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A Parametric Study of the Impact of the Cooling Water Site Specific Conditions on the Efficiency of a Pressurized Water Reactor Nuclear Power Plant  [PDF]
Mohamed M. A. Ibrahim,Mohamed R. Badawy
International Journal of Nuclear Energy , 2014, DOI: 10.1155/2014/569658
Abstract: In this study, the thermal analysis for the impact of the cooling seawater site specific conditions on the thermal efficiency of a conceptual pressurized water reactor nuclear power plant (PWR NPP) is presented. The PWR NPP thermal performance depends upon the heat transfer analysis of steam surface condenser accounting for the key parameters such as the cooling seawater salinity and temperature that affect the condenser overall heat transfer coefficient and fouling factor. The study has two aspects: the first one is the impact of the temperature and salinity within a range of (290?K–310?K and 0.00–60000?ppm) on the seawater thermophysical properties such as density, specific heat, viscosity, and thermal conductivity that reflect a reduction in the condenser overall heat transfer coefficient from 2.25?kW/m2?K to 1.265?kW/m2?K at temperature and salinity of 290?K and 0.00?ppm and also from 2.35?kW/m2?K to 1.365?kW/m2?K at temperature and salinity of 310?K and 60000?ppm, whereas the second aspect is the fouling factor variations due to the seawater salinity. The analysis showed that the two aspects have a significant impact on the computation of the condenser overall heat transfer coefficient, whereas the increase of seawater salinity leads to a reduction in the condenser overall heat transfer coefficient. 1. Introduction Thermal power plants are built for prescribed specific design conditions based on the targeted power demand, metallurgical limits of structural elements, statistical values of environmental conditions, and so forth. At design stage, a cooling medium temperature is chosen for each site considering long term average climate conditions. However, the working conditions deviate from the nominal operating conditions in practice. For this reason, efficiency in electricity production is affected by the deviation of the instantaneous operating temperature and salinity of seawater cooling water of nuclear power plant from the design temperature of the cooling medium extracted from environmental to transfer waste heat to the atmosphere via condenser. In the literature, there are few works published such as that of Kokaji [1] which indicated that all the nuclear power plants in Japan are located on the seaside and use a great amount of seawater for cooling condenser like fossil power generation plants. The resultant thermal discharge which may have an impact on the environmental of the surrounding areas is one of the most important issues to be given a serious consideration, next to prevention from affecting the local residents by released
Study on the Evaluation and Simulation of Steady State Behavior and Reactor Safety Concept for Integral Pressurized Water Reactor  [PDF]
Salah Ud-Din Khan,Minjun Peng,Muhammad Zubair
Information Technology Journal , 2011,
Abstract: In this study, a research has been carried out on the normal operational state concerning with the reactor under different load changes as well as simulation of pool type reactor concept by the use of thermal hydraulic system code Relap5/Mod3.4 for Integral Pressurized Water Reactors (IPWR s). These reactors belongs to the class of small and medium sized nuclear reactors (SMR s) and are designed in such away that most of the primary loop components are housed in a single Reactor Pressure Vessel (RPV). In this study, the reactor under study is the Inherent safe uranium zirconium hydride nuclear power reactor INSURE-100. Since, the conceptual design study phase of this reactor has been completed so the current research focuses on the normal operation which is inter related with different load characteristics as well as for the safety concept the considered reactor is simulated as a pool type. The effect of load changes on the different components of the reactor have been studied and are figure out in the graphical as well as in a parametric way. The current study also focuses on the simulation of pool type reactor concept, for this case the pressurized containment, which is partially filled with water and acts as a suppression pool. Finally the load percentage results and the steady state results obtained from the pool type reactor concept have been extracted from the graphical approach and are depicted in a parametric way and are sorted out in a table. The purpose of conducting the research is to evaluate the steady state behavior as well as safety analysis of the reactor.
Comparison of Small Modular Reactor and Large Nuclear Reactor Fuel Cost  [PDF]
Christopher P. Pannier, Radek Skoda
Energy and Power Engineering (EPE) , 2014, DOI: 10.4236/epe.2014.65009

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter construction times. SMRs also promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are public parameters for many near-term SMR projects. The fuel cost of electricity generation for selected SMRs and large reactors is simulated, including calculation of optimal tails assay in the uranium enrichment process. The results are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 70% higher fuel costs than large light water reactors using 2014 nuclear fuels market data. Fuel cost sensitivities to reactor design parameters are presented.

Passive Cooldown Performance of Integral Pressurized Water Reactor  [PDF]
Shoubao Dai, Chunnan Jin, Jingfu Wang, Yuxiang Chen
Energy and Power Engineering (EPE) , 2013, DOI: 10.4236/epe.2013.54B097

The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.

The SPES3 Experimental Facility Design for the IRIS Reactor Simulation  [PDF]
Mario Carelli,Lawrence Conway,Milorad Dzodzo,Andrea Maioli,Luca Oriani,Gary Storrick,Bojan Petrovic,Andrea Achilli,Gustavo Cattadori,Cinzia Congiu,Roberta Ferri,Marco Ricotti,Davide Papini,Fosco Bianchi,Paride Meloni,Stefano Monti,Fabio Berra,Davor Grgic,Graydon Yoder,Alessandro Alemberti
Science and Technology of Nuclear Installations , 2009, DOI: 10.1155/2009/579430
Abstract: IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.
Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor  [PDF]
Junli Gou,Suizheng Qiu,Guanghui Su,Douna Jia
Science and Technology of Nuclear Installations , 2009, DOI: 10.1155/2009/473795
Abstract: A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.
Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant  [PDF]
Analia Bonelli,Oscar Mazzantini,Martin Sonnenkalb,Marcelo Caputo,Juan Matias García,Pablo Zanocco,Marcelo Gimenez
Science and Technology of Nuclear Installations , 2012, DOI: 10.1155/2012/620298
Abstract: A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs. 1. Introduction The Central Nuclear Atucha 2 (CNA-2) is a nuclear power plant (NPP) with a two-loop, 745 MWe, Pressurized Heavy Water Reactor (PHWR), designed by Siemens-KWU and being under final construction in Lima, Argentina. The NPP is cooled and moderated by heavy water like a similar unit of smaller power (CNA-I) in operation at the same site since 1974. The reactor pressure vessel is very large and has a diameter of ~7.4?m. In difference to other PWRs the upper and lower plenum is to a large content occupied by filler pieces made of steel to reduce the necessary heavy water inventory (Figure 1). The reactor core consists of 451 vertical natural Uranium fuel assemblies located in the same number of coolant channels, connected each to the lower and upper reactor plenum. Each assembly consists of 37 fuel rods. The thermohydraulic design of the core divides the channels into five zones. For the external zones, specially designed flow limiters (drossels) are installed, so that the coolant flow in each channel zone is proportional to the average generated power in it, achieving almost the same channel outlet temperature for all the zones (Figure 2). The coolant channels are within the large moderator (MOD) tank. For reactivity reasons the moderator in it is maintained at a lower temperature than the reactor coolant. This is
Observation of the Isotopic Evolution of Pressurized Water Reactor Fuel Using an Antineutrino Detector  [PDF]
N. S. Bowden,A. Bernstein,S. Dazeley,R. Svoboda,A. Misner,T. Palmer
Physics , 2008, DOI: 10.1063/1.3080251
Abstract: By operating an antineutrino detector of simple design during several fuel cycles, we have observed long term changes in antineutrino flux that result from the isotopic evolution of a commercial Pressurized Water Reactor (PWR). Measurements made with simple antineutrino detectors of this kind offer an alternative means for verifying fissile inventories at reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes.
Application of CFD Codes in Nuclear Reactor Safety Analysis  [PDF]
T. H?hne,E. Krepper,U. Rohde
Science and Technology of Nuclear Installations , 2010, DOI: 10.1155/2010/198758
Abstract: Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/G?rlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance. 1. Introduction The last decade has seen an increasing use of three-dimensional CFD codes to predict steady state and transient flows in nuclear reactors because a number of important phenomena such as pressurized thermal shocks, coolant mixing, and thermal striping cannot be predicted by traditional one-dimensional system codes with the required accuracy and spatial resolution. CFD codes contain models for simulating turbulence, heat transfer, multiphase flows, and chemical reactions. Such models must be validated before they can be used with sufficient confidence in NRS applications. The necessary validation is performed by comparing model results against measured data. However, in order to obtain a reliable model assessment, CFD simulations for validation purposes must satisfy strict quality criteria given in the Best Practice Guidelines (BPGs). Our partner for CFD code qualification is ANSYS CFX [1], which is one of the leading CFD codes worldwide. Based on this partnership the models developed are implemented into the code and thus contribute to the code qualification. In principle the presented simulation could be performed by any other actual CFD-code. The following topical issues, where CFD calculations have been performed, will be briefly discussed in the paper: (1)coolant mixing, (2)corizontal stratified flow phenomena in the Hot Leg of PWR,(3)Debris transport phenomena in multidimensional water
Core Optimization Simulation for a Pressurized Water Reactor  [PDF]
A. Hussain,C. Xinrong
Information Technology Journal , 2009,
Abstract: In this study, a research has been carried out for the design of an optimal core configuration for a TRISO fueled compact sized PWR core. This is a light water cooled and moderated reactor that employs TRISO fuel particles in zirconium-sheathed fuel rods. The combination of PWR technology and TRISO fuel has been preferred for research to get the benefits of TRISO fuel in terms of enhanced integrity against the release of fission fragments and high negative temperature coefficient of reactivity in well proven PWR technology. This PWR design possesses additional safety features associated with the default design of TRISO fuel particle, which makes its use suitable even in a densely populated area. The designed core can be utilized for heating and desalination purposes or at any remotely located research facility. The current research study has been focused on the core configuration, instead of selecting one of the standard fuel lattices which are mostly being used in nuclear power plants; an inventive fuel lattice has been suggested for the optimal design. The TRISO fuel particle size and fuel pitch have also been optimized to achieve a compact size core. Neutronic transport theory lattice code WIMS-D/4 was used for the calculation of group constants (D, Σa and vΣf) and infinite multiplication factor (k∞). This calculated data were used in diffusion theory code CITATION for the purpose of achieving effective multiplication factor (keff) and estimated life of the core. The detailed and thorough analyses revealed that core configuration plays a dominant role in determination of compactness and excess reactivity of the core. The amount of excess reactivity has been increased and core size has been condensed by designing an optimal core.
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