bluuebunny/crossref_metadata_2025 huggingface dataset:
Towards Understanding Melt Eruption Phenomena During Molten Corium Concrete Interactions
Authors: Robb Kevin R. , Corradini Michael L.
Published: April 2011

A defense-in-depth feature for advanced light water reactors to cope with beyond design basis accidents is the ability to cool and stabilize ex-vessel core melt debris. Several international experimental programs have investigated core-concrete interactions and debris cooling of ex-vessel core melts. These experimental programs have identified various phenomena which affect melt coolability and may enhance it. One such phenomenon, melt eruptions, occurs when gas from the underlying decomposing concrete entrains melt up through and onto a solidified crust, which separates the molten melt from the cooling water. Previous modeling and experimental work have shown this cooling mechanism can have a large impact on melt coolability. Previous melt eruption models are reviewed and a new synthesis model is proposed. Reviewing past experimental evidence and modeling efforts indicate the geometry of the flow area impacts the amount of melt ejected. To understand the potential flow area available for melt eruptions, past experimental evidence is reviewed, a steady state analysis of flow area is performed and non-steady state considerations are discussed.
Irradiation-Induced Embrittlement of INCONEL 600 Flux Detectors in CANDU Reactors
Authors: Griffiths M. , Bickel G. A. , Douglas S. R.
Published: April 2011

The early design of flux detectors in use in CANDU® reactors consisted of multiple-paired-coiled detectors sheathed in INCONEL® 600 and wrapped around a Zircaloy carrier rod. The signal from these detectors was carried by INCONEL 600 coaxial cables that were threaded through the carrier-rod and fed back to instrumentation outside of the reactor core. Failures occurred in early applications that were related to the environment in which the detectors operated (moist or dry air). These failures were mitigated by ensuring that the detectors operated with a He cover gas that acts both to limit corrosion and as an efficient conductor of heat (generated by nuclear heating). An improved design was introduced consisting of single-individual-replaceable detectors (SIRs) with performance improved by eliminating brazed joints that were susceptible to nitric acid attack. Post-irradiation examination of encapsulated INCONEL 600 coiled flux detectors that had failed after several years of service in a CANDU reactor showed that many of the detector wires inside the carrier rod were broken. It was concluded that failure occurred because the He cover gas was lost and replaced by air. The examination also showed that those parts of the INCONEL 600 detector wires that had operated at the highest temperature, as evidenced by the degree of localised corrosion, were the most severely embrittled. Parts of the same detector wires that were clearly operating at lower temperatures, because they were shiny, remained ductile. Calculations indicate that if the INCONEL 600 material temperature exceeds about 573K (300°C) then it is conceivable that significant swelling and associated embrittlement could be expected. This is consistent with the observations indicating that the INCONEL detector wires that had experienced operation at elevated temperatures (>300°C) were embrittled.
Scenarios for the Transmutation of Actinides in CANDU Reactors
Authors: Hyland Bronwyn , Gihm Brian
Published: April 2011

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU® reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past [1–4]. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100 to 1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.
Development of Superfine Spherical Silica Grout as an Alternative Grouting Material for the Geological Disposal of Long-Lived Radioactive Waste
Authors: Naito Morimasa , Kishi Hirokazu , Fukuoka Naomi , Yamada Tsutomu , Ishida Hideaki
Published: April 2011

As an alternative grouting material for the geological repository of long-lived radioactive waste, the “Superfine Spherical silica Grout” (SFSG) material is developed using a fine spherical silica and a fine calcium hydroxide. The developed SFSG material takes an advantage of its smaller particle size distribution (max. ∼1 micron or less) than those of the cementitious materials, and also provides a low alkaline environment so as to reduce unfavorable effects on the long-term performance of geological disposal system. The SFSG is a mixture of the “super fine silica powder”, the superfine calcium hydroxide and additives such as superplasticizer. Presently, the mixture being investigated for grouting materials is focused on water/binder ratio (W/B) of 1.2. Some preliminary laboratory experiments were carried out to characterize its fundamental properties from the viewpoint of practical use for geological disposal, which is required to be equivalent with the conventional cementitious materials in terms of penetrability, strength, pH performance and workability. From a series of experiments, it was concluded that SFSG is expected to become an alternative grouting material for a geological repository.
Management and Practice of CPR1000 Nuclear Power Plants: Commissioning and Startup
Authors: Yang Yu , Can Wang
Published: April 2011

Nuclear power plants commissioning is an entire inspection and verification process of systems and equipments carried out by commissioning party, to prove that the overall performance of system, equipment and unit can meet the design requirements, operation criteria and nuclear safety criteria, and the plant is qualified for long-term stable and safe operation. As commissioning work has the characteristic of long lasting time, broad scope, high technology and lots of interfaces, and so on, it must be carried out under a set of scientific management system, to accord with requirements of “six controls” of commissioning work. The text aims at researching and investigating into commissioning management system that is in accordance with CPR1000 nuclear power plant commissioning through analysis of constitutes and practice situation of Ling AO phase II project commissioning management system, so as to instruct and standardize the commissioning management of nuclear power plants.
Efficiency of Sulfoaluminate Cement for Solidification of Simulated Radioactive Borate Liquid Waste
Authors: Sun Qina , Li Junfeng , Wang Jianlong , Ouyang Shixi , Li Qiang , Wu Minghui
Published: April 2011

To investigate the solidification efficiency of sulfoaluminate cement (SAC) and to provide more information for formula optimization, SAC blending zeolite, accelerator and Dura fiber was used as matrix materials for solidification of simulated radioactive borate liquid waste. The simulated radioactive borate liquid waste was prepared with boric acid and sodium hydroxide using drinking water. The performances of solidified waste forms were evaluated mainly basing on matrix compressive strength and leachability. The 28d compressive strength of the solidified waste forms were tested according to Chinese National Standard GB 14569.1-1993, and experiments on water/freezing/irradiation/impact resistance were also carried out. Nuclides Sr, Cs and Co were substituted by their non-radioactive isotopes respectively in leachability test, and the testing procedures were consistent with Chinese National Standard GB 7023-1986. Experimental results showed that it was feasible to solidify borated liquid wastes with SAC. The 28d compressive strength was 13.9MPa, nearly twice of the standard in GB 14569.1-1993. Strength losses in water/freezing/irradiation/impact resistance tests met the demands of GB 14569.1-1993 well. In the leaching test, the 42d leaching rates were 3.39×10−5 cm/d, 4.45×10−5 cm/d and 4.07×10−7 cm/d for Sr2+, Cs+ and Co2+ respectively, much lower than GB 14569.1-1993 limits. Results of leaching test also showed that the leaching mechanism of Co2+ was different from that of Sr2+ and Cs+.
Application of the Advanced Construction Technologies to Shimane Nuclear Power Station Unit No.3 of the Chugoku Electric Power Co., Inc.
Authors: Inoue Takashi , Fushiki Katsumi , Oka Akifumi , Murayama Kouichi
Published: April 2011

Shimane Nuclear Power Station Unit No.3 (hereinafter NS-3) with 1,373 MW of maximum electric power being supplied to The Chugoku Electric Power Co., Inc. (Hereafter The Chugoku EPCO) is now under construction together with Hitachi-GE Nuclear Energy (HGNE). The groundbreaking (first establishment in construction business permission) was held in Dec. 2005, and is aimed at starting operation in Dec. 2011. HGNE finished installing the Reactor Pressure Vessel (RPV), the heaviest equipment in the plant in July 2009. HGNE also installed the shroud, one of the most important equipment in Aug. and a giant crawler crane was used for the installation. Regarding electrical equipment, HGNE installed a central control panel into the Control Building (C/B). The installation of main products moves ahead one after another, and the construction is going smoothly. Currently, the main construction has been proceeded to mechanical and electrical work by HGNE from building one by the building contractors. This paper reports the construction status of NS-3 and introduces applied construction technologies and construction management methods.
Experimental Studies on Pool Boiling Characteristics of Titanium Dioxide-Water Nano-Fluids
Authors: Okawa Tomio , Kamiya Takahito
Published: April 2011

It is known that dispersion of a small amount of nanometer-sized particles in liquid can cause substantial improvement of the critical heat flux in pool boiling. Nanofluids (colloidal suspensions of nanoparticles in a base fluid) may therefore be used as the coolant in industrial applications in which high-heat-flux removal is needed. If it is supposed that the deposition of nanoparticles onto the heated surface during nucleate boiling is the main cause of the CHF enhancement in nanofluids, a certain time period is considered to be necessary for the CHF to be improved. In view of this, preliminary experiments were performed in the present work to investigate the time scale of CHF improvement; here, distilled water was used as a base fluid, and TiO2 and copper were selected as the materials of nanoparticles and heated surface, respectively. Under a particular experimental conditions of nanoparticle concentration and nucleate boiling heat flux (40 mg/l and 500 kW/m2), an approximate time scale of CHF improvement was 10 min; this value might not be negligibly short in some nanofluid applications. The measured time-variations of the wall superheat during the nucleate boiling in nanofluid suggested that longer time periods are required for the CHF enhancement at lower heat fluxes and lower nanoparticle concentrations. In particular, 40 min was not sufficient for the wall superheat to reach a steady-state value at the lowest nanoparticle concentration of tested in this work (9 mg/l).
Calculation of Fuel Burn Up and Radioactive Inventory for HEU Fuel Element of Tehran Research Reactor
Authors: Hosseini Seyed Abolfazl , Vosoughi Naser , Hosseini Mohammad
Published: April 2011

This paper presents a new approach for fuel burn up evaluation and radioactive inventory calculation used in Tehran Research Reactor. The approach is essentially based upon the utilization of a program written by C# which integrates the cell and core calculation codes, i.e., WIMSD-4 and CITVAP, respectively. Calculation of fuel burn up and radioactive inventories has been done for 26 core configuration of Tehran Research Reactor with HEU fuel element. The present inventory and fuel enrichment of each fuel element have been calculated.
Analysis of Spray Tests in a Multi-Compartment Geometry Using the GOTHIC Code
Authors: Andreani Michele , Erkan Nejdet
Published: April 2011

The effect of spray on the gas mixing in the multi-compartment geometry of the PANDA facility is investigated using the GOTHIC code. Two tests using two connected vessels are analysed, where water was injected from a full-cone spray nozzle placed in the upper part of one vessel. In the vessel where the spray was located, the upper region was filled with a helium-rich mixture. The remaining volume of the vessel and the second vessel were filled with a denser fluid. The difference between the two tests was the presence of air: one test was performed with steam and helium only, whereas in the second test air was also present in the gas mixture in both vessels. These experiments have been analysed with the advanced containment code GOTHIC. The results showed that the code tends to overpredict the depressurisation rate when the standard correlations for natural convection heat transfer between fluid and structures are used. With respect to mixing, generally good results were obtained. The timing of the stratification break-up above the nozzle injection, however, could not be reproduced. Finally, the large effect of the initial mixture composition on the evolution of the gas distribution in the connected vessel was well predicted.
Advanced Construction Technologies for the Ohma Nuclear Power Plant Reactor Building of Electric Power Development Co., Ltd.
Authors: Obata Tatsuya , Urashima Akihito , Watanabe Kiyokatsu , Miyahara Tsumoru
Published: April 2011

Electric Power Development Co., Ltd has been constructing Ohma Nuclear Power Plant aiming to start commercial operation in Nov. 2014. Ohma Nuclear Power Plant is located in Ohma-town, Aomori Prefecture and is a landmark power plant in which Mixed Oxide fuels can be loaded in the full core of the reactor. Hitachi-GE Nuclear Energy Ltd. and Kajima JV, both have extensive experience of nuclear power plant construction, are the main contractors of this project and supply the entire engineering, manufacturing of all major components, and execute the construction and commissioning for the reactor building. Ohma-town is located at the northernmost part of Aomori Prefecture bordering Tsugaru strait, where is exposed to severe cold and constant strong wind in winter. Such severe weather conditions make the construction very hard, however, Hitachi and KAJIMA tries to complete the project on schedule and on budget applying highly reliable advanced construction technologies, such as open-top and parallel construction method, all whether construction method, and large scale modularization technology. The groundbreaking (acquisition of the first construction permission) was already completed in May 2008. Its civil work steadily progressed, and the rock inspection was completed in Oct. 2009. Base mat will be completed in July 2010, and both building work and mechanical work go into full swing after installation of RCCV lower liner module.
Study on Applicability of Simple Conversion Factor for In-Situ Activity Measurement
Authors: Onishi Yuko , Kawasaki Satoru , Kato Masami
Published: April 2011

In Japan, the clearance system has been in effect since December 2005, and at present, the cleared material is recycled at a site in affiliation with the nuclear installation. In the near future, it is assumed that a considerable amount of cleared material will be transported from nuclear facilities that are being decommissioned, so that the material can be recycled and used by general industry. In order to operate the clearance system with certainty, it is essential to devise a method for promptly measuring the radionuclide concentration of cleared materials that are suspected of exceeding the clearance level. That is, preparation for the contingency that an uncertain radiation source mixes with a clearance item by chance. In general, in-situ measurement of low-level radionuclide concentration is difficult since accurate information on the shape of the target, as well as on the distribution of radionuclide concentration in the target, is required. Therefore, we investigated a method for simple and rapid determination of radionuclide concentration, which does not require accurate information on the volume and density of the target. To avoid the influence of volume or density, we examined the adoption of a conversion factor with units of “(Bq/g)/cps”. ISOCS (In-situ Object Counting System) Calibration Software (Model S573 Version 4.0, CANBERRA Inc.) was used to determine the conversion factor. Metal or concrete was assumed to be stored and piled in a stockyard of a recycling facility. The piled objects were arranged as a cone with a gradient of 50%, in accordance with the limitations specified by the “Waste Management and Public Cleansing Law”. The volume and bulk density were assumed to be 0.1–10 m3 and 0.5–10 g/cm3, respectively. The conversion factor for a cone with volume of 1 m3 and density of 1 g/cm3 could be applied to the piled objects having a volume greater than or equal to 1 m3 and a density greater than or equal to 0.5 g/cm3, resulting in an increase in tolerance by a factor of two. These results show that in-situ measurements of radionuclide concentration can be simplified by using a conversion factor with units of “(Bq/g)/cps”. This approach can be used to carry out prompt measurements on suspected materials in the field.
The Prediction of Long-Term and Emergency Sealability of Silicone and EPDM Gaskets
Authors: Nakano Mitsuyuki , Sasaki Haruko , Hanashima Kanji
Published: April 2011

At nuclear power plants, containment vessel acts as pressure barrier in such an emergency as a loss-of-coolant accident. It is important as safety equipment to prevent radiological from leaking outside. Rubber gaskets, which are used for sealing faces of containment vessel, are needed to maintain certain sealability not only in stable condition but also in an emergent situation. Among important characteristics of the rubber gasket are not only physical property changes after general aging test (for example, tensile strength changes after heat resistance test) that indicate long-term stability of gasket itself, but also after radiation resistance test which gives potential to good substitute characteristic in terms of sealability in such cases. Physical property changes after general aging test do not always substitute sealability in an emergency, because they do not reflect effects of radiation. That is why nuclear power plant engineers must choose suitable rubber materials that have high performance in radiation resistance. In Japan, silicone rubber gaskets have been used for containment vessel for a long time since 1970th, but in the United States, ethylene propylene diene terpolymer (EPDM) gaskets have been widely used. NICHIAS has silicone rubber and EPDM materials for containment vessel and these gaskets have been used in nuclear power plant. But it is not obvious why different materials have been used in two countries because few relative comparisons of the two materials have been carried out. Especially silicone rubber and EPDM gaskets have many combinations of chemical compositions, so it is difficult to evaluate gasket suitability for containment vessel. There are many kinds of studies concerning long-term stability and life of gasket, but we all must know what characteristics relate sealability under radiation exposure condition and are suitable for guidepost of sealability under previous condition. This report compares silicone rubber gasket with EPDM gasket on physical property changes under irradiation and thermal treatment. We report compression set test results about one each type of silicone rubber and EPDM gasket under irradiation from Co60. And we also report relationships in physical property changes between irradiation and thermal treatment. Finally NICHIAS predicts long-term and emergency sealability of these gaskets from the results of evaluation. We hope it will be a part of design guideline of rubber gasket for containment vessel.
Nuclear Analysis During Modification of NPP Core
Authors: Slavickas Andrius
Published: April 2011

Reactor power and neutron activity control is the main key for safe reactor operation. Reactivity coefficients and effects are main measures to estimate reactor control and safety. These characteristics outline reactors behavior during usually exploitation and accident events. Reactivity coefficients and effects quantify the effect, which various parameters (e.g. fuel and graphite temperatures, amount of steam) have for the core neutron activity. Many modifications of RBMK-1500 reactor cores in Ignalina NPP were made during their lifetime. Reactor core modifications like load of higher enriched fuel with burnable absorber and new design control rods affected reactivity coefficients and effects. Neutron-physical parameters calculations of reactor core states with variant fuel loads and new design control rods were performed using QUABOC/CUBBOC-HYCA software. The changes of reactivity coefficients and effects were quantified in this paper.
Study on Subcooled Flow Boiling Heat Transfer Under Vibration Conditions of Sinusoids and Seismic Shock Waves
Authors: Ohtake Hiroyasu , Mazume Takehiko , Koizumi Yasuo
Published: April 2011

Onset of nucleate boiling — ONB — and critical heat flux — CHF — on subcooled flow boiling under vibration condition, focusing on liquid velocity, amplitude and frequency of vibration, were investigated experimentally. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. The test section was set on a vibration table. The acceleration was 1.3 and 4.2 m/s2, respectively; the frequency was 2, 4 and 20 Hz, respectively. The test section was arranged for horizontal position facing upward and for vertical position, respectively: the vibration directions were horizontal and vertical to orientation of the heater. The present experimental results showed that critical heat fluxes under the vibration conditions of both vibration directions, i.e., horizontal and vertical vibrations to the heater, were higher than those for steady flow. The CHF under the vibration condition was increased with an increasing of acceleration of vibration. According to present observations, coalesced bubble on the heater was frequently released by vibration of the test heater. This behavior causes the CHF to become higher under the vibration condition. Furthermore, some experiments were demonstrated under vibration condition of seismic shock wave at the South Hyogo prefecture Earthquake in 1995. It was confirmed from the present experimental results that thermal limit of heater was a conservative evaluation during earthquake.
Conceptual Design of European Advanced Boiling Water Reactor (EU-ABWR)
Authors: Kamei Kazuhiro , Yamazaki Hiroshi , Hamazaki Ryoichi , Shimizu Takeo
Published: April 2011

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR has a large power generation capacity 1600MWe and the safety features meets European standards and regulation. This paper describes the conceptual design of EU-ABWR.
Analysis of Updated SuperCritical Water Heat Transfer Correlations for Vertical Bare Tubes
Authors: Mokry Sarah , Gupta Sahil , Farah Amjad , King Krysten , Pioro Igor
Published: April 2011

In support of developing SuperCritical Water-cooled Reactors (SCWRs), studies are currently being conducted for heat-transfer at supercritical conditions. This paper presents an analysis of heat-transfer to SuperCritical Water (SCW) flowing in bare vertical tubes as a first step towards thermohydraulic calculations in a fuel-channel. A large set of experimental data, obtained in Russia, was analyzed. Two updated heat-transfer correlations for forced convective heat transfer in the normal heat transfer regime to SCW flowing in a bare vertical tube were developed. It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) with high coolant temperatures (350–625°C). Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. The analyzed experimental dataset was obtained for SCW flowing upward in a 4-m-long vertical bare tube. The data was collected at pressures of about 24 MPa for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values for mass flux ranged from 200–1500 kg/m2s, for heat flux up to 1250 kW/m2 and inlet temperatures from 320–350°C. The Mokry et al. correlation was developed as a Dittus-Boelter-type correlation, with thermophysical properties taken at bulk-fluid temperatures. Alternatively, the Gupta et al. correlation was developed based on the Swenson et al. approach, where the majority of thermophysical properties are taken at the wall temperature. An analysis of the two updated heat-transfer correlations is presented in this paper. Both correlations demonstrated a good fit (±25% for Heat Transfer Coefficient (HTC) values and ±15% for calculated wall temperatures) for the analyzed dataset. Thus, these correlations can be used for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for SCW heat exchangers, for future comparisons with other independent datasets and for the verification of computer codes for SCWR core thermohydraulics.
Investigations on the Impact of Heavy Nuclides Resonance Elastic Scattering Models on Resonance Integrals
Authors: Peng Sitao , Zhang Shaohong , Jiang Xiaofeng
Published: April 2011

In today’s cross section data processing process, asymptotic scattering model is employed by NJOY for the neutron/nucleus elastic scattering interactions in the epithermal energy region, which means that the energy of a scattered neutron is always lower than its incident energy and it falls evenly within the interval of [αE, E]. This model has recently been proved to have non-ignorable errors at some resonances of heavy nuclides. In this study, to investigate the impact of heavy nuclides resonance elastic scattering models to the resonance integrals, exact scattering kernel is employed and a deterministic code Estuary is developed to efficiently solve the neutron slowing down problem. Numerical results demonstrate that with the use of Estuary, results given in the literature obtained by the Monte Carlo method can be reproduced. With the resonance cross section approximately represented by the single-level Breit-Wigner formulation, investigations are made for different resonance parameters for both asymptotic and exact scattering models. Relations between errors and these related parameters are summarized.
Analyses of Single-Phase Heat Transfer and Onset of Nucleate Boiling in Rod Bundles
Authors: Pe´turaud P. , Salko R. K. , Bergeron A. , Avramova M. N.
Published: April 2011

Thermal-Hydraulic (T/H) core code prediction of the existence and localization of boiling zones is crucial in the framework of axial offset anomaly (AOA) risk assessment of PWR cores. In this prospect, an experimental program — NESTOR — has been completed by Commissariat a` l’Energie Atomique (CEA, France), Electricite´ de France (EDF, France) and Electric Power Research Institute (EPRI, USA). The aim of the NESTOR program has been to develop an accurate prediction model for the onset of nucleate boiling (ONB) boundary in a nuclear fuel bundle based on an ONB wall superheat criterion associated with a dedicated single-phase heat transfer model. The experimental scope of NESTOR program involved using two loops to measure axial velocities in sub-channels and heater rod surface temperatures, respectively, in identical 5×5 rod bundles. The first set of experimental measurements were devoted to a bundle configuration containing only simple support grids (SSG) in order to resemble, as closely as possible, a bare rod bundle. Test data analyses in this configuration have since been jointly carried out by the three NESTOR partners, each using its own T/H core code (FLICA IV for CEA, THYC-COEUR for EDF and VIPRE-I for US Penn State University on behalf of EPRI). This paper describes the analyses results and conclusions based on SSG configuration data. The data analyses methodology consisted of three successive stages: (i) T/H core code calibration — determination of specific input data related to the bundle configuration and required by further core code simulations of the tests; (ii) Single-phase heat transfer analysis — development of dedicated single-phase heat transfer models using single-phase test data along with sub-channel-averaged temperatures and velocities obtained from T/H code simulations. The heat transfer models were unique for each of the three codes and included both a heat transfer correlation and grid enhancement correction factor; (iii) ONB test analysis — assessment of an ONB wall superheat criterion based on stage (ii) models and, if necessary, development of a new ONB wall superheat criterion. The analyses showed that while the heat transfer models could correctly represent the single-phase test data, the ONB wall superheat is over-predicted by 1–3.5 K when compared to experimental values and open literature wall superheat correlations. However, the actual impact of this inconsistency on prediction of ONB boundary localization in this experimental SSG bundle configuration is low. Similar concurrent data analyses for tests on a mixing vane grid bundle configuration are under progress and results should be available by the end of 2010.
Computational Simulation of Natural Convection of a Molten Core in Lower Head of a PWR Pressure Vessel
Authors: Vieira Camila B. , Su Jian
Published: April 2011

The paper reports computational simulation of natural convection of a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel. Numerical simulations were carried out for a two-dimensional semi-circular slice geometry which represents the lower head of a PWR pressure vessel. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate is simulated by using a commercial computational fluid dynamics (CFD) software ANSYS CFX 12.0. The bottom circular surface and top flat surface are both kept at a same constant temperature. The Boussinesq model is used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (shear stress transport) turbulent model is used to model the turbulent eddy viscosity. Two Prandtl numbers are considered, 6.13 and 7.0. Five Rayleigh numbers were simulated for each Prandtl number: 109, 1010, 1011, 1012, and 1013. The average Nusselt numbers on the bottom surface calculated by the computational simulations are in excellent agreement with Mayinger et al. (1976) correlation, with the average Nusselt number on the top flat surface calculated by the computational simulation is in excellent agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations only at Ra = 109. Numerical results of the velocity and temperature fields show the α-phenomena at lower Rayleigh number and ν-phenomena at lower Rayleigh number.
CFD Modeling of Subcooled Boiling in Vertical Bubbly Flow Condition Using ANSYS CFX 12
Authors: Pellacani Filippo , Mestre Silvana Matturro , Vicent Sergio Chiva , Juan Rafael Macian
Published: April 2011

Subcooled boiling in upward non-isothermal turbulent bubbly flow in tubes is numerically modeled using ANSYS-CFX 12 in this contribution. The approach is based on the RPI wall boiling model developed by Kurul and Podowski [1]. The interfacial non-drag forces are also investigated and included in the model. The Antal [5] model with different coefficients is used for the wall lubrication force. The lift force was calculated in two ways, with the Tomiyama model [4] and based on a constant value. The void fraction axial profiles for high pressure subcooled boiling in tubes are compared against the experimental data of Bartolomej [17] [18]. The pressure varies from 3 up to 6.89 MPa and the wall heat flux from 0.38 to 1.2 MW m−2. The mass flux range is between 900 and 1500 kg s−1m−2. The computational domain has up to 70000 nodes, and represents one eighth of a vertical pipe. In general, the models give predictions in close agreement with experimental results if the axial mean values are considered. The main difficulties for the simulation are observed for flow in transition to flow regimes with high void fraction, when the bubbly flow is not able to maintain the spherical condition of the bubbles, which is a requirement of the boiling models.
An Experimental Investigation Into Potential Nuclear Reactor Vessel Flow Instabilities
Authors: Brown Jeffrey A. , Rowland James W. , Fernando H. Joseph
Published: April 2011

An investigation into the increase in Plant Protection System (PPS) alarms at a three-unit US Pressurized Water Reactor (PWR) plant has determined that the alarms are the result, in part, of a hydraulic instability that has developed within the Reactor Coolant System (RCS) following the replacement of the steam generators in all three units of the Palo Verde Nuclear Generating Station (PVNGS). An experimental effort has been established by Arizona Public Service Company and Arizona State University in an attempt to determine the cause of these instabilities. Preliminary investigations have determined that the time scale of these instabilities is consistent with larger scale transient flow processes of the reactor vessel. Accordingly, the flow characteristics were assessed and localized flow measurements made using a one-fifth scale physical model of the upper plenum region of the reactor core of the Combustion Engineering System 80 reactor vessel to verify the postulation that large vortex structures referred to as “precessing” vortices [Ref. 1] affect the core exit flow conditions resulting in the noted flow instabilities. The physical model investigation was complemented by numerical analysis based on a Computational Fluid Dynamics (CFD) code performed for the same geometry. Benchmarking of the CFD model by the scaled physical model is intended to provide increased confidence in the CFD code. If verified, the CFD code may be modified so as to establish corrective actions for this condition, where physical modeling would probably be time consuming and cost prohibitive. The initial results for the physical and computational models demonstrate very good agreement between the measured and calculated flows in the upper-plenum region. The results of the complementary experimental and analytic evaluations do not support the presence of any large scale vortices of appropriate space scales that could affect flow conditions within the upper-plenum region. The elimination of the reactor vessel as the source of the instabilities suggests that the replacement steam generators may be the root cause of the flow instabilities. There is a possibility, however, that frequencies pertinent to vortices may be triggering mechanisms for flow instabilities in the entire system.
Development of Buildup Factors for Updating the ANSI/ANS-6.4.3 Standard
Authors: Sanders Charlotta E.
Published: April 2011

The ANSI/ANS-6.4.3-1991 Standard, Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, is currently being updated by an American Nuclear Society (ANS) Working Group. The ANSI/ANS-6.4.3, 1991 standard, which is “withdrawn” due to the failure to meet the requirement of the ANS to have standards updated every ten years, contains buildup factor values that are derived from data that is over seventeen years old. In addition, computer technology has significantly improved since 1991, allowing for more complicated, computationally demanding codes to be utilized. Therefore, the ANSI/ANS-6.4.3-1991 standard is being re-visited to include updated data and to include modern codes. Gamma-ray buildup factors and attenuation coefficients are being generated for common shielding materials (e.g., concrete, steel, water, etc.) utilizing the ENDF/B-VI.8 cross-section data library, which is distributed by Brookhaven National Laboratory’s National Nuclear Data Center (NNDC). One modern code, Monte Carlo N-Particle (MCNP5/MCNPX), is being compared with PALLAS-1D (VII) and Anisotropic Source-Flux Iteration Technique (ASFIT-VARI), which were used to develop values included in ANSI/ANS-6.4.3-1991. PALLAS-1D (VII) is a code for direct integration of transport equation in one-dimensional plane and spherical geometries while ASFIT-VARI is a gamma-ray transport code system for one-dimensional finite systems. MCNP5 is a general purpose Monte Carlo radiation transport code that tracks particles (e.g., neutron and photons) at numerous energies in a three dimensional configuration of materials. The MCNP5 radiation transport code require response function input to provide dose and exposure output. Mass energy-absorption coefficients and mass energy-transfer coefficients are required to develop absorbed dose and exposure responses. The National Institute of Standards and Technology (NIST) provide and maintain the most up-to-date mass energy-absorption coefficient and mass energy-transfer coefficient database currently available. The NIST values are used in these initial buildup factor calculations to prove the validity of the methodology used and allow for preliminary comparisons. The energy absorption (dose in material) and exposure buildup factors are calculated at mfp values of interest, consistent with ANSI/ANS-6.4.3-1991 up to forty mfp. Comparisons between the new buildup factors and the previous results presented in the ANSI/ANS-6.4.3-1991 standard indicate that there is fairly good agreement. Differences in buildup factor values can be attributed to differences in cross-section data libraries, numerical methods, and physics treatments within the respective codes.
Development of Advanced Two-Fluid Model for Boiling Two-Phase Flow in Rod Bundles
Authors: Yoshida Hiroyuki , Hosoi Hideaki , Suzuki Takayuki , Takase Kazuyuki
Published: April 2011

Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as rod bundles in nuclear reactors. However, two-fluid model include a lot of constitutive equations. Then, applicability of these constitutive equations must be verified by use of experimental results, and the two-fluid model has problems that the results of analyses depend on accuracy of constitutive equations. To solve these problems, we have been developing an advanced two-fluid model. In this model, an interface tracking method is combined with the two-fluid model to predict large interface structure behavior accurately. Interfacial structures larger than a computational cells, such as large droplets and bubbles, are calculated using the interface tracking method. And droplets and bubbles that are smaller than cells are simulated by the two-fluid model. Constitutive equations to evaluate the effects of small bubbles or droplets on two-phase flow are required in the advanced two-fluid model as same as a conventional two-fluid model. However, dependency of small bubbles and droplets on two-phase flow characteristic is relatively small, and the experimental results to verify the equations are not required much. In this study, we modified the advanced two-fluid model to improve the stability of the numerical simulation and reduce the computational time. Moreover, the modified model was incorporated to the 3-dimensional two-fluid model code ACE-3D. In this paper, we describe the outline of this model and the modification performed in this study. Moreover, the numerical results of two-phase flow in various flow conditions.
Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID
Authors: Cho H. K. , Yun B. J. , Park I. K. , Jeong J. J.
Published: April 2011

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.
Unstructured Adaptive Mesh Technique for Gas-Liquid Two-Phase Flows
Authors: Ito Kei , Kunugi Tomoaki , Ohshima Hiroyuki
Published: April 2011

In the design study of large-sized sodium-cooled fast reactors in Japan (JSFR), the suppression of gas entrainment (GE) phenomena at a free surface in the reactor vessel is very important to establish an economically superior design. However, the GE phenomena are highly non-linear and too difficult to be evaluated theoretically. Therefore, we are developing high-precision CFD method for gas-liquid two-phase flows to evaluate the GE phenomena accurately. To reproduce the GE phenomena by CFDs, there are three key issues, i.e. geometry dependency, interfacial dynamics and locality. Former two issues are already addressed by employing unstructured mesh schemes and a high-precision simulation method for gas-liquid two-phase flow based on the PLIC (Piecewise Linear Interface Calculation) method, respectively. In fact, the simulation results of the GE phenomena in a simple GE experiment showed good agreements with experimental data. Recently, therefore, we focus on the locality of the GE phenomena. In our previous study (presented in ICONE17), the two-dimensional unstructured adaptive mesh technique for single-phase flows was developed to address the third issue. The isotropic cell refinement method was employed and the connection cell method was proposed to eliminate the edge incompatibility. The verification/validation results showed that the developed unstructured adaptive mesh technique succeeded in providing a high-precision solution, even though a poor-quality distorted mesh at the initial state was employed. In this study, the unstructured adaptive mesh technique is extended to the numerical simulations of gas-liquid two-phase flows. The redistribution methods of two-phase flow variables are newly developed to satisfy the conservations of the variables, i.e. the volumes of gas and liquid phases, the location of interfaces and the momentum of each phase. This improved unstructured adaptive mesh technique for gas-liquid two-phase flows is validated by solving the well-known slotted disk revolution and dam-break problems. As a result, the unstructured adaptive mesh technique succeeds in maintaining the slotted-disk shape after one revolution and shows more than first order accuracy (grid convergence) in the slotted-disk revolution problem. In addition, thanks to the momentum-conservative formulation, the dam-break phenomenon is well simulated by the unstructured adaptive mesh technique. Especially, wave-breaking phenomena are simulated by refined cells near the gasliquid interface. It should be noted that these simulation results are obtained by using relatively small number of cells because of the efficient mesh adaptation by the unstructured adaptive mesh technique.
Rising Velocity of a Gas Plug in Vertical-Rectangular-Narrow Channels
Authors: Miyashita Tohru , Koizumi Yasuo , Ohtake Hiroyasu
Published: April 2011

Bubble rising velocity in stagnant water in rectangular channels was examined. The width of the flow channels and the gap space between parallel walls were varied from 10 mm through 150 mm and from 1 mm through 10 mm, respectively. When the bubble had plug shape in the long side and also the short side, the bubble velocity took the same velocity as that in a circular pipe that had the same periphery. When the bubble lost the plug shape in the long side, the rising velocity became fast as the long side shape departed from the plug shape. When the long side was large enough for the bubble to have the shape of a bubble in open space, the bubble rising velocity was expressed well with the expression for the bubble rising velocity in open space. As the long side became narrow, the bubble rising velocity became slower than that for open space. When the gap spacing was quite narrow; 1 mm, and the long side was less than 15 mm, the bubble stopped rising halfway in the flow channel. When the bubble took the shape of a plug, the bubble rising velocity did not depend on the film thickness between the bubble and the long-side wall in the present experimental range.
The Summarization of Power Coordinated Control of Reactor and Turbine
Authors: Dong Xiaosheng , Li Xiaofeng , Liu Guangming , Tian Yajie , Xu Guobin , Huang Weijun , Jiang Guojin
Published: April 2011

This paper describes the power coordinated control system in PWR NPP. The emphasis of this paper is to describe the key parameters selection, several coordinated control solution and operation mode selection. The some factors should be considered during control system design are also introduced in this paper.