bluuebunny/crossref_metadata_2025 huggingface dataset:
Lumped Parameter Model for Dynamic Performances of Plate-Fin Recuperator
Authors: Ding Ming , Wang Jie , Yang Xiaoyong , Shi Lei , Su Qingshan
Published: September 2008

A lumped parameter model was developed to study dynamic performances of plate-fin recuperator in high temperature gas-cooled reactor with direct helium turbine cycle (HTGR-GT). For the core heat capacitance of recuperator was far larger than heat capacitance and thermal flow rate of helium, it was reasonable to ignore the influence of heat capacitance of fluid on dynamic characteristics of recuperator and develop the lumped parameter model with infinite core heat capacitance. The model was solved by four-order Rounge-Kutta method, considering the influence of temperature on helium thermal properties. Based on the lump parameter model, transient response of outlet temperatures of recuperator was analyzed when step and ramp changes of inlet temperatures of recuperator took place in hot side, as well as mass flow rate of recuperator. Transient responses of the core temperature and outlet temperatures of helium were also analyzed while power was regulated in course of normal operation and total electric load was rejected from full power.
Irradiation Embrittlement Monitoring Programs of RPV’s in the Slovak Republic NPP’s
Authors: Kupcˇa L’udovi´t
Published: September 2008

Four types of surveillance programs were (are) realized in Slovak NPP’s: • “Standard Surveillance Specimen Program” (SSSP) was finished in Jaslovske´ Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4; • “Extended Surveillance Specimen Program” (ESSP), was prepared for Jaslovske´ Bohunice NPP V-2 with aim to validate the SSSP results; • For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program “Modern Surveillance Specimen Program” (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life; • For the Bohunice V-1 NPP was finished “New Surveillance Specimen Program” (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years; • New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV’s austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: • increasing of neutron fluence measurement accuracy; • substantial improvement the irradiation temperature measurement; • fixed orientation of samples to the centre of the reactor core; • minimum differences of neutron dose for all the Charpy-V notch and COD specimens; • the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV’s unit-1 and 2, based on the fundamental postulate — to provide the irradiation embrittlement monitoring till the end of units operation. The “New Surveillance Specimen Program” (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The results showed us that the NPP V-1 RPV’s can be operated without any technical limitation for minimum next 20 years. Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1 and 2) was prepared as the part of research project dealing with the WWER-440 units ageing management.
Fire Risk Analysis for Armenian NPP Confinement
Authors: Poghosyan Shahen , Malkhasyan Albert , Bznuni Surik , Amirjanyan Armen
Published: September 2008

Major fire occurred at Armenian NPP (ANPP) in October 1982 showed that fire-induced initiating events (IE) can have dominant contribution in overall risk of core damage. Probabilistic Safety Assessment study for fire-induced initiating events for ANPP was initiated in 2002. Analysis was performed for compartments fires in which could result in failure of components which are necessary for reactor cold shutdown. Analysis shows that main risk from fire at ANPP is conditioned by fire in cable tunnels 61–64. Meanwhile fire in confinement compartments don’t have significant contribution to overall risk of core damage. The exception is so called “confinement valves’ compartment” (room #A-013/2) fire (more than 7.5% of CDF) in which fire could result in the loss of coolant accident with unavailability of primary makeup system, which directly leads to core damage. Detailed analysis of this problem that is common for typical WWER-440/230 reactors with no hermetic MCPs and recommendations for solution are presented in this paper.
Axial Burnup Profile Influence on Criticality Safety of WWER Spent Fuel
Authors: Bznuni Surik , Amirjanyan Armen , Poghosyan Shahen
Published: September 2008

Criticality safety assessment for WWER-440 NUHOMS® cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in a such way that the neutron multiplication factor keff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. Nuclear and Radiation Safety Centre of Armenian Nuclear Regulatory Authority (NRSC ANRA) in order to improve future fuel storage efficiency, initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon wasn’t taken into account in the Safety Analysis Report of NUHOMS® spent fuel storage constructed on the site of ANPP. Although ANRA doesn’t yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burn up profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality safety analysis of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn’t taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The Actinides and Actinides + Fission Products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the keff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect. Using Burn up profiles of Control Assemblies in both approaches leads to much more increasing than in case of Working Assemblies. Usually increasing burn up leads to decreasing Δkeff, hence decreasing end effect. However for WWER-440 Control Assemblies that worked only within 6th (operative) group increasing burn up leads to increasing of the end effect.
Financing Strategies for a Nuclear Fuel Cycle Facility
Authors: Shropshire David , Chandler Jess
Published: September 2008

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the government case. The uncertainty in operations, leading to lower than optimal processing rates (or annual plant throughput), is the most detrimental issue to achieving low unit costs. Conversely, lowering debt interest rates and the required return on investments can reduce costs for private industry.
Development and Testing of an Improved Westinghouse Containment Design Basis Analysis Methodology
Authors: Ofstun Richard P. , Rybicki Jon R. , Smith Lawrence C.
Published: September 2008

Westinghouse utilizes separate computer models for the loss-of-coolant accident (LOCA) mass and energy (M&E) release and containment response calculations for the containment design basis accident (DBA) analysis. The current LOCA M&E release model uses very conservative assumptions to calculate the rate of steam generator (SG) energy transfer to the reactor coolant system (RCS). Westinghouse is upgrading the containment DBA analysis to use WCOBRA/TRAC (WC/T) for the LOCA M&E release calculation and GOTHIC for the containment response calculation. Furthermore, the formerly separate calculations can now be performed concurrently to allow communication between the RCS and containment calculation models. WC/T utilizes more representative models for calculating the rate of SG energy transfer to the RCS than the currently approved LOCA M&E release calculation model. The full length emergency core heat transfer separate-effects and system-effects test (FLECHT-SEASET) data was used to validate the WC/T SG model for the post-blowdown LOCA M&E release calculation. The WC/T SG model originally calculated a higher SG energy transfer rate than the test data, which is conservative for containment analysis. The WC/T code was updated to include recommended changes to model the SG tube quench front. Following implementation of the recommended software changes, WC/T calculates a more realistic, yet still conservative, SG energy transfer rate. A WC/T plant model was developed to perform a LOCA M&E release calculation, and the WC/T results were compared to results from the currently approved LOCA M&E release calculation model. WC/T calculated a lower post-blowdown release rate than the currently approved LOCA M&E release calculation model. The lower post-blowdown release rate will provide additional peak pressure margin for plants that rely on passive cooling, containment spray, and/or fan coolers to control the containment peak pressure, and is also expected to extend the time of ice melt-out for ice condenser containments. The upgraded containment DBA analysis model that utilizes WC/T and GOTHIC concurrently will establish a conservative, but more representative long-term containment pressure and temperature response. Results based upon the GOTHIC containment model running with the WC/T plant model have shown a pressure reduction of approximately 10 psi (70 kPa) in the post-blowdown pressure transient.
Experiments on Damping of Acoustic Waves in Water-Filled Pipes
Authors: Chatoorgoon Vijay , Li Qizhao
Published: September 2008

A simple, fundamental experimental study was conducted to better understand acoustic wave propagation is fluid-filled pipes. Three experiments were undertaken: the first with zero flow and a closed outlet end, the second with turbulent flow and an open outlet end and the third with zero flow and an open outlet end. The intent was to obtain data for model comparison and to determine the effect of turbulent flow on the system response. New insights are obtained and reported.
The Impact of RELAP5 Pipe Break Flow Rates Associated With Reverse Flow Limiter Removal for Steam Generator Replacement
Authors: Zheng Dong , Jarvis Julie M. , Vieira Allen T.
Published: September 2008

Pipe break flow rates are calculated for a main feedwater line break (FWLB) in the main steam valve vault (MSVV) for a PWR Steam Generator Replacement (SGR). A reverse flow limiter is installed in the original steam generator (OSG) feedwater nozzle to limit the blowdown flowrate in the event of a postulated FWLB. This feature is not incorporated in the replacement steam generator (RSG) design. The change in RSG nozzle design in conjunction with new operating conditions results in increased FWLB mass and energy releases which can impact environmental temperatures and pressures and flooding levels. In the United States, benchmarking for safety related analyses is necessary in consideration of 10CFR50.59 requirements. RELAP5/MOD3 is used to model the pipe break flowrates for a FWLB at different break locations. The benchmark FWLB blowdown releases are larger than the OSG design basis blowdown releases due to differences in RELAP5/MOD3 versions which are found to have different algorithms for subcooled choked flow. The SGR FWLB blowdown release rates are determined to have minimal impact on the compartment temperature and pressure response. However, the flooding levels and associated equipment qualification are potentially impacted. Modeling techniques used to minimize the impact of the SGR blowdown releases on MSVV flooding levels include modeling flashing effects, more realistic RSG temperature distribution, inventory depletion and Auxiliary Feedwater (AFW) flow initiation time, and considering loss of offsite power scenarios. A detailed flooding hazard evaluation is needed, which considers the actual main feedwater isolation times to ensure that environmentally qualified safety related components, required to mitigate the effects of a FWLB inside the MSVV, can perform their safety function prior to being submerged.
Single Rod Heat Transfer Tests to Study the Effects of Crud Deposition
Authors: Wang G. , Byers W. A. , Karoutas Z. E. , Hochreiter L. E. , Young M. Y. , Jacko R. J.
Published: September 2008

Researchers have performed many studies to try to understand crud formation on the fuel pin clad surfaces, which has been observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling and precipitation reactions. Crud deposits, may cause an unexpected change in core power distribution known as crud induced power shift (CIPS) or axial offset anomaly (AOA) if boron species accumulate in the deposits. If the crud deposit is thick enough, the potential exists for fuel rod surface dryout. The Westinghouse Advanced Loop Tester (WALT) at the George Westinghouse Science and Technology Center (STC) will be utilized to examine the effects of crud formation on fuel pin clad dryout. This paper describes a single heated rod test facility designed and fabricated by Westinghouse to study the effects of crud deposition at PWR reactor operating conditions. This test loop is a single rod facility with or without grid support structures on the heater rod and can be used for forced convection and sub-cooled boiling tests with and without crud deposition. Sub-cooled boiling experiments at PWR reactor operating conditions are currently being performed at this facility. The single electrical heated rod in this test facility is instrumented with four movable thermocouples to measure the inside wall temperatures at four azimuthal locations within the rod. In addition, there are two fixed thermocouples to measure the inlet and outlet temperatures of the water flowing on the outside of the heated rod such that an overall energy balance (i.e. comparing the heat absorbed by the water coolant to the measured rod electrical power) can be performed on the facility. This paper will present forced convection and boiling heat transfer curves for clean rod surfaces. Comparison with forced convection correlations and sub-cooled boiling correlations are also presented in this paper.
Effect of Obstacle Shape on the Deposition Rate in Annular Two-Phase Flow
Authors: Fujita Takayuki , Minamitani Jun , Okawa Tomio , Kataoka Isao
Published: September 2008

In annular two-phase flow, it is known that the deposition rate of droplets markedly increases if an obstacle is placed in a flow channel. In most of the available models to evaluate the obstacle effect on deposition rate, turbulence augmentation at the downstream of obstacle is considered as the key mechanism of the deposition enhancement. In this study, it was examined where the increase of droplet deposition occurred around the obstacle to reveal the primary cause of deposition enhancement. The test section was a round tube of 5 mm in inside diameter and air and water were used as test fluids; the double film extraction technique was adopted to measure the deposition rate. To specify the position where the deposition enhancement occurred, the liquid film was extracted at the upstream, middle and downstream of the flow obstacle. As a result, it was indicated that the significant amount of droplets was deposited not only at the downstream of flow obstacle but also at the upstream of obstacle. Using five tubular obstacles of different inside and outside diameters, the influence of obstacle geometry on the deposition rate was investigated. It was shown that the position of deposition enhancement was influenced significantly by the obstacle geometry.
Development of a Slim Manipulator Type Fuel Handling Machine for a Commercialized Fast Reactor
Authors: Chikazawa Yoshitaka , Usui Shinichi , Konomura Mamoru , Sadahiro Daisuke , Tozawa Katsuhiro , Hori Toru , Toda Mikio , Kotake Shoji
Published: September 2008

A seismic analysis has been performed showing that the seismic interaction between the UIS and the FHM can be avoided eliminating the internal clearance of bearing at the FHM arm joint. An angular contact ball bearing is suitable for the new FHM since it can eliminate gaps by preload pressure. A major problem of the FHM bearings is lubrication since the contact pressure between the steel rings and ball of the ball bearing is larger than that of the roller bearing used in the previous design. Additionally, FHM operating temperature is about 200deg-C and normal grease is not applicable under argon gas with sodium vapor to prevent contamination of grease in the primary sodium coolant. An endurance test with 1/10 scale bearings in the air has been performed to show applicability of angular contact ball bearings to the FHM arm joint. The results with 20,000cycle showed that bearings with combination of MoS2 coating steel rings and ceramics balls can be tolerable as the FHM operating condition. A full-scale bearing test in argon gas with sodium vapor has also been demonstrated to reveal bearing size and sodium vapor effects.
SCC Propagation Rate of Type 304, 304L Steels Under Oceanic Air Environment
Authors: Kosaki Akio
Published: September 2008

Corrosion integrity of canister in the concrete cask for spent fuel storage is very important because the canister serves to maintain the sealability over the storage period of 40 to 60 years. Natural exposure and accelerated corrosion tests of conventional stainless steels for canister, that are Type 304, 304L, and 316(LN), for concrete cask’s canister have been conducted by using many three Point Bending (3PB) test specimens and compared. The SCC propagation rates in Type 304 and 304L at the natural condition were about 1.2E−12 to 1.8E−11 m/s at the K (Stress Intensity Factor) range of 0.6 to 9.0 MPa√m, and that of the accelerate test (60 degrees C, 95%RHS., filled with NaCl mist) were about 1.0E−10 to 3.5E−9 m/s at the K range of 0.3 to 32 MPa√m. The SCC propagation rates under both natural and accelerated conditions were independent with K. Both da/dt values of the direct exposure test and of the under glass exposure test were in the same scattering band.
Development and Validation of the ECART Code for the Safety Analysis of Nuclear Installations
Authors: Parozzi F. , Paci S.
Published: September 2008

ECART can simulate the thermal-hydraulic behavior of LWR and GCR plants under severe accident conditions together with the transport of radiotoxic substances. This tool is still under improvement and assessment for new applications in non-nuclear risk studies, new advanced and fusion reactors. As regards accidents with fires within closed environments, specific models can simulate both thermal and chemical processes, accounting for combustion of gases and solids, as well as pool fires. The radiative heat transfer and the action of water sprays on atmosphere cooling and aerosol removal are properly taken into account, as verified by comparing the code predictions to full-scale experiments and to the consequences of fire accidents really occurred. About its application on tokamak fusion plants, a large validation activity is underway, mainly based on the analyses of ad hoc experimental programs or code benchmark promoted inside EURATOM Fusion Technology Programme. The correct simulation of the main phenomena occurring in ICE and STARDUST facilities, as well as the comparison with the results of codes employed in the fusion safety studies, demonstrates the applicability of ECART models in performing a realistic prediction of the whole incidental sequence, accounting both thermal-hydraulics and dust transport also inside fusion plants.
Non Invasive Water Level Monitoring on Boiling Water Reactors Using Internal Gamma Radiation: Application of Soft Computing Methods
Authors: Fleischer Sebastian , Hampel Rainer
Published: September 2008

To provide best knowledge about safety-related water level values in boiling water reactors (BWR) is essentially for operational regime. For the water level determination hydrostatic level measurement systems are almost exclusively applied, because they stand the test over many decades in conventional and nuclear power plants (NPP). Due to the steam generation especially in BWR a specific phenomenon occurs which leads to a water-steam mixture level in the reactor annular space and reactor plenum. The mixture level is a high transient non-measurable value concerning the hydrostatic water level measuring system and it significantly differs from the measured collapsed water level. In particular, during operational and accidental transient processes like fast negative pressure transients, the monitoring of these water levels is very important. In addition to the hydrostatic water level measurement system a diverse water level measurement system for BWR should be used. A real physical diversity is given by gamma radiation distribution inside and outside the reactor pressure vessel correlating with the water level. The vertical gamma radiation distribution depends on the water level, but it is also a function of the neutron flux and the coolant recirculation pump speed. For the water level monitoring, special algorithms are required. An analytical determination of the gamma radiation distribution outside the reactor pressure vessel is impossible due to the multitude of radiation of physical processes, complicated non-stationary radiation source distribution and complex geometry of fixtures. For creating suited algorithms Soft Computing methods (Fuzzy Sets Theory, Artificial Neural Networks, etc.) will be used. Therefore, a database containing input values (gamma radiation distribution) and output values (water levels) had to be built. Here, the database was established by experiments (data from BWR and from a test setup) and simulation with the authorised thermofluid code ATHLET.
ASTEC: An Integral Code for Simulation of Severe Light Water Reactor Accidents
Authors: Reinke N. , Neu K. , Allelein H.-J.
Published: September 2008

The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. Thus, the main fields of ASTEC application are intended to be accident sequence studies, uncertainty and sensitivity studies, probabilistic safety analysis level 2 studies as well as support to experiments. The modular structure of ASTEC allows running each module independently and separately, e.g. for separate effects analyses, as well as a combination of multiple modules for coupled effects testing and integral analyses. Among activities concentrating on the validation of individual ASTEC modules describing specific phenomena, the applicability to reactor cases marks an important step in the development of the code. Feasibility studies on plant applications have been performed for several reactor types such as the German Konvoi PWR 1300, the French PWR 900, and the Russian VVER-1000 and −440 with sequences like station blackout, small- or medium-break loss-of-coolant accident, and loss-of-feedwater transients. Subject of this paper is a short overview on the ASTEC code system and its current status with view to the application to severe accidents sequences at several PWRs, exemplified by selected calculations.
The Potential of Different Concepts of Fast Breeder Reactor for the French Fleet Renewal
Authors: Massara Simone , Tetart Philippe , Lecarpentier David , Garzenne Claude , Mourogov Alexandre
Published: September 2008

The performances of different concepts of Fast Breeder Reactor (Na-cooled, He-cooled and Pb-cooled FBR) for the current French fleet renewal are analyzed in the framework of a transition scenario to a 100% FBR fleet at the end of the XXIst century. Firstly, the modeling of these three FBR types by means of a semi-analytical approach in TIRELIRE - STRATEGIE, the EDF fuel cycle simulation code, is presented, together with some validation elements against ERANOS, the French reference code system for neutronic FBR analysis (CEA). Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and waste production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy. Thus, some of the key parameters defining the dynamic of FBR deployment are highlighted, to inform the orientation of R&D in the development and optimization of these systems.
Effect of Opening on the Response of Containment Model to Pressure Loading
Authors: Agrawal M. K. , Ravi Kiran A. , Ghosh A. K. , Kushwaha H. S.
Published: September 2008

The Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other phenomena which occur during simulated accident conditions in Nuclear power Plants. The facility consists of a concrete containment model having a volumetric scale ratio of 200:1 with respect to the actual containment of Indian Pressurized Heavy Water Reactor. The structure is designed for pressure of 1.73 Kg/cm2 for specified leak tightness. Adequacy to withstand design pressure is checked by test as well as numerical analysis before commissioning of the facility. Accordingly Containment building model has been analyzed by finite element method for internal design pressure and dead weight. Analysis has been carried out for the structure with and without the opening in the containment. Effect of opening on the response of containment has been studied. The paper includes the modeling methodology, maximum deflection and stress amplification around the opening for various models.
Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code
Authors: Gorzel A.
Published: September 2008

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.
Design Study of 300 MWt PWR Fueled With UO2 Coated Fuel Particle
Authors: Rivai Abu Khalid , Aziz Ferhat , Takahashi Minoru
Published: September 2008

A neutronic design was performed for 300 MWt Pressurized Water Reactor (PWR) with UO2 compacts made of coated fuel particles (CFP) comparing that with sintered pellets made of UO2 powder as ordinary fuel type. UO2 CFP type was enriched 4.8% of 235U and UO2 ordinary type was enriched 5% of 235U. Both reactors were operated with single batch refueling system with a cycle period of 3 years. The purpose of the design was to investigate the applicability of UO2 CFP type to PWR comparing with UO2 ordinary type that commonly used for PWR. The calculation was done with SRAC (Standard Reactor Analysis Code) computer code and nuclear library of JENDL-33. The results of calculation showed that k-effective for both type of fuel could be maintained at critical condition for 3 years operation without refueling. The k-effective and the Doppler coefficients have been calculated for all types of fuel at 600 K and 900 K degrees. The results of calculation showed that for all types of fuel Doppler coefficient was negative, which was good for inherent safety characteristic. The size optimization design showed that the active core dimensions of UO2 CFP type reactor was about 2 times larger than the UO2 ordinary type reactor.
Experiments on Gas Mixing and Stratification Driven by Jets and Plumes in Large-Scale, Multi-Compartment Geometries
Authors: Zboray Robert , Paladino Domenico , Auban Olivier
Published: September 2008

The present paper discusses experiments carried out to examine mixing of different gases (steam, air) and the evolution their distributions in large-scale, multi compartment geometry imitating nuclear reactor containment compartments. The flow and the mixing process in the experiments are driven by plumes and jets representing source structures with different momentum-to-buoyancy strength. The time evolution of the relevant parameters like gas concentrations, velocities and temperatures are followed using dedicated instrumentation. The data obtained is meant to be used for the validation and development of high-resolution, mainly CFD based, 3D computational tools for nuclear reactor containment safety analysis.
Numerical Simulation of Flow Around a Hot Particle With a Vapor Film Using a Volume-of-Fluid Interface Tracking Method
Authors: Yuan Minghao , Yang Yanhua
Published: September 2008

Movement characteristics of a melt particle with a vapor film moving in coolant could affect the distribution of phases and then the global behavior in premixing phase of steam explosion. Detail CFD simulation of flow around the particle could improve the understanding of the resistance on the particle. The work of this paper is to establish a CFD code to simulate the vapor film flow field for this purpose. The result of pressure and velocity distribution around hot particle could be used to calculate the total force exerted on the particle and then predict its movement characteristics. In order to treat with the water-vapor free surface and rapid vaporization on it, an interface tracking method based on Volume-of-Fluid (VOF) with Piecewise-Linear Interface Construction (PLIC) is proposed. Some special treatment is taken to deal with the discontinuous velocity field due to rapid phase change. Because the interface is reconstructed during the solution, radiation heat transfer which is dominant in the whole process could be taken into account more easily. Continuum Surface Force (CSF) model is used for surface tension. Preliminary simulation results have been obtained. The results will be compared to experimental data later.
Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel
Authors: Wataru Masumi , Kato Hisashi , Kudo Satoshi , Oshima Naoko , Wada Koji , Narutaki Hirofumi
Published: September 2008

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility (Figure 1). Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack.
Best Estimate Methodology for Modeling Bubble Flows
Authors: Bolshov L. A. , Chudanov V. V. , Strizhov V. F. , Alekseenko S. V. , Meledin V. G. , Pribaturin N. A.
Published: September 2008

In this paper the numerical–experimental methodology of studying bubble flows is considered. The effective numerical CFD algorithms for 3D calculation of two-phase flows with explicitly chosen of interface boundary and calculation of surface tension forces are offered. The new high-precise experimental technique permitting to control and to describe 3D effects of bubble flows as average in time and space, so local instant is described. The main attention of this paper is focused on validation and verification of numerical model (before submitted at NURETH11 conference) as with usage of the tests, accessible from the literature and outcomes of experiments obtained in IT SB RAS (Novosibirsk).
BWR Steam Dryer for Extended Power Uprate
Authors: Le Moigne Yann , Andre´n Anders , Greis Ingemar , Kornfeldt Hans , Sundlo¨f Per , Sjunnesson Jan , Karlsson Anna-Karin
Published: September 2008

Steam Dryers for extended power up-rated conditions, based on a design with proven performance record, have been designed by Westinghouse with performance characteristics meeting all utility requirements. Extensive CFD analysis has been used to design the new dryer intended for the BWR 3000 plants. The analysis has utilized previous knowledge, acquired from experimental work, on steam flow characteristics in the BWR 3000 with its asymmetrically placed steam line nozzles. A design has been developed that is predicted to both decrease vibration levels and to solve a water-level measurement problem. An extensive experimental verification of the design is presently done.
Modeling of Non-Homogeneous Containment Atmosphere in the ThAI Experimental Facility Using a CFD Code
Authors: Babic´ Miroslav , Kljenak Ivo , Mavko Borut
Published: September 2008

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the simulation was to reproduce the non-homogeneous temperature and species concentration distributions in the ThAI experimental facility. A three-dimensional model of the ThAI vessel for the CFX4.4 code was developed. The flow in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also introduced. The calculated time-dependent variables together with temperature and concentration distributions at the end of experiment phases are compared to experimental results.
Corrosion Behaviors of Al-Steel-Sputtering-Treated Steel and SiC/SiC Composites in High Temperature LBE at Low Oxygen Concentration
Authors: Rivai Abu Khalid , Takahashi Minoru
Published: September 2008

Corrosion tests of Al and SS-304-sputtering-surface treated STBA26 (9Cr.1Mo.0.1Si) and SiC/SiC composites with BN (boron nitide) coating has been conducted in high temperature LBE of 700°C at low oxygen concentration of 6.8 × 10−7 wt% and the behavior was analyzed. The sputtering technique was used to protect the steel from corrosion. The thickness of sputtering-treated layer was 21.45 μm. All specimens were immersed in LBE in a pot for 1000 hours. The STBA26 (9Cr.1Mo.0.1Si) without surface treated were also tested for comparison with sputtering-treated steels. The results showed that sputtering-treated layer still remained on the base of STBA26. No penetration of LBE was observed in this layer. The layer could protect the steel from penetration of LBE. The result also showed that thin layer which contains aluminum oxide and chromium oxide was formed on the surface-treated layer, and it protected the base area. On the contrary, the penetration in base area was observed in the as received STBA26. In SiC/SiC composites, there appeared cracks in a thin surface area and LBE penetrated deeply into the material. The corrosion did not occur in this SiC/SiC composite in the high temperature LBE.
Goals, Requirements, and Design Implications for the Advanced High-Temperature Reactor
Authors: Forsberg Charles W.
Published: September 2008

The Advanced High-Temperature Reactor (AHTR), also called the liquid-salt-cooled Very High-Temperature Reactor (LS-VHTR), is a new reactor concept that has been under development for several years. The AHTR combines four existing technologies to create a new reactor option: graphite-matrix, coated-particle fuels (the same fuel as used in high-temperature gas-cooled reactors); a liquid-fluoride-salt coolant with a boiling point near 1400°C; plant designs and decay-heat-removal safety systems similar to those in sodium-cooled fast reactors; and a helium or nitrogen Brayton power cycle. This paper describes the basis for the selection of goals and requirements, the preliminary goals and requirements, and some of the design implications. For electricity production, the draft AHTR goals include peak coolant temperatures between 700 and 800°C and a maximum power output of about 4000 MW(t), for an electrical output of ∼2000 MW(e). The electrical output matches that expected for a large advanced light-water reactor (ALWR) built in 2025. Plant capital cost per kilowatt electric is to be at least one-third less than those for ALWRs with the long-term potential to significantly exceed this goal. For hydrogen production, the peak temperatures may be as high as 950°C, with a power output of 2400 MW(t). The safety goals are to equal or surpass those of the modular high-temperature gas-cooled reactor with a beyond-design-basis accident capability to withstand large system and structural failures (vessel failure, etc.) without significant fuel failure or off-site radionuclide releases. These safety goals may eliminate the technical need for evacuation zones and reduce security requirements and significantly exceed the safety goals of ALWRs. The plant design should enable economic dry cooling to make possible wider nuclear-power-plant siting options. Uranium consumption is to be less than that for a LWR, with major improvements in repository performance and nonproliferation characteristics.
Assessment of ICARE/CATHARE V1 Severe Accident Code
Authors: Chatelard Patrick , Fleurot Joe¨lle , Marchand Olivier , Drai Patrick
Published: September 2008

The ICARE/CATHARE code system has been developed by the French “Institut de Radioprotection et de Suˆrete´ Nucle´aire” (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the “International ICARE/CATHARE Users’ Club”, under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a “circuit” experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the standalone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version (“Users’ Guidelines”). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head.
Large Interface Simulation in Multiphase Flow Phenomena
Authors: Henriques Aparicio , Coste Pierre , Pigny Sylvain , Magnaudet Jacques
Published: September 2008

An attempt to represent multiphase multiscale flow, filling the gap between Direct Numerical Simulation (DNS) and averaged approaches, is the purpose of this paper. We present a kind of Large Interface (LI) simulation formalism obtained after a filtering process on local instantaneous conservation equations of the two-fluid model which distinguishes between small scales and large scales contributions. LI surface tension force is also taken into account. Small scale dynamics call for modelisation and large scale for simulation. Joined to this formalism, a criterion to recognize LI’s is developped. It is used in an interface recognition algorithm which is qualified on a sloshing case and a bubble oscillation under zero-gravity. This method is applied to a rising bubble in a pool that collapses at a free surface and to a square-base basin experiment where splashing and sloshing at the free surface are the main break-up phenomena.
Improving Outage Performance: Outage Optimization Process
Authors: LaPlatney Jere J.
Published: September 2008

Planned outage performance is a key measure of how well an Nuclear Power Plant (NPP) is operated. Performance during planned outages strongly affects virtually all of a plant’s performance metrics. In recognition of this fact, NPP operators worldwide have and continue to focus on improving their outage performance. The process of improving outage performance is commonly referred to as ‘Outage Optimization’ in the industry. This paper starts with a summary of the principles of Outage Optimization. It then provides an overview of a process in common use in the USA and elsewhere to manage the improvement of planned outages. The program described is comprehensive in that it involves managing improvement in both the Preparation and Execution phases of outage management.
“COSI”: The Complete Renewal of the Simulation Software for the Fuel Cycle Analysis
Authors: Boucher Lionel , Grouiller Jean-Paul
Published: September 2008

“COSI”, a software developed by the Nuclear Energy Direction at CEA, the French Atomic Energy Commission, is a code simulating a pool of nuclear electricity generating plants with its associated fuel cycle facilities. This code has been designed to study short, medium and long term options for the introduction of various types of nuclear reactors and for the usage of associated nuclear materials. It permits to study transition scenarios and gives due consideration to isotopic composition essentially of uranium, plutonium, minor actinides and some fission products.
Impact on Nuclear Waste Production of Different Fuel Cycle Strategies
Authors: Boucher Lionel , Grouiller Jean-Paul , Courtois Charles , David Sylvain , Maurin Matthieu
Published: September 2008

In the frame of the French law for the researches about waste management, different dynamic scenarios have been studied [1]. These scenarios are considering the French case and start from the present situation, which consists in a single stage of Plutonium recycling in PWRs. The scenarios described in this paper take into account two main options: Continuation of nuclear energy or phase out option.