In this work, a chemical kinetic modeling study of the high-temperature ignition and laminar flame behavior of Tetrahydrofuran (THF), a promising second-generation transportation biofuel, is presented. Stochastic Species Elimination (SSE) model reduction approach (Eldeeb and Akih-Kumgeh, Proceedings of ASME Power Conference 2018) is implemented to develop multiple skeletal versions of a detailed chemical kinetic model of THF (Fenard et al., Combustion and Flame, 2018) based on ignition delay time simulations at various pressures and temperature ranges. The detailed THF model contains 467 species and 2390 reactions. The developed skeletal versions are combined into an overall reduced model of THF, consisting of 193 species and 1151 reactions. Ignition delay time simulations are performed using detailed and reduced models, with varying levels of agreement observed at most conditions. Sensitivity analysis is then performed to identify the most important reactions responsible for the observed performance of the reduced model. Reaction rate parameter modification is performed for such reactions in order to improve the agreement of detailed and reduced model predictions with literature experimental ignition data. The work contributes toward improved understanding and modeling of the oxidation kinetics of potential transportation biofuels, especially cyclic ethers.

A quasi-dimensional (QD) simulation model is a preferred method to predict combustion in the gasoline engines with reliable results and shorter calculation time compared with multi-dimensional simulation. The combustion phenomena in spark ignition (SI) engines are highly turbulent, and at initial stage of the combustion process, turbulent flame speed highly depends on laminar burning velocity S L . A major parameter of the QD combustion model is an accurate prediction of the S L , which is unstable under low engine speed and ultra-lean mixture. This work investigates the applicability of the combustion model for evaluating the combustion characteristics of a high-tumble port gasoline engine operated under ultra-lean mixture (equivalence ratio up to ϕ = 0.5) which is out of the range of currently available S L functions initially developed for a single component fuel. In this study, the S L correlation is improved for a gasoline surrogate fuel (5 components). Predicted S L data from the conventional and improved functions are compared with experimental S L data taken from a constant-volume chamber under micro-gravity condition. The S L measurements are done at reference conditions at temperature of 300K, pressure of 0.1MPaa, and at elevated conditions whose temperature = 360K, pressure = 0.1, 0.3, and 0.5 MPaa. Results show that the conventional S L model over-predicts flame speeds under all conditions. Moreover, the model predicts negative S L at very lean ( ϕ ≤ 0.3) and rich ( ϕ ≥ 1.9) mixture while the revised S L is well validated with the measured data. The improved S L formula is then incorporated into the QD combustion model by a user-defined function in GT-Power simulation. The engine experimental data are taken at 1000 RPM and 2000 RPM under engine load IMEP n = 0.4–0.8 MPa (with 0.1 increment) and ϕ ranges are up to 0.5. The results shows that the simulated engine performances and combustion characteristics are well validated with the experiments within 6% accuracy by using the QD combustion model coupled with the improved S L . A sensitivity analysis of the model is also in good agreement with the experiments under cyclic variation (averaged cycle, high IMEP or stable cycle, and low IMEP or unstable cycle).

With the everlasting increase in the population, a huge surge in the electricity consumption can be noticed. Thus, the power and electricity generating power plants need to augment their performance to cope with this uprising problem. The main goal for most gas turbine power plants is to increase their efficiency and performance which can be achieved by increasing the turbine inlet temperature (TIT). However, increasing the TIT requires cooling of the turbine blades to extend its lifetime and avoid thermal stresses and oxidation rates. Usually, there are two routes to improve the turbine blade cooling, either scientist focus on the parameters that effect the cooling process such as the film cooling effectiveness, shape of holes and angle of injection, or the problem is approached from a thermodynamic point of view. It is well known that the air used to cool the turbine blades is bled from the compressor which causes a severe penalty on the thermodynamic efficiency and power output of the gas turbine. This paper main objective is to improve the gas turbine performance by lowering the temperature of the coolant lines bled from the compressor for turbine blade cooling resulting in a reduction in the amount of coolant mass flow rate required for turbine cooling which will reduce the penalty on the overall efficiency increasing it. For this purpose, three different configurations of Maisotsenko desiccant cooling systems were proposed to cool down coolant lines as well as the inlet air temperature. Optimization analysis was performed to determine the best operating parameters of the gas turbine as well as the cooling systems. Sensitivity analysis was conducted as well to investigate the effect of various variables on the gas turbine overall efficiency and the coolant mass flow rate. The results showed an increase in the overall efficiency from 42.57% to 43.83%, reduction in the amount of coolant mass flow rate that is bled from the compressor from 4.584 kg/s to 3.607 kg/s and in the cooling fraction from 4.72% to 3.9%.

Pipeline architecture consists in several stations in series configuration; hence the unavailability of one station impacts the availability of the whole pipeline. This lead to the need of optimizing the availability of each station in terms of configuration and number of units required in order to be able of satisfying the demand at any time. The loss of production cost in gas supply application is very high. Aero-derivatives gas Turbines are typically used as drivers in pipeline applications since they maximize train efficiency, minimizing gas consumption. PGT25+ aero-derivative Gas Turbines are among the most popular units applied in pipeline services. They merge demonstrated reliability performances together with a very limited outage duration impact that leads to very high Availability. Outage duration is optimized through modular replacement of both GG and HSPT that is facilitated by light weight of the machine. A Reliability Block Diagram has been built with the aim to optimize the Pipeline PGT25+ Gas Generator scheduled maintenance. Each block represents a Gas Generator while each station is realized taking into account the actual k-out-of-N configuration of each station units. Once the model has been created, a sensitivity analysis has been performed in order to estimate the impact of the Gas Generator cycle time (Gas Generator refurbishment time),that is what if larger or shorter than the baseline 6 months. Further, even a sensitivity study has been carried on to estimate the impact of the number of available spare parts on the delay that some units will suffer due to un-sufficient number of GG spare with consequent higher risk.

In order to understand how a boiler is performing/ operating, it is critical to obtain data throughout its operation. Data collection and storage methods have evolved through the years improving the quality and quantity of the data. Data is valuable for tracking current unit performance, troubleshooting and helping to narrow down any potential issues/ concerns with performance. Proper use of data collection and analysis may minimize the need for scheduled performance testing except when specific data points are required. This paper will discuss how sensitivity analysis can be utilized to determine the effect lack of/poor quality data has on the desired analysis. It discusses data collection and evaluation for various cases and the relevant ASME codes. Other key features of the paper are the various methods available for data representation, allowing the engineer to easily track key operating parameters.

The energy-water nexus is an area of increasing global concern and research. In several existing publications on the subject, the challenges of water use for power plant cooling and energy use for water supply are handled seperately. There is however also a need to consider the totality of interactions between the different elements of the engineered water and electricity systems, thus creating a system-of-systems model. A model of this form integrates water use for electricity supply and electricity use for water supply into a single framework, thus elucidating a wide range of interactions which can be influenced by policy and management decisions to achieve desired objectives. An engineering model capturing these interactions and based on first-pass models of the underlying physics of the various coupling and boundary points has been developed in previous work. In this work, the Jacobian of the resulting system of equations has been determined for a particular illustrative case. This Jacobian enables a sensitivity analysis of the inputs and outputs of this system-of-systems to changes in water and electricity demand to be carried out. As a concrete example, the Jacobian is used to examine the effect of a 10 % growth in both electricity and water demand on the set of system inputs and outputs.

The RCS flow measurement in PWRs is currently performed with a heat balance between primary and secondary systems: thus temperature heterogeneity impacts this measurement. In case of RCS flow shutdown (incident situation), a relative measurement of the RCS flow is monitored using an existing elbow tap. The goal of the present study is to assess the feasibility of using the existing plant elbow taps to accurately measure an absolute value of the RCS flow continuously and independently of the RCS temperature measurements. RCS flow is basically proportional to the square root of the differential pressure in the pipe elbow taps. Experiments on a scale model and sensitivity analyses with CFD simulations have been carried out and show that only few parameters have an influence on the proportionality coefficient. CFD is seen as able to predict this coefficient with an adequate accuracy. Potential applications of this method are RCS flow monitoring from start to full load after the change of primary coolant pumps or of steam generators.

Solar power generation technologies are categorized as Concentrated Solar Thermal Power (CSP) and PhotoVoltaic (PV). AREVA’s Compact Linear Fresnel Reflector (CLFR) system is a CSP power generation technology which compares favorably with other technologies in terms of its land efficiency and environmental impact. Analysis of the costs and benefits of solar technologies can inform their design and influence environmental and economic policies. This paper reports a comprehensive “cradle to grave” life cycle analysis (LCA) of AREVA’s CLFR technology. A unique element of this study is the availability of comprehensive inventory data from AREVA’s Reliance project, a 125 MWe Solar CLFR power plant under construction in India. Using actual project data showed the energy payback time was about 8.2 months and the greenhouse gas intensity was about 31 g-CO2/kWhe. Sensitivity analysis identified that the environmental performance is most sensitive to the solar intensity represented by direct normal irradiance. This study also compares AREVA’s CLFR technology with other leading solar power generation technologies. AREVA’s CLFR has the similar energy payback time and greenhouse gas intensity as other CSP technologies, and it has lower environmental impact compared to flat-plate PV systems.

Smart realization involves a steady increase in inverter-based components like Distributed Energy Resources (DER), energy storage systems, and plug-in electric vehicles. The harmonics related to DER inverters and the spread of power electronic devices raises concerns for utilities and customers. Harmonics can create component failures, thermal losses and control system malfunctions. In this paper the authors analyze the impact of multi-source harmonics from DERs inside distribution networks. The harmonics impacts are evaluated by harmonic measurement indices. Harmonic emission in a real distribution circuit is simulated with the help of power flow analysis. The results are presented with visualization techniques to give a better picture of harmonic propagation vs. different levels of harmonic source magnitude and angle. Due to effect of harmonic on network efficiency, a sensitivity analysis considering power factors is conducted.

Studies have been made on application of TRACG05, which is the latest version of TRACG code, to Best Estimate Plus Uncertainty (BEPU) approach for Anticipated Operational Occurrences (AOO) in BWR and ABWR. This paper presents validation results of TRACG05 against BWR and ABWR start-up test data. TRACG05 analyses of BWR and ABWR start-up tests, such as generator load rejection (pressurization event) and recirculation pump trip (core flow decrease event), were performed using standard plant nodalization and nominal inputs. In addition to the nominal cases, sensitivity studies and statistical analyses were conducted. The sensitivity studies include the sensitivities to TRACG05 model uncertainties, which have relatively high importance being tabulated in a Phenomena Identification and Ranking Table (PIRT), and nodalizaton. The sensitivity analyses include the random variation of TRACG05 model uncertainties, which have High ranking in PIRT, and initial core power. These uncertainties were preliminarily evaluated from comparisons of calculated results to separate effects test data, integral test data and component test data. TRACG05 calculated results of the nominal cases agree to the start-up test data with reasonable accuracy. For statistical analyses, the standard deviations of the key parameters related to the safety parameters (critical power ratio, etc) for AOO were calculated at each time step. The test data basically fall inside the 2 sigma bands. It indicates that the differences between calculated results and test data are within the uncertainties of TRACG05 model and initial core power that are taken into account for safety analyses in BEPU process. It is concluded that the TRACG05 code using the standard nodalization, inputs and model uncertainties is validated against the full-scale plant data and is capable of predicting the AOO transients.

Containment is the ultimate barrier which protects the radioactive substance from spreading to the atmosphere. Sensitivity analysis on AP1000 containment during postulated design basis accidents (DBAs) was studied by a dedicated analysis code PCCSAP-3D. The code was a three-dimensional thermal-hydraulic program developed to analyze the transient response of the containment during DBAs; and it was validated at a certain extent. Peak pressure and temperature were the most important phenomena during DBAs. The parameters being studied for sensitivity analysis were break source mass flow rate, containment free space, surface area and volume of heat structures, heat capacity of the containment shell, film coverage, cooling water tank mass flow rate and initial conditions. The results showed that break mass flow rate as well as containment free space had the most significant impact on the peak pressure and temperature during DBAs.

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

The purpose of seismic qualification of Structures, Systems and Components (SSCs) in nuclear power plants is to ensure that their intended safety function will not be compromised during and after a postulated earthquake event. The seismic performance of the equipment is generally evaluated using In-Structure Response Spectra (ISRS) at equipment-support locations as an input motion. Traditionally, these ISRS are generated based on design ground spectra prescribed by either U.S. Nuclear Regulatory Commission Regulatory Guide 1.60 or other design spectral shapes, which normally consider the frequencies content up to 33Hz. However, it has been recently recognized that probabilistic hazard-based site specific ground motion response spectra (GMRS) for Central and Eastern United States (CEUS) hard rock sites contains significant energy in the high frequency range, far beyond 33Hz. Since the motion at equipment support locations is highly affected by the dynamic characteristics of the soil or rock surrounding the building foundations and those of the structure itself, the adequacy of dynamic modeling and analysis techniques for determining the ISRS is critical to seismic qualification of safety-related equipment. This paper provides examples on dynamic modeling and analysis techniques required to accurately capture the structural responses for purposes of calculating ISRS throughout the frequency range of interest, including the high frequency responses typically expected at the CEUS sites. The discussion includes the selection of finite element mesh size, and sensitivity analysis performed to demonstrate that the propagation of these high frequencies through the different levels of the structure is properly captured. Other analytical considerations, such as the selection of time step size, for conducting time-history analysis, are also presented.

1000-MWe scale Pressurized Water Reactor (PWR) is taking service or under construction all over the world, and larger scale plant is studied and developed for its more competitive economics. Not only design basic accidents are analyzed for nuclear safety, the severe accident must also be considered to meet with the increasing requirement of safety. In the “nuclear power plant design safety regulation” (HAF102) issued by Nation Nuclear Safety Administration (NNSA), aim at the preventing and mitigating of severe accident, the regulation bring forward new requirement, which required that during design phase, NPP should consider setting the preventing and mitigation measurement of severe accident as actually as possible. As an approach to prevent the curium from melting down the vessel and entering the containment when a postulated severe accident occurs, In-vessel retention (IVR) of molten core debris via water cooling of the external surface of the reactor vessel has been introduced into AP1000. External reactor vessel cooling (ERVC) is assumed to be achieved keeping exterior surface of vessel at 400K. It is known to all that different scenario and process results in different IVR molten model. As the core melt, different IVR model is formed at different time, such as two-layer model, three-layer model and four layer model. It is necessary to study the IVR model when severe accident process moves on. This paper studies two-layer and three-layer IVR models and find the features of the models. Based on this, sensitivity study of important parameters has also been analyzed. It is useful for us to understand the mechanism of the molten pool. This paper has some directive significance on future IVR strategy research and also provides theoretical support to safety evaluation of PWR plants.

The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%. The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO 2 +BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MW th .

The ability to use FLUENT 12 or other CFD software to accurately model supercritical water flow through various geometries in diabatic conditions is integral to research involving coal-fired power plants as well as Supercritical Water-cooled Reactors (SCWR). The cost and risk associated with constructing supercritical water test loops are far too great to use in a university setting. Previous work has shown that FLUENT 12, specifically realizable k-ε model, can reasonably predict the bulk and wall temperature distributions of externally heated vertical bare tubes for cases with relatively low heat and mass fluxes. However, sizeable errors were observed for other cases, often those which involved large heat fluxes that produce deteriorated heat transfer (DHT) regimes, which involves a reduction in the rate of heat transfer between the wall and the fluid, and therefore a rise in wall temperature. These errors were believed to be caused by FLUENT’s over-estimation of changes in thermal physical properties of the fluid which occur around the pseudocritical point. The goal of this research is to gain a more complete understanding of how FLUENT 12 models supercritical water cases and where errors can be expected to occur. One control case is selected where expected changes in bulk and wall temperatures occur and they match empirical correlations’ predictions, and the operating parameters are varied individually to gauge their effect on FLUENT’s solution. The model used is the realizable k-ε, and the parameters altered are inlet pressure, mass flux, heat flux, and inlet temperature. As a result of this work, under supercritical conditions, the only finite limitations found in the k-ε model are in the mass flux which is predicted correctly above 300 kg/m 2 s, and in the temperature and related physical properties of water where they exceed the limitation of the fluid properties in the fluid database.

A study to characterize the steam waterhammer phenomena of a low pressure cooling injection (LPCI) system for a Mark 1 boiling water reactor (BWR) has been performed using RELAP5 and GOTHIC during a transient event. The scenario of particular interest was a manual switchover from shutdown cooling mode 3 to low pressure injection due to a loss of coolant accident (LOCA). This transient was initiated by opening the isolation valves of the two trains on a LPCI system into the torus. The torus was considered to be at atmospheric pressure and 20°C. The initial condition of the problem was set up such that the liquid was stagnant in the system. The initial temperature and pressure of the liquid, which was between the torus and isolation valves, was considered to be the same as the torus conditions. On the other hand, the initial condition of the liquid upstream of the isolation valves was chosen to be at 1 MPa and near saturation temperature. The analysis showed that the liquid in the system flashed into steam and discharged into the torus after the isolation valves started to open. Discharge of steam continued until the pressure in the LPCI system reached to a hydrostatic equilibrium with the torus. Following this, the cold liquid from the torus began to reflod the LPCI piping while condensing the steam at the liquid-steam interphase. This caused a mild steam waterhammer event when all of the steam condensed in the piping segments with closed ends. A sensitivity analysis showed that, the magnitude of the steam waterhammer predicted by both codes was sensitive to the number of nodes selected to model the piping, where the steam waterhammer phenomena occurred. Technical basis was obtained from the available literature and used as a guide to choose the number of nodes for the models in both codes. Once the steam waterhammer and the hydrodynamic properties associated with this transient were predicted by both codes, the forces exerted on critical pipe components were calculated. Also, selected thermal-hydraulic properties and hydrodynamic loads were compared between both code calculations. Comparisons showed reasonable agreements.

A RELAP5-3D input deck of the South Texas Project (STP) power plant was created in order to study the thermal-hydraulic behavior of the plant during normal operation (steady-state) and during a Loss of Coolant Accident (LOCA). It is important to study the sensitivity of selected output parameters such as the total coolant mass flow rate, the peak clad temperature, the secondary pressure, as a function of specific input parameters (reactor nominal power, vessel inlet temperature, steam generators primary side heat transfer coefficient, primary pressure etc.) in order to identify the variables that play a role in the uncertainty of the thermal-hydraulic calculations. RELAP5-3D, one of the most used best estimate thermal-hydraulic system codes, was coupled with DAKOTA, developed by Sandia National Laboratory for Uncertainty Quantification and Sensitivity Analysis in order to simplify the simulation process and the analysis of the results. In the present paper, the results of the sensitivity study for selected output parameters of the steady-state simulations are presented. The coupled software was validated by repeating one set of simulations using the RELAP5-3D standalone version and by analyzing the simulation results with respect of the physical expectations and behavior of the power plant. The thermal-hydraulic parameters of interest for future uncertainty quantification calculations were identified.

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1–3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI-2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the reactor pressure vessel will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues are included in the paper.

An economic and mathematical model for evaluating the nuclear electricity price is proposed, provided the reported costs of all production stages of the nuclear fuel cycle (NFC) are represented. As compared to known pricing models applicable for power generation (plant LCOE models, providing a least-cost generation alternative for complex market models), influences of uncertainties have been reduced due to representation of the integral nuclear-energy complex (NEC) as multiproduct and multiresource production system. The methodology exploits the Leontief’s interproduct balance model with specific matrix structure, and the technological features of a closed-loop NFC were factored in. The price properties of the modified Leontief’s model have been used for price evaluations. We presume the nuclear power plant (NPP) fixed assets estimates and the reliable prognostic data of the NEC performances are to be available on the annual basis. Conceivable variations of technological parameters can easily be employed to proceed with sensitivity analysis. The sensitivity of the nuclear electricity price with respect of fixed assets cost has been specified through the ‘production price’ basics and the defining value of the NPP capital cost share has been confirmed, as compared to the working capital impact including the fuel cost value.