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Thermal Hydraulic Aspects Of The Use Of Low Enrichment Uranium Fuel In The Mit Research Reactor
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Book Synopsis Thermal-hydraulic Aspects of the Use of Low Enrichment Uranium Fuel in the MIT Research Reactor by : Joseph B. Gehret
Download or read book Thermal-hydraulic Aspects of the Use of Low Enrichment Uranium Fuel in the MIT Research Reactor written by Joseph B. Gehret and published by . This book was released on 1984 with total page 304 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal Hydraulics Analysis of the MIT Research Reactor in Support of a Low Enrichment Uranium (LEU) Core Conversion by : Yu-Chih Ko (Ph. D.)
Download or read book Thermal Hydraulics Analysis of the MIT Research Reactor in Support of a Low Enrichment Uranium (LEU) Core Conversion written by Yu-Chih Ko (Ph. D.) and published by . This book was released on 2008 with total page 290 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The objectives of this study are to benchmark the in-house computer code for the MITR, and to perform the thermal hydraulic analyses in support of the LEU design studies. The in-house multi-channel thermal-hydraulics code, MULCH-II, was developed specifically for the MITR. This code was validated against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. Various fuel configurations are evaluated as part of the LEU core design optimization study. The criteria adopted for the LEU thermal hydraulics analysis for this study are the limiting safety system settings (LSSS), to prevent onset of nucleate boiling during steady-state operation, and to avoid a clad temperature excursion during the loss of flow transient. The benchmark analysis results showed that the MULCH-II code is in good agreement with other computer codes and experimental data, and hence it is used as the main tool for this study. In ranking the LEU core design options, the primary parameter is a low power peaking factor in order to increase the LSSS power and to decrease the maximum clad temperature during the transient. The LEU fuel designs with 15 to 18 plates per element, fuel thickness of 20 mils, and a hot channel factor less than 1.76 are shown to comply with these thermal-hydraulic criteria. The steady-state power can potentially be higher than 6 MW, as requested in the power upgrade submission to the Nuclear Regulatory Commission.
Book Synopsis Thermal Hydraulic Limits Analysis for the Massachusetts Institute of Technology Research Reactor Low Enrichment Uranium Core Conversion Using Statistical Propagation of Parametric Uncertainties by : Keng-Yen Chiang
Download or read book Thermal Hydraulic Limits Analysis for the Massachusetts Institute of Technology Research Reactor Low Enrichment Uranium Core Conversion Using Statistical Propagation of Parametric Uncertainties written by Keng-Yen Chiang and published by . This book was released on 2012 with total page 171 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor power upgraded from 6 MW to 7 MW is proposed in order to maintain the same reactor performance of the HEU core. Previous approaches in analyzing the impact of engineering uncertainties on thermal hydraulic limits via the use of engineering hot channel factors (EHCFs) were unable to explicitly quantify the uncertainty and confidence level in reactor parameters. The objective of this study is to develop a methodology for MITR thermal hydraulic limits analysis by statistically combining engineering uncertainties in order to eliminate unnecessary conservatism inherent in traditional analyses. This methodology was employed to analyze the Limiting Safety System Settings (LSSS) for the MITR LEU core, based on the criterion of onset of nucleate boiling (ONB). Key parameters, such as coolant channel tolerances and heat transfer coefficients, were considered as normal distributions using Oracle Crystal Ball for the LSSS evaluation. The LSSS power is determined with 99.7% confidence level. The LSSS power calculated using this new methodology is 9.1 MW, based on core outlet coolant temperature of 60 'C, and primary coolant flow rate of 1800 gpm, compared to 8.3 MW obtained from the analytical method using the EHCFs with same operating conditions. The same methodology was also used to calculate the safety limit (SL) to ensure that adequate safety margin exists between LSSS and SL. The criterion used to calculate SL is the onset of flow instability. The calculated SL is 10.6 MW, which is 1.5 MW higher than LSSS, permitting sufficient margin between LSSS and SL.
Book Synopsis Impact Assessment for the MIT Research Reactor Low Enrichment Uranium Fuel Fabrication Tolerances by : Dakota J. Allen
Download or read book Impact Assessment for the MIT Research Reactor Low Enrichment Uranium Fuel Fabrication Tolerances written by Dakota J. Allen and published by . This book was released on 2020 with total page 109 pages. Available in PDF, EPUB and Kindle. Book excerpt: In the framework of non-proliferation policy, the Massachusetts Institute of Technology Reactor (MITR) is planning to convert from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel. A new type of high-density LEU fuel based on a monolithic U-10Mo alloy is being qualified to allow the conversion of all remaining U.S. high performance research reactors including the MITR. The purpose of this study is to understand the impact of proposed MITR LEU "FYT" fuel element fabrication tolerances on the operation and safety limits of the MITR. Therefore, the effects of fabrication specification parameters on all levels of the core, ranging from full-core alterations to individual spots on the fuel plates were analyzed. Evaluations at the design tolerances, and beyond, were conducted through neutronics and thermal hydraulics calculations. The first step was analyzing the separate effects that parameters, including enrichment, fuel mass loading, fuel plate thickness, and impurities, have on the reactor physics of the core. These analyses were used to develop curve fits to predict the effect of these parameters on the excess reactivity of fresh fuel inserted into the LEU core. These models could then be used to estimate the effect on fuel cycle length to ensure the tolerances would not cause significant changes to the operating cycle of MITR. These analyses estimated the margin to criticality present in the core and ensured that the reactivity shutdown margin (SDM) was not violated. Other parameters such as coolant channel gap and local fuel homogeneity cause primarily local impacts including the power distribution within the fuel element, and related impacts to thermal hydraulic margins. This modeling was necessary to ensure that these parameters would not cause the margin to MITR's thermal hydraulic safety limit, the onset of nucleate boiling (ONB), to be violated. The final step was a covariance analysis of the combined effects at a full-core and element level. This combined effect analysis assured that the core would maintain proper safety and operational margins with a realistic distribution of off-nominal parameters. Given the comprehensive analysis performed, the current design fabrication tolerances were determined to provide acceptable fuel cycle length and safety margins consistent with the MITR LEU preliminary safety analysis report, and a basis for updating these tolerances during planned manufacturing-scale plate fabrication demonstrations has been established.
Book Synopsis Evaluation of the Thermal-hydraulic Operating Limits of the HEU-LEU Transition Cores for the MIT Research Reactor by : Yunzhi Diana Wang
Download or read book Evaluation of the Thermal-hydraulic Operating Limits of the HEU-LEU Transition Cores for the MIT Research Reactor written by Yunzhi Diana Wang and published by . This book was released on 2009 with total page 115 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 plates per element, compared to the existing HEU design of 15 plates per element. A transitional conversion strategy, which consists of replacing three HEU elements with fresh LEU fuel elements in each fuel cycle, is proposed. The objective of this thesis is to analyze the thermo-hydraulic safety margins and to determine the operating power limits of the MITR for each mixed core configuration. The analysis was performed using PLTEMP/ANL ver 3.5, a program that was developed for thermo-hydraulic calculations of research reactors. Two correlations were used to model the friction pressure drop and enhanced heat transfer of the finned fuel plates: the Carnavos correlation for friction factor and heat transfer, and the Wong Correlation for friction factor with a constant heat transfer enhancement factor of 1.9. With these correlations, the minimum onset of nucleate boiling (ONB) margins of the hottest fuel plates were evaluated in nine different core configurations, the HEU core, the LEU core and seven mixed cores that consist of both HEU and LEU elements. The maximum radial power peaking factors were assumed at 2.0 for HEU and 1.76 for LEU in all the analyzed core configurations. The calculated results indicate that the HEU fuel elements yielded lower ONB margins than LEU fuel elements in all mixed core configurations. In addition to full coolant channels, side channels next to the support plates that form side coolant channels were analyzed and found to be more limiting due to higher flow resistance. The maximum operating powers during the HEU to LEU transition were determined by maintaining the minimum ONB margin corresponding to the homogeneous HEU core at 6 MW. The recommended steady-state power is 5.8 MW for all transitional cores if the maximum radial peaking is adjacent to a full coolant channel and 4.9 MW if the maximum radial peaking is adjacent to a side coolant channel.
Book Synopsis Development of a Low Enrichment Uranium Core for the MIT Reactor by : Thomas Henderson Newton
Download or read book Development of a Low Enrichment Uranium Core for the MIT Reactor written by Thomas Henderson Newton and published by . This book was released on 2006 with total page 312 pages. Available in PDF, EPUB and Kindle. Book excerpt: (cont.) Thermal-hydraulic calculations using the multi-channel thermal-hydraulics analysis code MULCH-II indicated that the peak power channel will remain below the Onset of Nucleate Boiling under all normal operating conditions as well as loss of flow conditions. In addition, using MCNP and the thermal-hydraulics/point kinetics code PARET it was shown that all reactivity coefficients were negative and that the LEU core could withstand a step reactivity insertion of $3.69 without reaching cladding softening temperature, thus increasing the allowable reactivity for an incore experiment. Finally, it is possible to use the proposed design to increase the neutron flux by increasing core power, but with a correspondingly reduced refueling cycle length.
Book Synopsis LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor by : Yinjie Zhao
Download or read book LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor written by Yinjie Zhao and published by . This book was released on 2022 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Research Reactor (MITR) is in the process of converting from the current 93%-enriched U-235 highly-enriched uranium (HEU) fuel to the low enriched uranium (LEU, 20%-enriched U-235) fuel, as part of the global non-proliferation initiatives. A high-density, monolithic uraniummolybdenum (U-10Mo) fuel matrix is chosen. The fuel element design is changed from 15-plate finned HEU fuel to 19-plate unfinned LEU fuel with the same geometry. The reactor power increases from 6.0 MW to 7.0 MW thermal, and primary coolant flow rate increases from 2000 gpm to 2400 gpm. Detailed analyses were completed for initial LEU core with 22 fuel elements, and demonstrated both neutronic and thermal hydraulic safety requirements are met throughout equilibrium cycles. An alternative conversion strategy is proposed which involves a gradual transition from an all-HEU core to an all-LEU core by replacing 3 HEU fuel elements with fresh LEU fuel elements during each fuel cycle. The objectives of this study are to demonstrate that the primary coolant system can be safely modified for 2400 gpm operation, and to perform steady-state and loss-of-flow (LOF) transient thermal-hydraulic analyses for the MITR HEU-LEU transitional mixed cores to evaluate this alternative conversion strategy. The primary technical challenge for the 20% increase in primary flow rate with existing piping system is flow-induced vibration. Several experiments were performed to measure and quantify vibration acceleration and velocity on three main hydraulic components to determine if higher flowrates cause excessive vibration. The test results show that the maximum vibration velocity is 9.70 mm/s, the maximum vibration acceleration is 0.98 G at the current flow rate 2000 gpm and no significant spectral change in the vibration profile at 2550 gpm. Therefore, it can be concluded that the existing piping system can safely support 2400 gpm primary flow operation. Thermal hydraulics analysis was performed using RELAP5 MOD3.3 code and STAT7 code. The MITR transitional mixed core input models were constructed to simulate the reactor primary system. Two scenarios, steady-state and loss-of-flow transient were simulated at power level of 6 MW. RELAP5 results show that during steady state, there is significant safety margin ( 10 °C) to onset of nucleate boiling for both HEU and LEU fuel. The maximum core temperature occurs at HEU fuel in Mix-core 3, the maximum wall temperature reached was 89 °C. During the LOF transient case, the result shows that The HEU fuel element is more limiting than the LEU in transitional cores. Nucleate boiling is predicted to occur only in the HEU hot channel during the first 50 seconds after the pump coastdown. The peak cladding temperatures are much lower than the fuel temperature safety limit of UAl[subscript x] fuel plates, which is 450 °C. From the STAT7 calculation results, the operational limiting power at which onset of nucleate boiling (ONB) occurs in all cases show significant margins from the Limiting System Safety Setting (LSSS) over-power level. The lowest margin for LEU element during the mixed core transition is at Mix-7, 11.43 MW with a 4.03 MW power margin. For the HEU element, the lowest margin during the transition is at Mix-2, 8.51 MW with a 1.11 MW power margin. The location at which ONB is always expected to occur is F-Plate Stripe 1 and 4 for the LEU fuel element; side plate for the HEU fuel element with the HEU element is always more limiting.
Book Synopsis Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors by : Ferry Roelofs
Download or read book Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors written by Ferry Roelofs and published by Woodhead Publishing. This book was released on 2018-11-30 with total page 464 pages. Available in PDF, EPUB and Kindle. Book excerpt: Thermal Hydraulics Aspects of Liquid Metal cooled Nuclear Reactors is a comprehensive collection of liquid metal thermal hydraulics research and development for nuclear liquid metal reactor applications. A deliverable of the SESAME H2020 project, this book is written by top European experts who discuss topics of note that are supplemented by an international contribution from U.S. partners within the framework of the NEAMS program under the U.S. DOE. This book is a convenient source for students, professionals and academics interested in liquid metal thermal hydraulics in nuclear applications. In addition, it will also help newcomers become familiar with current techniques and knowledge. Presents the latest information on one of the deliverables of the SESAME H2020 project Provides an overview on the design and history of liquid metal cooled fast reactors worldwide Describes the challenges in thermal hydraulics related to the design and safety analysis of liquid metal cooled fast reactors Includes the codes, methods, correlations, guidelines and limitations for liquid metal fast reactor thermal hydraulic simulations clearly Discusses state-of-the-art, multi-scale techniques for liquid metal fast reactor thermal hydraulics applications
Author :United States. Congress. House. Committee on Science and Technology. Subcommittee on Energy Development and Applications Publisher : ISBN 13 : Total Pages :1968 pages Book Rating :4.3/5 (121 download)
Book Synopsis Conversion of Research and Test Reactors to Low-enriched Uranium (LEU) Fuel by : United States. Congress. House. Committee on Science and Technology. Subcommittee on Energy Development and Applications
Download or read book Conversion of Research and Test Reactors to Low-enriched Uranium (LEU) Fuel written by United States. Congress. House. Committee on Science and Technology. Subcommittee on Energy Development and Applications and published by . This book was released on 1985 with total page 1968 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Абхазия written by and published by . This book was released on 2011 with total page 574 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Masters Theses in the Pure and Applied Sciences by : Wade H. Shafer
Download or read book Masters Theses in the Pure and Applied Sciences written by Wade H. Shafer and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 407 pages. Available in PDF, EPUB and Kindle. Book excerpt: Masters Theses in the Pure and Applied Sciences was first conceived, published, and disseminated by the Center for Information and Numerical Data Analysis and Synthesis (CINDAS) * at Purdue University in 1 957, starting its coverage of theses with the academic year 1955. Beginning with Volume 13, the printing and dissemination phases of the activity were transferred to University Microfilms/Xerox of Ann Arbor, Michigan, with the thought that such an arrangement would be more beneficial to the academic and general scientific and technical community. After five years of this joint undertaking we had concluded that it was in the interest of all con cerned if the printing and distribution of the volumes were handled by an interna tional publishing house to assure improved service and broader dissemination. Hence, starting with Volume 18, Masters Theses in the Pure and Applied Sciences has been disseminated on a worldwide basis by Plenum Publishing Cor poration of New York, and in the same year the coverage was broadened to include Canadian universities. All back issues can also be ordered from Plenum. We have reported in Volume 29 (thesis year 1984) a total of 12,637 theses titles from 23 Canadian and 202 United States universities. We are sure that this broader base for these titles reported will greatly enhance the value of this important annual reference work. While Volume 29 reports theses submitted in 1984, on occasion, certain univer sities do report theses submitted in previous years but not reported at the time.
Book Synopsis Thermal-hydraulic Calculations for the GRR-1 Research Reactor Core Conversion to Low Enriched Uranium Fuel by : C. Housiadas
Download or read book Thermal-hydraulic Calculations for the GRR-1 Research Reactor Core Conversion to Low Enriched Uranium Fuel written by C. Housiadas and published by . This book was released on 1999 with total page 36 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Friction Pressure Drop Measurements and Flow Distribution Analysis for LEU Conversion Study of MIT Research Reactor by : Susanna Yuen-Ting Wong
Download or read book Friction Pressure Drop Measurements and Flow Distribution Analysis for LEU Conversion Study of MIT Research Reactor written by Susanna Yuen-Ting Wong and published by . This book was released on 2008 with total page 151 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal hydraulic analyses, propose different fuel element designs for optimization of thermal hydraulic performance of the LEU core. Since proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the friction pressure drop is required to be verified experimentally. The objectives of this study are to measure the friction coefficient in both laminar and turbulent flow regions, and to develop empirical correlations for the finned rectangular coolant channels for the safety analysis of the MITR. A friction pressure drop experiment is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is measured for both flat and finned coolant channels of various channel heights. Experiment data show that the Darcy friction factors for laminar flow in finned rectangular channels are in good agreement with the existing correlation if a pseudo-smooth equivalent hydraulic diameter is considered; whereas a new friction factor correlation is proposed for the friction factors for turbulent flow. Additionally, a model is developed to calculate the primary flow distribution in the reactor core for transitional core configuration with various combinations of HEU and LEU fuel elements.
Book Synopsis Thermal-hydraulic Optimization for High Production of Low-enriched Uranium Based Molybdenum-99 by : Jeffrey William Scott
Download or read book Thermal-hydraulic Optimization for High Production of Low-enriched Uranium Based Molybdenum-99 written by Jeffrey William Scott and published by . This book was released on 2009 with total page 136 pages. Available in PDF, EPUB and Kindle. Book excerpt: Globally, more than 20 million samples of technetium-99 are used annually to diagnose many different forms of cancer. Despite this, there are only four main nuclear reactors worldwide that produce molybdenum-99, the parent isotope of technetium-99m. In an attempt to address the growing demand, as well as to motivate the current reactors to stop using high-enriched uranium, the University of Missouri Research Reactor (MURR) has begun to produce molybdenum-99 using low-enriched uranium. The fission of uranium-235, for which one of the products is molybdenum-99, generates approximately 2 kW of heat per 4-gram target. Continually removing this heat from the fission reaction is the limiting factor in the high-volume production process. The objective of this thesis is to find a reactor wedge setup with maximum rate of heat removal, thereby maximizing the amount of molybdenum-99 that can be created in the reactor. To maximize heat transfer, a quasi 1-D analytic model, calibrated numerically, is created to hydraulically and thermally model the coolant flow through the current reactor setup. Using flow network modeling (FNM), this analytic model is expanded to analyze other potential geometries that could maximize heat transfer. The findings show that for a single channel under the existing reactor configuration, a maximum of 19.18 kW can be dispersed. By opening the drain and slightly shrinking the channel diameter, this can be improved to 22.77 kW. When the system is expanded to 10 parallel channels and the drain is fully opened, over 250 kW can theoretically be achieved, though the neutronics of the MURR reactor will likely limit this.
Book Synopsis Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report by :
Download or read book Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n, a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.
Book Synopsis Nuclear Reactor Thermal Hydraulics and Other Applications by : Donna Guillen
Download or read book Nuclear Reactor Thermal Hydraulics and Other Applications written by Donna Guillen and published by BoD – Books on Demand. This book was released on 2013-02-13 with total page 204 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book includes contributions from researchers around the world on numerical developments and applications to predict fluid flow and heat transfer, with an emphasis on thermal hydraulics computational fluid dynamics. Our ability to simulate larger problems with greater fidelity has vastly expanded over the past decade. The collection of material presented in this book augments the ever-increasing body of knowledge concerning the important topic of thermal hydraulics. Featured topics include coolant channel analysis, thermal hydraulic transport and mixing, as well as hydrodynamics and heat transfer processes. The contents of this book will interest researchers, scientists, engineers and graduate students.
Book Synopsis Thermal Hydraulic Mixing Transients in the MIT Research Reactor Core Tank by : Lin-Wen Hu
Download or read book Thermal Hydraulic Mixing Transients in the MIT Research Reactor Core Tank written by Lin-Wen Hu and published by . This book was released on 1996 with total page 360 pages. Available in PDF, EPUB and Kindle. Book excerpt:
