruno , Manenc Christelle , Payot Fre´de´ric , Gaillard Claude , Guillot Je´roˆme , Janot Michel , Morassano Bruno , Pepino Michel , Simondi-Teisseire Be´atrice , Garnier Yves , Gre´goire Gilles , Masson Ve´ronique , Pantera Laurent
Published: September 2008

The purpose of this paper is to provide a preliminary overview of the phenomena observed during the experimental phase of the PHEBUS Fission Product Test FPT3. This experiment was the last in the series of 5 in-pile integral experiments performed by IRSN in the PHEBUS facility operated by the CEA on the site of Cadarache. Unlike the previous tests, FPT3 used boron carbide as absorber material instead of silver-indium-cadmium, so varying an important parameter impacting physico-chemical phenomena. FPT3 test course was in agreement with the pre-defined test protocol, including a 8,5-day irradiation phase, a fuel bundle degradation phase which lasted less than 5 hours and a 4-day long-term phase that consisted of an aerosol stage dedicated to the analysis of aerosol deposition mechanisms inside the containment vessel and a chemistry stage devoted to the analysis of the iodine chemistry. During the experiment, both the on-line instrumentation and the periodic samplings worked quite well. The fuel degradation progress could be analysed through both temperatures inside the bundle and gaseous concentration measurements performed in the circuit and inside the containment vessel. Some major events, like fuel clad and absorber rod failures or material relocations, were clearly correlated to both bundle and circuit instrumentation signals. The post test non destructive examinations of the fuel bundle (X-radiography, X- and γ-tomographies and γ-scanning) allowed to compare FPT2 and FPT3 bundle final degradation states. On-line γ-detector measurements coupled with numerous post test gamma-counted sequential samplings help for the characterization of the iodine behaviour inside the containment vessel during the degradation and the long term phases. The whole set of measurements appears self-consistent and provides new data for the iodine solubility inside the sump, the iodine gaseous fraction and the organic versus molecular iodine distribution inside the containment atmosphere.
Thermal Hydraulic Characteristics of the Integral Type Reactor, SMART-P for Validation of the Heat Removal Capability
Authors: Choi K. Y. , Cho S. , Yi S. J. , Park H. S. , Choi N. H. , Chung M. K.
Published: September 2008

The SMART is an integral type reactor with new innovative design features aimed at achieving a highly enhanced safety and improved economics. This paper focuses on the thermal hydraulic experimental program for the development of SMART. Thermal hydraulic responses for the transient operations of the SMART-P are experimentally investigated by using an integral effect test facility. The test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P with a power of 65MWt. In the present study, the VISTA facility was subjected to various transient conditions in order to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. Several experiments, including a heatup, a main coolant pump (MCP) speed change, and a power change, have been performed to investigate the heat transfer characteristics and the natural circulation performance of the primary system and the Passive Residual Heat Removal System (PRHRS) of the SMART-P by using the VISTA facility. Performance tests of a passive residual heat removal system (PRHRS) have also been carried out for its design optimization. Besides, several design basis accidents, such as an increase or a decrease of the feedwater flow, a loss of coolant flow, a control rod withdrawal, and a limited case of a loss of coolant accident (LOCA) on the line to the gas cylinder are under investigation in order to understand the thermal-hydraulic responses and finally to verify the system design of the SMART-P. Especially, the details of the experimental results for a loss of feedwater accident and a power increase accident due to a control rod withdrawal are explored in the present study.
Reactor Controllability of 3-Region-Core Molten Salt Reactor System: A Study on Load Following Capability
Authors: Yamamoto Takahisa , Mitachi Koshi , Nishio Masatoshi
Published: September 2008

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.
Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS
Authors: Kleinhietpaß Ingo D. , Unger Hermann , Wagner Hermann-Josef , Koch Marco K.
Published: September 2008

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport.
Debris Thermal Hydraulics Modeling of QUENCH Experiments
Authors: Kisselev Arcadii E. , Kobelev Gennadii V. , Strizhov Valerii F. , Vasiliev Alexander D.
Published: September 2008

Porous debris formation and behavior in QUENCH experiments (QUENCH-02, QUENCH-03) plays a considerable role and its adequate modeling is important for thermal analysis. This work is aimed to the development of a numerical module which is able to model thermal hydraulics and heat transfer phenomena occurring during the high-temperature stage of severe accident with the formation of debris region and molten pool. The original approach for debris evolution is developed from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The debris model is based on the system of continuity, momentum and energy conservation equations, which consider the dynamics of volume-averaged velocities and temperatures of fluid, solid and gaseous phases of porous debris. The different mechanisms of debris formation are considered, including degradation of fuel rods according to temperature criteria, taking into consideration some correlations between rod layers thicknesses; degradation of rod layer structure due to thermal expansion of melted materials inside intact rod cladding; debris formation due to sharp temperature drop of previously melted material due to reflood; and transition to debris of material from elements lying above. The porous debris model was implemented to best estimate numerical code RATEG/SVECHA/HEFEST developed for modeling thermal hydraulics and severe accident phenomena in a reactor. The model is used for calculation of QUENCH experiments. The results obtained by the model are compared to experimental data concerning different aspects of thermal behavior: thermal hydraulics of porous debris, radiative heat transfer in a porous medium, the generalized melting and refreezing behavior of materials, hydrogen production.
Refueling Liquid-Salt-Cooled Very High-Temperature Reactors
Authors: Forsberg Charles W. , Peterson Per F. , Cahalan James E. , Enneking Jeffrey A. , MacDonald Phil
Published: September 2008

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000°C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500°C, values that imply minimum refueling temperatures between 400 and 550°C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper.
An Integrated Simulator for the Maintenance Optimization of Light Water Reactors
Authors: Yoshimura Shinobu , Furuta Kazuo , Isobe Yoshihiro , Sagisaka Mitsuyuki , Noda Michiyasu , Wada Tetsuya , Akiba Hiroshi
Published: September 2008

An attempt was made to develop an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) based on PFM (Probabilistic Fracture Mechanics). The concept of the simulator is to provide a method to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency (both from cost and profit). The simulator will also provide a guideline regarding social acceptance of risk-based decision makings. This study has been conducted under “Innovative and Viable Nuclear Energy Technology (IVNET) Development Project” financially supported by Japanese METI.
Numerical and Experimental Studies on Impact Loaded Concrete Structures
Authors: Saarenheimo Arja , Hakola Ilkka , Ka¨rna¨ Tuomo , Hyva¨rinen Juhani , Tuomala Markku
Published: September 2008

An experimental set-up has been constructed for medium scale impact tests. The main objective of this effort is to provide data for the calibration and verification of numerical models of a loading scenario where an aircraft impacts against a nuclear power plant. One goal is to develop and take in use numerical methods for predicting response of reinforced concrete structures to impacts of deformable projectiles that may contain combustible liquid (“fuel”). Loading, structural behaviour, like collapsing mechanism and the damage grade, will be predicted by simple analytical methods and using non-linear FE-method. In the so-called Riera method the behavior of the missile material is assumed to be rigid plastic or rigid visco-plastic. Using elastic plastic and elastic visco-plastic material models calculations are carried out by ABAQUS/Explicit finite element code, assuming axisymmetric deformation mode for the missile. With both methods, typically, the impact force time history, the velocity of the missile rear end and the missile shortening during the impact were recorded for comparisons.
Uncertainty Analysis for a Depressurised Loss of Forced Cooling Event of the PBMR Reactor
Authors: Jansen van Rensburg Pieter A. , Sage Martin G.
Published: September 2008

This paper presents an uncertainty analysis for a Depressurised Loss of Forced Cooling (DLOFC) event that was performed with the systems CFD (Computational Fluid Dynamics) code Flownex for the PBMR reactor. An uncertainty analysis was performed to determine the variation in maximum fuel, core barrel and reactor pressure vessel (RPV) temperature due to variations in model input parameters. Some of the input parameters that were varied are: thermo-physical properties of helium and the various solid materials, decay heat, neutron and gamma heating, pebble bed pressure loss, pebble bed Nusselt number and pebble bed bypass flows. The Flownex model of the PBMR reactor is a 2-dimensional axi-symmetrical model. It is simplified in terms of geometry and some other input values. However, it is believed that the model adequately indicates the effect of changes in certain input parameters on the fuel temperature and other components during a DLOFC event. Firstly, a sensitivity study was performed where input variables were varied individually according to predefined uncertainty ranges and the results were sorted according to the effect on maximum fuel temperature. In the sensitivity study, only seven variables had a significant effect on the maximum fuel temperature (greater that 5°C). The most significant are power distribution profile, decay heat, reflector properties and effective pebble bed conductivity. Secondly, Monte Carlo analyses were performed in which twenty variables were varied simultaneously within predefined uncertainty ranges. For a one-tailed 95% confidence level, the conservatism that should be added to the best estimate calculation of the maximum fuel temperature for a DLOFC was determined as 53°C. This value will probably increase after some model refinements in the future. Flownex was found to be a valuable tool for uncertainly analyses, facilitating both sensitivity studies and Monte Carlo analyses.
CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0
Authors: To´th Sa´ndor , Le´gra´di Ga´bor , Aszo´di Attila
Published: September 2008

From the aspect of planning the power upgrading of nuclear reactors — including the VVER-440 type reactor — it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960 mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper.
Risk-Informed Assessment of Degraded Containment Vessels
Authors: Petti Jason P. , Spencer Benjamin W. , Graves Herman L.
Published: September 2008

This study couples structural analyses of four typical U.S. nuclear power plant containments with existing probabilistic risk analysis (PRA) models to assess the effects of degradation. This method is used to determine the increase in the early release frequencies (risk) due to postulated cases of corrosion in the steel liners and shells, as well as other forms of degradation.
COCOSYS and ASTEC Analyses of Iodine Multi-Compartment Tests in the ThAI-Facility
Authors: Weber G. , Allelein H.-J. , Funke F. , Kanzleiter T.
Published: September 2008

Post test calculations with the accident codes COCOSYS-AIM and ASTEC-IODE were performed on the iodine multi-compartment tests of the German ThAI program (Thermal hydraulics, Aerosols, Iodine). In the tests transport and adsorption/desorption behavior of gaseous I2 were measured in the 60 m3 vessel with a five-compartment configuration under severe accident conditions. The thermal hydraulic modules in COCOSYS and in the containment part of ASTEC (CPA) are nearly identical but not the iodine modules AIM and IODE. The adsorption/desorption model in AIM is based on ThAI data whereas in IODE correlations derived from laboratory-scale tests are used. A 50-zone nodalisation of the ThAI vessel was used with both codes. COCOSYS-AIM and ASTEC-IODE describe qualitatively correctly the I2 concentration differences of several orders of magnitude in periods with stratified atmosphere and the slow homogenization in a convective mixed atmosphere. However, both codes overestimate the gaseous I2 concentration at high relative humidity. The most likely reason is a slow reaction of deposited I2 to the non-volatile FeI2 on the steel surfaces, which has not been modeled sufficiently yet. Further experimental investigations in the ThAI facility are envisaged. A considering of the ThAI data in the I2 adsorption/desorption correlations may improve future ASTEC-IODE results. Nevertheless, the analyses of the large-scale ThAI iodine tests have been an important validation step for COCOSYS-AIM and ASTEC-IODE demonstrating the capability of multi-compartment I2 treatment.
Euratom Research Framework Programme on Reactor Systems
Authors: Deffrennes Marc , Hugon Michel , Manolatos Panagiotis , Van Goethem Georges , Webster Simon
Published: September 2008

The activities of the European Commission (EC) in the field of nuclear energy are governed by the Treaty establishing the European Atomic Energy Community (EURATOM). The research activities of the European Union (EU) are designed as multi-annual Framework Programmes (FP). The EURATOM 6th Framework Programme (EURATOM FP-6), covering the period 2002–2006, is funded with a budget of 1, 230 million Euros and managed by the European Commission. Beyond the general strategic goal of the EURATOM Framework Programmes to help exploit the potential of nuclear energy, in a safe and sustainable manner, FP-6 is designed to contribute also to the development of the “European Research Area” (ERA), a concept described in the Commission’s Communication COM(2000)6, of January 2000. Moreover EURATOM FP-6 contributes to the creation of the conditions for sharing the same nuclear safety culture throughout the EU-25 and the Candidate Countries, fostering the acceptance of nuclear power as an element of the energy mix. This paper gives an overview of the research activities undertaken through EURATOM FP-6 in the area of Reactor Systems, covering the safety of present reactors, the development of future safe reactors, and the needs in terms of research infrastructures and education & training. The actions under FP-6 are presented in their continuity of a ctions under FP-5. The perspectives under FP-7 are also provided. Other parts of the EURATOM FP, covering Waste Handling and Radiation Protection, as well as Fusion Energy, are not detailed in this paper.
Silicon Carbide and Uranium Oxide Based Composite Fuel Preparation Using Polymer Infiltration and Pyrolysis
Authors: Singh Abhishek K. , Zunjarrao Suraj C. , Singh Raman P.
Published: September 2008

Ceramic composite pellets consisting of uranium oxide, U3O8, particles in a silicon carbide matrix are fabricated using a novel processing technique based on polymer infiltration and pyrolysis (PIP). In this process, spherical particles of depleted uranium oxide, in the form of U3O8, are dispersed in liquid allylhydridopolycarbosilane (AHPCS), and subjected to pyrolysis up to 900°C under a continuous flow of ultra high purity (UHP) argon. Pyrolysis of AHPCS produces near-stoichiometric amorphous SiC at 900°C. Multiple polymer infiltration and pyrolysis (PIP) cycles are required to minimize open porosity and densify the silicon carbide matrix, in order to enhance the mechanical strength of the material. Structural characterization is carried out after first pyrolysis to investigate chemical interaction between U3O8 and SiC. The physical and mechanical properties are also quantified, and it is shown that this processing scheme promotes uniform distribution of uranium fuel source along with a high ceramic yield of the parent matrix. Furthermore, the processing technique involves lower energy requirements than conventional sintering processes currently in practice.
Numerical Investigation of Thermal Stratification in the Primary Circuit of VVER-440 Type Reactors
Authors: Boros Ildiko´ , Aszo´di Attila , Le´gra´di Ga´bor
Published: September 2008

Thermal stratification in the primary loops and in the connected pipes can limit the lifetime of the piping, or lead to penetrating cracks due to the stresses caused by the temperature differences and the cyclic temperature changes. Therefore it is essential to determine the thermal hydraulic parameters of the stratified flow. The determination of the affected pipes can be based on the international operational experience and on engineering consideration. The most affected pipes in PWRs are the pressurizer surge line, the injection pipe of the emergency core cooling systems and the feedwater injection pipe of the steam generators. CFD codes can provide an appropriate tool for the examination of the development and the breaking up of the stratification and the determination of the temperature distribution. However, the challenge of the uncertainty of the boundary conditions has to be faced because of the unknown flow circumstances. According to an extensive evaluation, performed in 1998 by the VEIKI, in the VVER-440/213 units of Paks NPP the most affected pipe is the pressurizer surge line [1]. To find out the possible thermal stratification in the surge line, a temperature monitoring system was installed on the YP20 leg of the surge line of the Unit 1 of the Paks NPP in 2000. The measurements showed that during the heat-up period there is a thermal stratification almost all time in the surge line [2]. The maximum temperature differences reach 140 K (140 °C). The surge line has been modeled with the CFD code CFX-5.7. The performed transient simulations confirmed the existence of a thermal stratification in the surge line, but showed permanent recirculation of colder coolant in the lower layer, caused by the asymmetric arrangement of the surge line legs and the asymmetric connection of the two legs to the main loop. In this paper, the surge line model and the results of the transient simulations are presented. The CFD model of the injection pipe of the high pressure Emergency Core Cooling System and the performed simulations for the analysis of occurrence of thermal stratification are presented as well.
Ultrasonic Cross-Correlation Flow Measurement: Theory, Noise Contamination Mechanisms, and a Noise Mitigation Technique
Authors: Jenkins D. M. , Lysak P. D. , Capone D. E. , Brown W. L. , Askari V.
Published: September 2008

Based on past experience with ultrasonic cross-correlation flow meters in power plant environments, the presence of spatially correlated noise due to pressure waves, vibration, or sources other than transport of turbulent eddies will cause a bias in the time delays measured by the meter. Several techniques were developed to detect the existence of such correlated noise and correct for its effect at plant conditions. An analytical and experimental investigation was performed to further understand the basic physics of the noise mechanisms. The dominant error mechanisms investigated in this work were speed of sound perturbations due to pressure fluctuations and beam path length changes due to wall vibration. An analytical model was formulated which estimates the signal level of the flow meter based on the turbulent velocity field. From this model, an estimate of the system noise which would cause contamination could be determined. A test at a water tunnel facility was performed in order to evaluate the noise mechanisms. During this test, measurements were taken with and without controlled noise sources. Pressure and acceleration measurements were used to evaluate a coherent noise removal technique developed to mitigate the impact of noise in the ultrasonic cross-correlation flow measurement. The coherent noise removal technique was shown to be effective in removing noise during the water tunnel test.
Detailed Core Design and Flow Coolant Conditions for Neutron Flux Maximization in Research Reactors
Authors: Teruel F. E. , Uddin Rizwan
Published: September 2008

Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. In addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0E14 n·cm−2s−1), a moderate thermal neutron flux zone (2.5E14 n·cm−2s−1), and a low thermal flux zone (1.0E14 n·cm−2s−1). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions.
Visualization of Sub-Cooled Flow Film Boiling in Horizontal Channel on Flat Heat Exchangers
Authors: Wu Wen , Chen Peipei , Jones Barclay G. , Newell Ty A.
Published: September 2008

The observation of sub-cooled flow film boiling was performed on a single-side-heated flat heat exchanger by using refrigerant R134a as the testing fluid. A stable vapor film was observed with the co-existence of sub-cooled bulk liquid. Vapor bubbles might release from the film, with the number and frequency changing under different conditions. The purpose of this paper is to describe these observations, and to provide a quantitative analysis of the phenomena. Results are compared with Berenson’s model on horizontal heating surface, with discussion and suggestion made.
Margin Assessment of AP1000 Loss of Flow Transient
Authors: Carlin Edward L. , Hilton Peter A. , Sung Yixing
Published: September 2008

The Reactor Coolant System (RCS) of the AP1000 plant consists of two circulating loops. Each loop contains two canned motor Reactor Coolant (RC) pumps that have a rotating inertia to provide RCS flow coastdown if power to the pumps is lost. Westinghouse analysis of the complete loss of flow (CLOF) accident in support of the AP1000 design certification was based on the USNRC-approved traditional methodology applied to operating plants. The RCS response during the transient was predicted using the LOFTRAN code based on a reactivity insertion curve highly skewed to the bottom of the reactor core, but the calculation of Departure from Nucleate Boiling Ratio (DNBR) was performed assuming a top-skewed axial power profile. A more realistic margin assessment can be made by using an improved method similar to Westinghouse RAVE methodology recently approved by the USNRC. The improved method uses the three-dimensional kinetic nodal code SPNOVA coupled with the reactor core thermal-hydraulic code VIPRE-W for predicting the reactor core response during the CLOF transient. The improved method significantly improves margin predictions by generating core power distributions consistent with the trip reactivity changes for the DNBR calculation. The margin assessment showed that the improved method resulted in a 19% DNBR increase as compared to the traditional method for the AP1000 CLOF transient.
A Study on a Hybrid Approach for Diagnosing Faults in Nuclear Power Plant
Authors: Yang M. , Zhang Z. J. , Peng M. J. , Yan S. Y. , Wang H. , Ouyang J.
Published: September 2008

Proper and rapid identification of malfunctions is of premier importance for the safe operation of Nuclear Power Plants (NPP). Many monitoring or/and diagnosis methodologies based on artificial and computational intelligence have been proposed to aid operator to understand system problems, perform trouble-shooting action and reduce human error under serious pressure. However, because no single method is adequate to handle all requirements for diagnostic system, hybrid approaches where different methods work in conjunction to solve parts of the problem interest researchers greatly. In this study, Multilevel Flow Models (MFM) and Artificial Neural Network (ANN) are proposed and employed to develop a fault diagnosis system with the intention of improving the success rate of recognition on the one hand, and improving the understandability of diagnostic process and results on the other hand. Several simulation cases were conducted for evaluating the performance of the proposed diagnosis system. The simulation results validated the effectiveness of the proposed hybrid approach.
Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor
Authors: Ando Masato , Kubo Shigenobu , Kamishima Yoshio , Iitsuka Toru
Published: September 2008

The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed.
Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT
Authors: Jevremovic Tatjana , Hursin Mathieu , Satvat Nader , Hopkins John , Xiao Shanjie , Gert Godfree
Published: September 2008

The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Rebalancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined.
Spectrum Shifting as a Mechanism to Improve Performance of VHTRs with Advanced Actinide Fuels
Authors: Tsvetkov Pavel V. , Ames David E. , Pritchard Megan L. , Alajo Ayodeji B.
Published: September 2008

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of achieving ultra-long life VHTR (Very High Temperature Reactor) configurations by utilizing minor actinides as a fuel component. The postulated principal mechanism is an enhanced involvement of self-generated fissile compositions based on spent LWR fuel. Since pebble bed and prismatic core designs permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties of minor actinides by neutron spectrum shifting through configuration adjustments. Depending on neutron spectra, neptunium, americium and curium may contribute to small reactivity swings (self-stabilization) over prolonged irradiation periods. The presented analysis is focused on achievability of spectral variations and their potential impact. In principle, promising core features and performance characteristics have been demonstrated.
Analysis of Fluid Flow and Heat Transfer in a Falling Film Using Volume of Fluid Method
Authors: Vishnoi A. K. , Chandraker D. K. , Vijayan P. K.
Published: September 2008

This paper deals with the applicability of VOF method for interface tracking with heat transfer and validation of the VOF approach using experimental data. A vertical channel flow problem in which the liquid is falling inside a vertical channel along one of the walls from the top is analysed and liquid–air interface is tracked. In the same problem analysis of heat transfer from the wall has been incorporated. This approach has a potential to predict liquid film thickness in a heated tube/subchannel which will lead to the evaluation of critical power (power corresponding to critical heat flux).
Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure
Authors: Karalevicˇius Renatas , Dundulis Gintautas , Rimkevicˇius Sigitas , Usˇpuras Eugenijus
Published: September 2008

The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident.
Comprehensive Cost Estimation Method for Decommissioning
Authors: Kudo Kenji , Kawatsuma Shinji , Rindo Hiroshi , Watabe Kozo , Tomii Hiroyuki , Shiraishi Kunio , Yagi Naoto , Fukushima Tadashi , Zaitsu Tomohisa
Published: September 2008

Japan Atomic Energy Research Institute (JAERI) played a leading role in basic research in the field of atomic energy research and development, while Japan Nuclear Cycle Development Institute (JNC) did a major role in FBR cycle development and high level waste disposal. According to Japanese government’s decision in December 2001, JAERI and JNC was merged as of October 1st. 2005. The new organization, Japan Atomic Energy Agency (JAEA) is an institute for comprehensive R&D for atomic energy, and which is the largest research and development institute among Japanese Governmental organization. Its missions are basic research on atomic, R&D for nuclear fuel cycle, decommissioning and disposal for own facilities and waste, contribution to safety and non-proliferation, etc. The JAEA owns a number of nuclear facilities: research reactors such as JRR-2 and Joyo, prototype reactors such as ATR “Fugen” and FBR “Monju”, fuel cycle plants such as Uranium Enrichment Demonstration Plant at ningyo-toge, MOX fuel plants at Tokai, Reprocessing Plant at Tokai, and Hot Laboratories such as JRTF and FMF. As a part of preparation of the mergence, JNC and JAERI have jointly developed a comprehensive cost estimation method for decommissioning, based on decommissioning and upgrading experiences of JAERI and JNC. This method has adopted more estimation formulae for typical decommissioning activities than ever, so as to be more reliable. JAERI and JNC had estimated by using the comprehensive estimation method for decommissioning, and concluded the total cost for decommissioning would be 600 billion yen (approx. 5 billion USD).
Experimental Evaluation of Pool Fire Suppression Performance of Sodium Leak Collection Tray in Open Air
Authors: Parida F. C. , Rao P. M. , Ramesh S. S. , Malarvizhi B. , Gopalakrishnan V. , Rao E. H. V. M. , Kasinathan N. , Kannan S. E.
Published: September 2008

In the event of sodium leakage from heat transfer circuits of fast breeder reactors (FBR), liquid sodium catches fire in ambient air leading to production of flame, smoke and heat. One of the passive fire protection methods involves immediate collection of the leaking sodium to a sodium hold-up vessel (SHV) covered with a sloping cover tray (SCT) having a few drain pipes and one vent pipe (as in Fig. 1). As soon as the liquid sodium falls on the sloping cover tray, gravity guides the sodium through drain pipes into the bottom tray in which self-extinction occurs due to oxygen starvation. This sodium fire protection equipment called leak collection tray (LCT) works without the intervention of an operator and external power source. A large number of LCTs are strategically arranged under the sodium circulating pipe lines in the FBR plants to serve as passive suppression devices. In order to test the efficacy of the LCT, four tests were conducted. Two tests were with LCT having three drain pipes and rest with one. In each experiment, nearly 40 kg of hot liquid sodium at 550 °C was discharged on the LCT in the open air. Continuous on-line monitoring of temperature at strategic locations (∼ 28 points) were carried out. Colour videography was employed for taking motion pictures of various time-dependent events like sodium dumping, appearance of flame and release of smoke through vent pipes. After self-extinction of sodium fire, the LCT was allowed to cool overnight in an argon atmosphere. Solid samples of sodium debris in the SCT and SHV were collected by manual core drilling machine. The samples were subjected to chemical analysis for determination of unburnt and burnt sodium. The results of the four tests revealed an interesting feature: LCT with three drain pipes showed far lower sodium collection efficiency and much higher sodium combustion than that with just one drain pipe. Thermal fluctuations in temperature sensor located near the tip of the drain pipe have indicated that transient freezing and remelting processes are responsible for this phenomenon. Moreover comparison of test results between present and earlier experiments has revealed that the LCT with funnel shaped SCT is superior to that with boat shaped SCT.
Spatially Informed Plant PRA Models for Security Assessment
Authors: Wheeler Timothy A. , Thomas Willard , Thornsbury Eric
Published: September 2008

Traditional risk models can be adapted to evaluate plant response for situations where plant systems and structures are intentionally damaged, such as from sabotage or terrorism. This paper describes a process by which traditional risk models can be spatially informed to analyze the effects of compound and widespread harsh environments through the use of “damage footprints.” A “damage footprint” is a spatial map of regions of the plant (zones) where equipment could be physically destroyed or disabled as a direct consequence of an intentional act. The use of “damage footprints” requires that the basic events from the traditional probabilistic risk assessment (PRA) be spatially transformed so that the failure of individual components can be linked to the destruction of or damage to specific spatial zones within the plant. Given the nature of intentional acts, extensive modifications must be made to the risk models to account for the special nature of the “initiating events” associated with deliberate adversary actions. Intentional acts might produce harsh environments that in turn could subject components and structures to one or more insults, such as structural, fire, flood, and/or vibration and shock damage. Furthermore, the potential for widespread damage from some of these insults requires an approach that addresses the impacts of these potentially severe insults even when they occur in locations distant from the actual physical location of a component or structure modeled in the traditional PRA.
On the Development of the ISAAC Graphic Model
Authors: Kim K. R. , Kim S. D. , Song Y. M.
Published: September 2008

Recently the development of graphical methods using the ISAAC code’s calculation data has been started in order to show the Wolsong 1 & 2 PHWR plants behaviour during the severe accidents. This graphic model is designed to provide two basic functions: one is to provide the graphical display of several plants systems together with the important parameters. For example, the representative T/H behaviour, fuel behaviour, fuel channel behaviour, reactor core behaviour, containment behaviour and fission product behaviour are going to be displayed in a graphic window. The other function is the control capability equipped with the controllable valves and pumps in the PHWR SAMG. In this paper, details of the elementary technical aspects of the ISAAC graphic model are presented which are the system structure, ISAAC variable definition, data communication methods and graphical display generation.