subject:(Reactor physics)

University of Ontario Institute of Technology

1. Patel, Amin. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.

Degree: 2010, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/87

► Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several… (more)

Subjects/Keywords: Applied reactor physics; Transport theory; Diffusion theory; CANDU; Nuclear reactor

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APA (6^th Edition):

Patel, A. (2010). Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/87

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Patel, Amin. “Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.” 2010. Thesis, University of Ontario Institute of Technology. Accessed February 16, 2019. http://hdl.handle.net/10155/87.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Patel, Amin. “Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.” 2010. Web. 16 Feb 2019.

Vancouver:

Patel A. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. [Internet] [Thesis]. University of Ontario Institute of Technology; 2010. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/10155/87.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Patel A. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. [Thesis]. University of Ontario Institute of Technology; 2010. Available from: http://hdl.handle.net/10155/87

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

The Ohio State University

2. Kennedy, Ryanne Ariel. Quantifying Uncertainty in Reactor Flux/Power Distributions.

Degree: PhD, Nuclear Engineering, 2011, The Ohio State University

URL: http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901

► The design and development of a conceptual system for power measurements in a reactor core using in-core sensors has been an ongoing focus of research… (more)

▼ The design and development of a conceptual system for power measurements in a reactor core using in-core sensors has been an ongoing focus of research performed at The Ohio State University. Previous work focused on the development of software that constructs three-dimensional core power distribution using signals from an array of Constant Temperature Power Sensors distributed in the reactor core. The monitoring system processes the sensor signals from the reactor core in such a manner that probabilistic information, in the form of probability distribution functions (pdfs), of reactor power density is obtained at the sensor locations. The desired pdfs of power density at sensor locations were obtained through the implementation of the estimation algorithm Dynamic System Doctor (DSD). Additional work in this area proposed methods to interpolate the pdfs at the measurement points for uniform core compositions. The interpolation methods presented in this thesis extend estimation of the uncertainty in power/flux distribution to heterogeneous cores by combining the knowledge from in-core sensors and reactor core neutron transport calculations. Two significant benefits of such an estimation process are the following:1. It allows quantification of the uncertainty on peaking factors as well as global power/flux distributions and hence evaluation of potential for power upgrades2. It provides data for a best-estimate quantification of design margins.The interpolation method is an adaptation of an interpolation procedure presented by L. Read and utilizes the flux/power distributions obtained from a realistic reactor core model as interpolating functions. The pdfs at the interpolated power/flux values are calculated using a weighted linear interpolation of the sensor pdfs. Four weighting schemes are considered for the power/flux interpolation algorithm. The proposed interpolation methods and each weighting scheme are tested using flux data from models of the Ohio State University Research Reactor (OSURR). The objective of the testing of each scheme is to determine its accuracy/conservatism, and to gauge whether the interpolation is physically representative of the process under consideration. Different methods are compared as to their ability to accurately estimate the pdfs at different positions in the reactor core. In view of the uncertainty in the current material composition of OSURR due to burnup of fuel during the past two decades, the Monte Carlo code MCNP5 is used to generate the “experimental data”. This approach also allows control of the uncertainty on the tally results by changing the number of neutron simulations performed. For the test problems, the OSURR core model flux distributions used as distribution interpolating functions (DIF) were generated by PENTRAN (Parallel Environment Neutral particle Transport) 3-D discrete ordinates code (version 9.36k). Three example problems were used to test the linear interpolation schemes. In these three examples the interpolating method is applied to the flux distribution in… Advisors/Committee Members: Aldemir, Tunc (Advisor).

Subjects/Keywords: Nuclear Engineering; nuclear reactor; flux; uncertainty quantification; reactor physics

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APA (6^th Edition):

Kennedy, R. A. (2011). Quantifying Uncertainty in Reactor Flux/Power Distributions. (Doctoral Dissertation). The Ohio State University. Retrieved from http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901

Chicago Manual of Style (16^th Edition):

Kennedy, Ryanne Ariel. “Quantifying Uncertainty in Reactor Flux/Power Distributions.” 2011. Doctoral Dissertation, The Ohio State University. Accessed February 16, 2019. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901.

MLA Handbook (7^th Edition):

Kennedy, Ryanne Ariel. “Quantifying Uncertainty in Reactor Flux/Power Distributions.” 2011. Web. 16 Feb 2019.

Vancouver:

Kennedy RA. Quantifying Uncertainty in Reactor Flux/Power Distributions. [Internet] [Doctoral dissertation]. The Ohio State University; 2011. [cited 2019 Feb 16]. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901.

Council of Science Editors:

Kennedy RA. Quantifying Uncertainty in Reactor Flux/Power Distributions. [Doctoral Dissertation]. The Ohio State University; 2011. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901

Georgia Tech

3. Abou Jaoude, Abdalla. Design of a mixed-spectrum long-lived reactor with improved proliferation resistance.

Degree: PhD, Mechanical Engineering, 2017, Georgia Tech

URL: http://hdl.handle.net/1853/59275

► Long-lived fast reactors have been suggested as an effective way of spreading nuclear energy to new countries. These small reactors can be produced at centralized… (more)

▼ Long-lived fast reactors have been suggested as an effective way of spreading nuclear energy to new countries. These small reactors can be produced at centralized locations, shipped to area of need, then returned to the main hub at the end of their lifetime for decommissioning. Such ‘hub-spoke’ arrangements disincentivizes states front building sensitive front and back-end technology; however, critics argue they still pose a proliferation risk due to the large quantity of weapon-grade plutonium they produce during their operating lifetime. The dissertation attempts to address this issue by proposing a mixed-spectrum core configuration. A fast neutron zone can increase fissile material production, while a thermalized zone reduces plutonium quality. Moderating material (ZrH1.6) is inserted within peripheral assemblies, while the center of the core maintains a fast configuration. Assemblies are then shuffled to ensure all are exposed to the thermalized spectrum. This allows the new design to simultaneously improve proliferation resistance and reduce fast fluence damage, a limiting criteria for long-lived core designs. The objectives are achieved with minimal impact on overall performance. Core lifetime can be maintained at 25 years, without the need for any additional fuel. Inherent passive safety criteria can be met, and power peaking phenomena at the fast/thermal interface was deemed to be manageable. Different design variants that can alleviate power peaking or leverage the ability of thorium-cycle breeding in the epithermal regime, were also investigated. Mixed-spectrum cores pushes the boundaries of what deterministic codes are capable of modeling accuracy. The REBUS suite of codes is modified to provide a more accurate tool to explore the design space. MCNP6 is then used for detailed analysis and safety evaluation of optimal core configurations. The thesis demonstrates the viability of using a mixed-spectrum reactor design to improve proliferation resistance of long-lived cores. The main identified tradeoff was an increase in overall resource consumption, a slightly larger core size, and the reliance on shuffling midway through the core lifetime. Advisors/Committee Members: Erickson, Anna (advisor), Petrovic, Bojan (committee member), Hertel, Nolan (committee member), Stulberg, Adam (committee member), Stauff, Nicolas (committee member).

Subjects/Keywords: Reactor physics; Proliferation resistance; Mixed-spectrum reactors; Reactor design; Safety evaluation

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APA (6^th Edition):

Abou Jaoude, A. (2017). Design of a mixed-spectrum long-lived reactor with improved proliferation resistance. (Doctoral Dissertation). Georgia Tech. Retrieved from http://hdl.handle.net/1853/59275

Chicago Manual of Style (16^th Edition):

Abou Jaoude, Abdalla. “Design of a mixed-spectrum long-lived reactor with improved proliferation resistance.” 2017. Doctoral Dissertation, Georgia Tech. Accessed February 16, 2019. http://hdl.handle.net/1853/59275.

MLA Handbook (7^th Edition):

Abou Jaoude, Abdalla. “Design of a mixed-spectrum long-lived reactor with improved proliferation resistance.” 2017. Web. 16 Feb 2019.

Vancouver:

Abou Jaoude A. Design of a mixed-spectrum long-lived reactor with improved proliferation resistance. [Internet] [Doctoral dissertation]. Georgia Tech; 2017. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1853/59275.

Council of Science Editors:

Abou Jaoude A. Design of a mixed-spectrum long-lived reactor with improved proliferation resistance. [Doctoral Dissertation]. Georgia Tech; 2017. Available from: http://hdl.handle.net/1853/59275

McMaster University

4. McDonald, Michael H. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.

Degree: MASc, 2011, McMaster University

URL: http://hdl.handle.net/11375/11223

►

The supercritical water cooled reactor (SCWR) is a Generation IV reactor concept that features light water coolant in a supercritical state. Canada is developing… (more)

▼

The supercritical water cooled reactor (SCWR) is a Generation IV reactor concept that features light water coolant in a supercritical state. Canada is developing a pressure tube variant of the supercritical water reactor as an evolution of the CANDU reactor. The main advantages of the pressure tube SCWR are an improved thermal efficiency over current reactors, enhanced safety through passive safety features, and plant simplifications. The objective of this thesis was to investigate current fuel and core designs for the Canadian SCWR concept. Simulations of 2-D lattice cells for fuel assemblies containing 43 and 54 fuel elements were performed using the neutron transport code WIMS-AECL. Safety parameters and fuel burnup performance were investigated here. Three dimensional full core simulations were performed using the diffusion code RFSP. These studies examined batch fueling, cycle length, radial and axial power profiles, linear element ratings, and reduction of axial power peaking through graded enrichment along the fuel channel. Finally, a study of reactivity transients was performed using the FUELPIN heat transfer/point kinetics code. The main results of the studies show that the coolant density change that occurs as water passes through the pseudocritical point strongly affects fuel performance. It is concluded that the 54 element assembly design is acceptable in terms of coolant void reactivity performance with lattice pitch smaller than 26 cm. To meet the burnup target, a fuel enrichment of about 5% is required. From the RFSP studies, this level of fuel enrichment will provide an operating period of 370 days between refueling. Relatively high axial power peaking is observed at the beginning of cycle conditions. A main finding is that the proposed reactor power level of 2540 MWth produces unacceptably high linear element ratings. This is confirmed using the FUELPIN code. A reduction in linear element rating is suggested for consideration.

Master of Applied Science (MASc)

Advisors/Committee Members: Novog, D. R., Engineering Physics and Nuclear Engineering.

Subjects/Keywords: Reactor physics; supercritical water cooled reactor; fuel; Nuclear Engineering; Nuclear Engineering

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APA (6^th Edition):

McDonald, M. H. (2011). Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/11223

Chicago Manual of Style (16^th Edition):

McDonald, Michael H. “Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.” 2011. Masters Thesis, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/11223.

MLA Handbook (7^th Edition):

McDonald, Michael H. “Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.” 2011. Web. 16 Feb 2019.

Vancouver:

McDonald MH. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. [Internet] [Masters thesis]. McMaster University; 2011. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/11223.

Council of Science Editors:

McDonald MH. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. [Masters Thesis]. McMaster University; 2011. Available from: http://hdl.handle.net/11375/11223

University of Ontario Institute of Technology

5. Mohapatra, Subhramanyu. Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors.

Degree: 2016, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/740

► To avoid the computational effort associated with full-core neutron transport calculations, full-core neutronics calculations for Pressurized Heavy-Water Reactors (PHWRs) are usually performed in diffusion theory… (more)

Subjects/Keywords: Applied reactor physics; Superhomogenization; PHWR; SPH factors

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APA (6^th Edition):

Mohapatra, S. (2016). Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/740

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Mohapatra, Subhramanyu. “Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors.” 2016. Thesis, University of Ontario Institute of Technology. Accessed February 16, 2019. http://hdl.handle.net/10155/740.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Mohapatra, Subhramanyu. “Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors.” 2016. Web. 16 Feb 2019.

Vancouver:

Mohapatra S. Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors. [Internet] [Thesis]. University of Ontario Institute of Technology; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/10155/740.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Mohapatra S. Investigation of sub-cell homogenization for PHWR lattice cells using superhomogenization factors. [Thesis]. University of Ontario Institute of Technology; 2016. Available from: http://hdl.handle.net/10155/740

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Texas A&M University

6. Kitcher, Evans Damenortey. Xenon Induced Power Oscillations in a Generic Small Modular Reactor.

Degree: 2016, Texas A&M University

URL: http://hdl.handle.net/1969.1/157767

► As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution.… (more)

▼ As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a preliminary investigation into excess reactivity control options for the SMR design was conducted confirming the generally held notion that existing PWR control mechanisms can be used in iPWR SMRs with similar effectiveness. With the desire to operate the SMR under the boron free coolant condition, erbium oxide fuel integral burnable absorber rods were identified as a possible means to retain the dispersed absorber effect of soluble boron in the reactor coolant in replacement. Advisors/Committee Members: Chirayath, Sunil S (advisor), Charlton, William S (committee member), Poston, John W (committee member), Bangerth, Wolfgang (committee member).

Subjects/Keywords: Xenon Oscillations; Small Modular Reactor; Multi-physics

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APA (6^th Edition):

Kitcher, E. D. (2016). Xenon Induced Power Oscillations in a Generic Small Modular Reactor. (Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/157767

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Kitcher, Evans Damenortey. “Xenon Induced Power Oscillations in a Generic Small Modular Reactor.” 2016. Thesis, Texas A&M University. Accessed February 16, 2019. http://hdl.handle.net/1969.1/157767.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Kitcher, Evans Damenortey. “Xenon Induced Power Oscillations in a Generic Small Modular Reactor.” 2016. Web. 16 Feb 2019.

Vancouver:

Kitcher ED. Xenon Induced Power Oscillations in a Generic Small Modular Reactor. [Internet] [Thesis]. Texas A&M University; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1969.1/157767.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Kitcher ED. Xenon Induced Power Oscillations in a Generic Small Modular Reactor. [Thesis]. Texas A&M University; 2016. Available from: http://hdl.handle.net/1969.1/157767

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

University of Colorado

7. Vasiliou, AnGayle Konstance. Thermochemical Conversion of Biomass: A Molecular Viewpoint.

Degree: PhD, Chemistry & Biochemistry, 2011, University of Colorado

URL: http://scholar.colorado.edu/chem_gradetds/53

► This dissertation describes experiments performed to study the thermal decomposition of biomass from a molecular viewpoint. The structure of biomass consists of three major… (more)

▼ This dissertation describes experiments performed to study the thermal decomposition of biomass from a molecular viewpoint. The structure of biomass consists of three major parts: cellulose, hemicellulose and lignin. Thermochemical conversion of biomass, specifically pyrolysis and gasification, yields a complex mixture of light gases, condensable vapors and aromatic tars. The goal for the gasification of biomass is to maximize the production of syngas (CO and H2 ) and minimize the production of aromatic tars. This thesis provides thermochemical information particularly related to cellulose decomposition. The current technology for the conversion of biomass to biofuels is hindered by the lack of fundamental knowledge concerning detailed mechanisms and kinetic parameters that govern the process. In order to approach this problem, this work provides such information for furan, furfural, acetaldehyde and propionaldehyde, known intermediates in the pyrolysis of cellulose. The thermal decomposition of the aforementioned biomass molecules was formed in a microtubular reactor with pressures of 75-100 torr and up to temperatures of 1700 K corresponding to residence times of roughly 30-100 μs in the heated reactor. The biomass molecules were entrained in the carrier gases He or Ar and passed through the reactor. The thermal decomposition of the molecules occurs during transit through the heated reactor and products are cooled upon expansion into a vacuum chamber. The pyrolysis product beam was interrogated by three unique schemes: Photoionization Time of Flight Mass Spectroscopy (PIMS) using 10.5 eV light, Matrix Isolation Infrared (IR) Spectroscopy and PIMS using tunable iv vacuum ultraviolet (VUV) radiation at the chemical dynamics beamline of the Advanced Light Source located at Lawrence Berkley National Laboratory in Berkley, CA. Unlike previous studies of biomass decomposition, these experiments were able to identify the initial pyrolysis products. The first half of this thesis will deal with the thermal decomposition pathways and kinetics of furan and furfural. Earlier G2(MP2) electronic structure calculations predicted that furan will thermally decompose to acetylene, ketene, carbon monoxide, and propyne at lower temperatures. At higher temperatures, these calculations forecast that propargyl radical could result. We see all these products as well as the formation of aromatic hydrocarbons at higher concentrations. This is the first study to show radicals present in biomass decomposition. Thermal decomposition of furfural generates furan and thus follows the same mechanistic pathways as described above. The second half of this manuscript details the thermal decomposition of acetaldehyde and three isotopologues CH3CDO, CD3CHO and CD3CDO as well as benzaldehyde. As thermal decomposition products of CH3CHO, we have identified CH3 (PIMS), CO (IR, PIMS), H (PIMS), H2 (PIMS), CH2CO (IR, PIMS), CH2=CHOH (IR, PIMS) and HCCH (IR, PIMS). The mechanism for decomposition of benzaldehyde is analogous to… Advisors/Committee Members: Gayfree Barney Ellison, Veronica Bierbaum, John Daily.

Subjects/Keywords: biomass; decomposition; reactor; thermal; Physical Chemistry; Physics

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APA (6^th Edition):

Vasiliou, A. K. (2011). Thermochemical Conversion of Biomass: A Molecular Viewpoint. (Doctoral Dissertation). University of Colorado. Retrieved from http://scholar.colorado.edu/chem_gradetds/53

Chicago Manual of Style (16^th Edition):

Vasiliou, AnGayle Konstance. “Thermochemical Conversion of Biomass: A Molecular Viewpoint.” 2011. Doctoral Dissertation, University of Colorado. Accessed February 16, 2019. http://scholar.colorado.edu/chem_gradetds/53.

MLA Handbook (7^th Edition):

Vasiliou, AnGayle Konstance. “Thermochemical Conversion of Biomass: A Molecular Viewpoint.” 2011. Web. 16 Feb 2019.

Vancouver:

Vasiliou AK. Thermochemical Conversion of Biomass: A Molecular Viewpoint. [Internet] [Doctoral dissertation]. University of Colorado; 2011. [cited 2019 Feb 16]. Available from: http://scholar.colorado.edu/chem_gradetds/53.

Council of Science Editors:

Vasiliou AK. Thermochemical Conversion of Biomass: A Molecular Viewpoint. [Doctoral Dissertation]. University of Colorado; 2011. Available from: http://scholar.colorado.edu/chem_gradetds/53

Georgia Tech

8. Remley, Kyle E. Development of methods for high performance computing applications of the deterministic stage of comet calculations.

Degree: PhD, Mechanical Engineering, 2016, Georgia Tech

URL: http://hdl.handle.net/1853/58610

► The Coarse Mesh Radiation Transport (COMET) method is a reactor physics method and code that has been used to solve whole core reactor eigenvalue and… (more)

▼ The Coarse Mesh Radiation Transport (COMET) method is a reactor physics method and code that has been used to solve whole core reactor eigenvalue and flux distribution problems. A strength of the method is its formidable accuracy and computational efficiency. COMET solutions are computed to Monte Carlo accuracy on a single processor in a runtime that is several orders of magnitude faster than stochastic calculations. However, with the growing ubiquity of both shared and distributed memory parallel machines and the desire to extend the method to allow for coupling to multiphysics and on-the-fly response generation, serial implementations of COMET calculations will become less desirable. It is under this motivation that an implementation for a parallel execution of deterministic COMET calculations has been developed. COMET involves inner and outer iterations; inner iterations involve local calculations that can be carried out independently, making the algorithm amenable to parallelization. However, considerations for decomposing a problem and the distribution of data must be made. To allow for efficient parallel implementation of a distributed algorithm, changes to response data access and sweep order are made, along with considerations for communications between processors. The parallel code is implemented on several variants of the C5G7 benchmark problem to assess the scalability of the algorithm, and it is found that problems with larger numbers of coarse meshes increase the scalability of the code, which is an encouraging result. The code is further tested for full core reactor problems, where extremely efficient wall clock times (on the order of minutes) for solutions are achieved. Finally, application of the parallel code to novel implementations of COMET (e.g., problems with high flux expansions) is discussed. Advisors/Committee Members: Rahnema, Farzad (advisor), Petrovic, Bojan (committee member), Zhang, Dingkang (committee member), Morley, Tom (committee member), Haghighat, Alireza (committee member).

Subjects/Keywords: Coarse mesh transport; Parallel computing; Reactor physics

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APA (6^th Edition):

Remley, K. E. (2016). Development of methods for high performance computing applications of the deterministic stage of comet calculations. (Doctoral Dissertation). Georgia Tech. Retrieved from http://hdl.handle.net/1853/58610

Chicago Manual of Style (16^th Edition):

Remley, Kyle E. “Development of methods for high performance computing applications of the deterministic stage of comet calculations.” 2016. Doctoral Dissertation, Georgia Tech. Accessed February 16, 2019. http://hdl.handle.net/1853/58610.

MLA Handbook (7^th Edition):

Remley, Kyle E. “Development of methods for high performance computing applications of the deterministic stage of comet calculations.” 2016. Web. 16 Feb 2019.

Vancouver:

Remley KE. Development of methods for high performance computing applications of the deterministic stage of comet calculations. [Internet] [Doctoral dissertation]. Georgia Tech; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1853/58610.

Council of Science Editors:

Remley KE. Development of methods for high performance computing applications of the deterministic stage of comet calculations. [Doctoral Dissertation]. Georgia Tech; 2016. Available from: http://hdl.handle.net/1853/58610

Oregon State University

9. Peguero, Lara M. Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor.

Degree: MS, Nuclear Engineering, 2016, Oregon State University

URL: http://hdl.handle.net/1957/59812

► Interest in increased fuel supply stability has driven an investigation into possible alternate fuel for use in the WWR-SM research reactor at the Institute of… (more)

▼ Interest in increased fuel supply stability has driven an investigation into possible alternate fuel for use in the WWR-SM research reactor at the Institute of Nuclear Physics in Uzbekistan. The WWR-SM is a high-power, pool-type research reactor currently utilizing IRT-4M fuel made by a single Russian supplier. A candidate for new fuel is the Korean-made High-flux Advanced Neutron Application Reactor (HANARO) fuel rod, which maintains the low 19.75% enrichment of the current fuel but has a different configuration. To examine the safety and operational parameters of the HANARO fuel rods in the WWR-SM core, the Monte Carlo N-Particle transport code is used to model the neutronics and the PLTEMP code from Argonne National Lab is used to compute thermal-hydraulic parameters. While the core structure will remain unchanged, the plate-type IRT-3M assembly models will be replaced with models of assemblies using HANARO fuel rods that preserve the outer dimensions and central flux trap for irradiation of samples. Three candidate assemblies are assessed for neutronic viability. Two contain the standard HANARO fuel rods with different pitches between 16 fuel rods arranged within the space between basic structures of the IRT assemblies. The other assembly contains a modified HANARO fuel rod with a reduced fuel meat radius; this smaller rod allows for more fuel rods in the assembly, thus increasing the heat transfer area. Neutronic results indicate viability of one of the standard HANARO fuel rod assemblies. The modified HANARO fuel rod assemblies do not maintain sufficient excess reactivity when placed in the core configuration to operate at a critical state, making standard WWR-SM operations impossible. Whearas the undermoderated standard HANARO fuel rod assemblies have a significantly positive temperature coefficient of reactivity making them unsafe in positive temperature transient conditions. When 24 optimal pitch assemblies are in the WWR-SM core model, the core has sufficient excess reactivity to maintain a critical state and produces flux levels within the target range of 10¹² to 3 x 10¹⁴ n cm⁻² s⁻¹. Power distributions and depletion analysis suggest that the cycle length for this core would be at least comparable to the 24 day cycle for IRT-3M fuel assemblies, if not better. However, this core configuration has too much excess reactivity and does not meet the operational shutdown margin requirement. Minor modifications could be made to the core configuration to meet the shutdown margin but are not explored here. Thermal-hydraulic analysis of the optimal pitch HANARO fuel rod shows that low flow rates and the existing pressure drop across the WWR-SM core would result in nucleate boiling along the length of the fuel rod, though flow remains stable. In addition, the maximum temperature in the fuel in some simulations reaches temperatures within 20% of the operational design maximum of 350 °C. Temperature results from the PLTEMP code agree well with hand calculations with an average relative difference of 7%. Advisors/Committee Members: Palmer, Todd S. (advisor), Stetz, Albert (committee member).

Subjects/Keywords: Reactor physics; Light water reactors – Uzbekistan

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APA (6^th Edition):

Peguero, L. M. (2016). Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/59812

Chicago Manual of Style (16^th Edition):

Peguero, Lara M. “Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor.” 2016. Masters Thesis, Oregon State University. Accessed February 16, 2019. http://hdl.handle.net/1957/59812.

MLA Handbook (7^th Edition):

Peguero, Lara M. “Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor.” 2016. Web. 16 Feb 2019.

Vancouver:

Peguero LM. Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor. [Internet] [Masters thesis]. Oregon State University; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1957/59812.

Council of Science Editors:

Peguero LM. Feasibility Study of Using HANARO Fuel Rods in WWR-SM Reactor. [Masters Thesis]. Oregon State University; 2016. Available from: http://hdl.handle.net/1957/59812

University of Ontario Institute of Technology

10. Ferguson, Thomas. Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model.

Degree: 2018, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/940

► Superhomogenization (SPH) has gained interest in the industry as a possible method to overcome the inherent limitations of standard homogenization (SH) for full nuclearreactor-core neutronics… (more)

Subjects/Keywords: Applied reactor physics; Superhomogenization; SPH factors; PHWR

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APA (6^th Edition):

Ferguson, T. (2018). Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/940

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Ferguson, Thomas. “Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model.” 2018. Thesis, University of Ontario Institute of Technology. Accessed February 16, 2019. http://hdl.handle.net/10155/940.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Ferguson, Thomas. “Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model.” 2018. Web. 16 Feb 2019.

Vancouver:

Ferguson T. Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model. [Internet] [Thesis]. University of Ontario Institute of Technology; 2018. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/10155/940.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Ferguson T. Investigation of a SPH-based sub-cell homogenization for PHWR using a multi-cell model. [Thesis]. University of Ontario Institute of Technology; 2018. Available from: http://hdl.handle.net/10155/940

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

University of Cambridge

11. Lindley, Benjamin A. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel.

Degree: PhD, 2015, University of Cambridge

URL: https://www.repository.cam.ac.uk/handle/1810/247162 ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.642470

► Light water reactors (LWRs) are the world’s dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs… (more)

▼ Light water reactors (LWRs) are the world’s dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs or fast reactors. The thermal neutron spectrum in LWRs is less suitable for burning TRUs as this causes a build-up of TRU isotopes with low fission probability. This increases the fissile feed requirements, which tends to result in a positive void coefficient (VC) and hence the reactor is unsafe to operate. Use of reduced-moderation LWRs can improve TRU transmutation performance, but the VC is still severely limiting for these designs. Reduced-moderation pressurized water reactors (RMPWRs) and boiling water reactors (RBWRs) are considered in this study. Using thorium (Th) instead of U as the fertile fuel component can greatly improve the VC. However, Th-based transmutation is a much less developed technology than U-based transmutation. In this thesis, the feasibility and fuel cycle performance of full TRU recycle in Th-fuelled RMPWRs and RBWRs are evaluated. Neutronic performance is greatly improved by spatial separation of TRU and 233-6U, primarily implemented here using heterogeneous RMPWR and RBWR assembly designs. In a RMPWR, the water to fuel ratio must be reduced to around 50% of the normal value to allow full actinide recycle. If implemented by retrofitting an existing PWR, steady-state thermal-hydraulic constraints can still be satisfied. However, in a large break loss-of-coolant accident, the emergency core cooling system may not be able to provide water to the core quickly enough to prevent fuel cladding failure. A discharge burn-up of ~40 GWd/t is possible in RMPWRs. Reactivity control is a challenge due to the reduced worth of neutron absorbers in the hard neutron spectrum, and their detrimental effect on the VC, especially when diluted, as for soluble boron. Control rods are instead used to control the core. It appears possible to achieve adequate power peaking, shutdown margin and rod-ejection accident response. In RBWRs, it appears neutronically feasible to achieve very high burn-ups (~120 GWd/t) but the maximum achievable incineration rate is less than in RMPWRs. The reprocessing and fuel fabrication requirements of RBWRs are less than RMPWRs but more than fast reactors. A two-stage TRU burning cycle, where the first stage is Th-Pu MOX in a conventional PWR feeding a second stage continuous burn in a RBWR, is technically reasonable. It is possible to limit the core area to that of an ABWR with acceptable thermal-hydraulic performance. In this case, it appears that RBWRs are of similar cost to inert matrix incineration in LWRs, and lower cost than RMPWRs and Th- and U-based fast reactor recycle schemes.

Subjects/Keywords: 539.7; Thorium; Light Water Reactor; Plutonium; Transuranic; Reduced-moderation; Nuclear fuel cycle; Reactor physics

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APA (6^th Edition):

Lindley, B. A. (2015). The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel. (Doctoral Dissertation). University of Cambridge. Retrieved from https://www.repository.cam.ac.uk/handle/1810/247162 ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.642470

Chicago Manual of Style (16^th Edition):

Lindley, Benjamin A. “The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel.” 2015. Doctoral Dissertation, University of Cambridge. Accessed February 16, 2019. https://www.repository.cam.ac.uk/handle/1810/247162 ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.642470.

MLA Handbook (7^th Edition):

Lindley, Benjamin A. “The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel.” 2015. Web. 16 Feb 2019.

Vancouver:

Lindley BA. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel. [Internet] [Doctoral dissertation]. University of Cambridge; 2015. [cited 2019 Feb 16]. Available from: https://www.repository.cam.ac.uk/handle/1810/247162 ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.642470.

Council of Science Editors:

Lindley BA. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel. [Doctoral Dissertation]. University of Cambridge; 2015. Available from: https://www.repository.cam.ac.uk/handle/1810/247162 ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.642470

12. Berggren, Alexander. Development of a Mobile Reactor for Large Scale Water Treatment.

Degree: Physics, 2019, Umeå University

URL: http://urn.kb.se/resolve?urn=urn:nbn:se:umu:diva-155584

► Water pollution is one of many environmental problems that currently exists and inadequate treatment of industrial wastewater is contributing to further pollution. SpinChem AB's… (more)

Subjects/Keywords: Water treatment; Mobile reactor; Autopilot; Rotating Bed Reactor; Other Physics Topics; Annan fysik

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APA (6^th Edition):

Berggren, A. (2019). Development of a Mobile Reactor for Large Scale Water Treatment. (Thesis). Umeå University. Retrieved from http://urn.kb.se/resolve?urn=urn:nbn:se:umu:diva-155584

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Berggren, Alexander. “Development of a Mobile Reactor for Large Scale Water Treatment.” 2019. Thesis, Umeå University. Accessed February 16, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:umu:diva-155584.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Berggren, Alexander. “Development of a Mobile Reactor for Large Scale Water Treatment.” 2019. Web. 16 Feb 2019.

Vancouver:

Berggren A. Development of a Mobile Reactor for Large Scale Water Treatment. [Internet] [Thesis]. Umeå University; 2019. [cited 2019 Feb 16]. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:umu:diva-155584.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Berggren A. Development of a Mobile Reactor for Large Scale Water Treatment. [Thesis]. Umeå University; 2019. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:umu:diva-155584

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

KTH

13. Luszczek, Karol. Validation and Benchmarking of Westinghouse BWR lattice physics methods.

Degree: Reactor Technology, 2015, KTH

URL: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563

► A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to… (more)

▼ A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to becalculated and generates a proper set of data to be used by a 3-D full coresimulator. Due to advancement and complexity of modern Boiling WaterReactor assembly designs, a new deterministic lattice physics codeis being developed at Westinghouse Sweden AB, namely PHOENIX5.Each time a new code is written, its methodology of solving the neutrontransport equation, has to be validated to make sure it providesreliable output. In a wake of preparation for PHOENIX5 release andconsecutive validation efforts, a set of reference Monte Carlo calculationswas prepared, using the code Serpent. A depletion calculation with achosen type of branch cases was conducted. Methods implemented inPHOENIX5 are based on the Current Coupling Collision Probabilitymethod used in older versions of the code HELIOS. Therefore, a comparisonbetween reference Monte Carlo simulations and HELIOS 1.8.1is made, in order to discover problems inherent to the said method ofsolving the neutron transport equation. A special care should be givenduring PHOENIX5 validation, to issues highlighted in this work.Discrepancies in results of Serpent and HELIOS are attributed mostlyto disparities in the basic nuclear data used by the codes, as well as arange of approximations and corrections adopted by the deterministiccode.Serpent and HELIOS showed a good agreement in a typical voidrange (up to 90 % void) and ‘less’ challenging branches (coolant void,fuel temperature and spacer grid branches). More significant discrepanciesappeared for extreme cases with a very high void and control rodpresence (k1 differences as high as 1000 pcm) and rather pronouncedconcentrations of the natural boron dissolved in coolant (absolute differencesroughly at a level of 900 pcm). The issues do not seem to stemsolely from discrepancies in the nuclear data libraries used by Serpentand HELIOS.Moreover, a coolant void bias was consistently found in the resultsof branch calculation at changing coolant void. This confirms the analogousphenomenon found in previous studies of the CCCP based deterministiccodes. It most probably stems from the assumptions used bythe method while tackling the neutron transport equation, such as theflat source approximation, the isotropic scattering assumption and thetransport correction. An alternative transport correction approximationis proposed to alleviate this issue.

Subjects/Keywords: Serpent; HELIOS; lattice physics; benchmarking; Monte Carlo; neutronics; reactor physics

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Luszczek, K. (2015). Validation and Benchmarking of Westinghouse BWR lattice physics methods. (Thesis). KTH. Retrieved from http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Luszczek, Karol. “Validation and Benchmarking of Westinghouse BWR lattice physics methods.” 2015. Thesis, KTH. Accessed February 16, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Luszczek, Karol. “Validation and Benchmarking of Westinghouse BWR lattice physics methods.” 2015. Web. 16 Feb 2019.

Vancouver:

Luszczek K. Validation and Benchmarking of Westinghouse BWR lattice physics methods. [Internet] [Thesis]. KTH; 2015. [cited 2019 Feb 16]. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Luszczek K. Validation and Benchmarking of Westinghouse BWR lattice physics methods. [Thesis]. KTH; 2015. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

University of Ontario Institute of Technology

14. Carisse, Katarzyna. Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.

Degree: 2014, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/458

► The neutron flux throughout a reactor core should be calculated, ideally, by solving the neutron transport equation for a highly detailed geometric model of the… (more)

Subjects/Keywords: Applied reactor physics; Advanced homogenization; CANDU; PT-SCWR

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Carisse, K. (2014). Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/458

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Carisse, Katarzyna. “Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.” 2014. Thesis, University of Ontario Institute of Technology. Accessed February 16, 2019. http://hdl.handle.net/10155/458.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Carisse, Katarzyna. “Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.” 2014. Web. 16 Feb 2019.

Vancouver:

Carisse K. Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR. [Internet] [Thesis]. University of Ontario Institute of Technology; 2014. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/10155/458.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Carisse K. Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR. [Thesis]. University of Ontario Institute of Technology; 2014. Available from: http://hdl.handle.net/10155/458

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

15. Domingos, Douglas Borges. Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB.

Degree: PhD, Tecnologia Nuclear - Reatores, 2014, University of São Paulo

URL: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-17122014-133601/ ;

►

Neste trabalho foi realizado uma comparação entre três tipos de alvos (UAl2-Al, U-Ni cilíndrico e U-Ni placa) para a produção de 99Mo por fissão do… (more)

▼

Neste trabalho foi realizado uma comparação entre três tipos de alvos (UAl2-Al, U-Ni cilíndrico e U-Ni placa) para a produção de 99Mo por fissão do 235U. Para isso foram desenvolvidas análises neutrônicas e termo-hidráulicas. Também foram realizados experimentos para se validar as metodologias de cálculo termo-hidráulica e neutrônica utilizadas neste trabalho. Para os cálculos neutrônicos foram utilizados os programas NJOY99.0, AMPX-II e HAMMERTECHNION, para geração das seções de choque, e os programas SCALE 6.0 e CITATION para os cálculos tridimensionais dos núcleos, queima do combustível e produção de 99Mo. Para os cálculos termo-hidráulicos foram utilizados os programas MTRCRIEAR1 e ANSYS CFX para calcular as variáveis térmicas e hidráulicas dos dispositivos de irradiação e compará-las a limites e critérios de projeto estabelecidos. Primeiro foram realizadas análises neutrônicas e termo-hidráulicas para o reator IEA-R1 com os alvos de UAl2-Al (10 miniplacas). As análises demonstraram que a atividade total obtida para o 99Mo nas miniplacas não atende à demanda dos hospitais brasileiros (450 Ci/semana) e que nenhum limite de projeto termo-hidráulico é ultrapassado. Em seguida foram realizados os mesmos cálculos para os três tipos de alvo no Reator Multipropósito Brasileiro (RMB). As análises neutrônicas demonstraram que os três alvos podem atender à demanda dos hospitais brasileiros. As análises termo-hidráulicos demonstram que será necessário uma velocidade mínima no dispositivo de irradiação de 7 m/s para o UAl2, de 8 m/s para o alvo de U-Ni cilíndrico e de 9 m/s para o alvo de U-Ni placa para que nenhum limite de projeto seja ultrapassado. Foram realizados experimentos em uma bancada de aferição de vazão para se validar a metodologia de cálculo termo-hidráulico. Os experimentos realizados para se validar os cálculos neutrônicos foram feitos no reator IPEN/MB-01. Todos os experimentos foram simulados com as metodologias acima descritas e os resultados comparados entre si. Os resultados das simulações apresentaram boa concordância com os resultados experimentais.

In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl2-Al, U-Ni cylindrical and U-Ni plate) used for the production of 99Mo by fission of 235U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of 99Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl2-Al (10 miniplates). Analyses have shown that the total activity…

Advisors/Committee Members: Santos, Adimir dos, Silva, Antonio Teixeira e.

Subjects/Keywords: física de reatores; radiofármacos; radiopharmaceuticals; reactor physics; termo-hidráulica; thermo-hydraulic

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APA (6^th Edition):

Domingos, D. B. (2014). Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB. (Doctoral Dissertation). University of São Paulo. Retrieved from http://www.teses.usp.br/teses/disponiveis/85/85133/tde-17122014-133601/ ;

Chicago Manual of Style (16^th Edition):

Domingos, Douglas Borges. “Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB.” 2014. Doctoral Dissertation, University of São Paulo. Accessed February 16, 2019. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-17122014-133601/ ;.

MLA Handbook (7^th Edition):

Domingos, Douglas Borges. “Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB.” 2014. Web. 16 Feb 2019.

Vancouver:

Domingos DB. Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB. [Internet] [Doctoral dissertation]. University of São Paulo; 2014. [cited 2019 Feb 16]. Available from: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-17122014-133601/ ;.

Council of Science Editors:

Domingos DB. Análises neutrônica e termo-hidráulica de dispositivos para irradiação de alvos tipo LEU de UAlx-Al e U-Ni para produção de Mo-99 nos reatores IEA-R1 e RMB. [Doctoral Dissertation]. University of São Paulo; 2014. Available from: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-17122014-133601/ ;

The Ohio State University

16. Abejon Orzaez, Jorge. Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor.

Degree: MS, Nuclear Engineering, 2009, The Ohio State University

URL: http://rave.ohiolink.edu/etdc/view?acc_num=osu1238045558

► The objective of this research is to, based on the original design for the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), develop an MCNPX model… (more)

Subjects/Keywords: Engineering; Nuclear Physics; Neutronics; advanced high temperature reactor; criticality

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APA (6^th Edition):

Abejon Orzaez, J. (2009). Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor. (Masters Thesis). The Ohio State University. Retrieved from http://rave.ohiolink.edu/etdc/view?acc_num=osu1238045558

Chicago Manual of Style (16^th Edition):

Abejon Orzaez, Jorge. “Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor.” 2009. Masters Thesis, The Ohio State University. Accessed February 16, 2019. http://rave.ohiolink.edu/etdc/view?acc_num=osu1238045558.

MLA Handbook (7^th Edition):

Abejon Orzaez, Jorge. “Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor.” 2009. Web. 16 Feb 2019.

Vancouver:

Abejon Orzaez J. Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor. [Internet] [Masters thesis]. The Ohio State University; 2009. [cited 2019 Feb 16]. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1238045558.

Council of Science Editors:

Abejon Orzaez J. Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor. [Masters Thesis]. The Ohio State University; 2009. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1238045558

Georgia Tech

17. Keller, Steven Ede. Flux-limited Diffusion Coefficient Applied to Reactor Analysis.

Degree: PhD, Nuclear Engineering, 2007, Georgia Tech

URL: http://hdl.handle.net/1853/16126

► A new definition of the diffusion coefficient for use in reactor physics calculations is evaluated in this thesis. It is based on naturally flux-limited diffusion… (more)

Subjects/Keywords: Transport cross section; Reactor physics

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APA (6^th Edition):

Keller, S. E. (2007). Flux-limited Diffusion Coefficient Applied to Reactor Analysis. (Doctoral Dissertation). Georgia Tech. Retrieved from http://hdl.handle.net/1853/16126

Chicago Manual of Style (16^th Edition):

Keller, Steven Ede. “Flux-limited Diffusion Coefficient Applied to Reactor Analysis.” 2007. Doctoral Dissertation, Georgia Tech. Accessed February 16, 2019. http://hdl.handle.net/1853/16126.

MLA Handbook (7^th Edition):

Keller, Steven Ede. “Flux-limited Diffusion Coefficient Applied to Reactor Analysis.” 2007. Web. 16 Feb 2019.

Vancouver:

Keller SE. Flux-limited Diffusion Coefficient Applied to Reactor Analysis. [Internet] [Doctoral dissertation]. Georgia Tech; 2007. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1853/16126.

Council of Science Editors:

Keller SE. Flux-limited Diffusion Coefficient Applied to Reactor Analysis. [Doctoral Dissertation]. Georgia Tech; 2007. Available from: http://hdl.handle.net/1853/16126

University of Manchester

18. Black, Greg. Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal.

Degree: Thesis (Eng.D.), 2014, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547

► The University of Manchester Greg BlackThesis submitted for the degree of Doctor of EngineeringIrradiated Graphite Waste: Analysis and Modelling of Radionuclide Production with a View… (more)

▼ The University of Manchester Greg BlackThesis submitted for the degree of Doctor of EngineeringIrradiated Graphite Waste: Analysis and Modelling of Radionuclide Production with a View to Long Term Disposal23rd June 2014The UK has predominantly used graphite moderator reactor designs in both its research and civil nuclear programmes. This material will become activated during operation and, once all reactors are shutdown, will represent a waste legacy of 96,000 tonnes [1]. The safe and effective management of this material will require a full understanding of the final radiological inventory. The activity is known to arise from impurities present in the graphite at start of life as well as from contamination products transported from other components in the reactor circuit. The process is further complicated by radiolytic oxidation which leads to considerable weightloss of the graphite components. A comprehensive modelling methodology has been developed and validated to estimate the activity of the principle radionuclides of concern, 3H, 14C, 36Cl and 60Co. This methodology involves the simulation of neutron flux using the reactor physics code WIMS, and radiation transport code MCBEND. Activation calculations have been performed using the neutron activation software FISPACT. The final methodology developed allows full consideration of all processes which may contribute to the final radiological inventory of the material. The final activity and production pathway of each radionuclide has been researched in depth, as well as operational parameters such as the effect of changes in flux, fuel burnup, graphite weightloss and irradiation time. Methods to experimentally determine the activity, and distribution of key radionuclides within irradiated graphite samples have been developed in this research using a combination of both gamma spectroscopy and autoradiography. This work has been externally validated and provides confidence in the accuracy of the final modelling predictions. This work has been undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE, and was funded by the Office for Nuclear Regulation.

Subjects/Keywords: 621.48; Nuclear Graphite; Radioactive Waste Management; Reactor Physics Modelling; Gamma Spectroscopy

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APA (6^th Edition):

Black, G. (2014). Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547

Chicago Manual of Style (16^th Edition):

Black, Greg. “Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal.” 2014. Doctoral Dissertation, University of Manchester. Accessed February 16, 2019. https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547.

MLA Handbook (7^th Edition):

Black, Greg. “Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal.” 2014. Web. 16 Feb 2019.

Vancouver:

Black G. Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal. [Internet] [Doctoral dissertation]. University of Manchester; 2014. [cited 2019 Feb 16]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547.

Council of Science Editors:

Black G. Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal. [Doctoral Dissertation]. University of Manchester; 2014. Available from: https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547

Georgia Tech

19. Kooreman, Gabriel. Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport.

Degree: PhD, Mechanical Engineering, 2016, Georgia Tech

URL: http://hdl.handle.net/1853/59134

► The neutron transport equation often is homogenized in order to simplify its solution procedure in some manner or another. There exist many methods for homogenizing… (more)

▼ The neutron transport equation often is homogenized in order to simplify its solution procedure in some manner or another. There exist many methods for homogenizing the neutron transport equation with different benefits and detriments. One promising method is the Consistent Spatial Homogenization (CSH) method developed and implemented in 1-D by Yasseri and Rahnema. The method, along with its successor, the Diffusion-Transport Homogenization (DTH) method are promising for their ability to reconstruct accurate fine-mesh angular flux profiles as well as reactor eigenvalue after a re-homogenization procedure. This work will explore the extension of both the CSH and DTH methods to higher spatial dimensionality in order to solve large-scale reactor eigenvalue problems. The CSH and DTH methods are based around iterated re-homogenization of the neutron transport equation with an auxiliary source term which is used to correct for heterogeneity effects of a given problem. The net effect of this is that the effects of heterogeneity are relegated to a source term, and the homogenized neutron transport equation is solved instead of the heterogeneous equation. This allows for implementation of simpler acceleration techniques to improve the speed and accuracy of the homogenized problem and in multiple dimensions helps to avoid the effects of complicated reactor geometries. The re-homogenization procedure brings the flux solution back to the heterogeneous discretization in order to generate better approximations for the homogenized cross sections, a better approximation of the auxiliary source term, and most importantly to reconstruct the full heterogeneous angular flux profile. In this work, the CSH and DTH methods are modified for increased spatial dimensionality and implemented using a 2-D SN discrete ordinates transport solver. This implementation is tested using Cartesian-mesh variants of the 2D-C5G7 benchmark problem and a 2-D full-scale boiling water reactor (BWR) benchmark problem. Advisors/Committee Members: Rahnema, Farzad (advisor), Petrovic, Bojan (committee member), Zhang, Dingkang (committee member), Morley, Tom (committee member), Densmore, Jeffery (committee member).

Subjects/Keywords: Neutron transport; Homogenization; Diffusion; Eigenvalue calculation; Reactor physics

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APA (6^th Edition):

Kooreman, G. (2016). Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport. (Doctoral Dissertation). Georgia Tech. Retrieved from http://hdl.handle.net/1853/59134

Chicago Manual of Style (16^th Edition):

Kooreman, Gabriel. “Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport.” 2016. Doctoral Dissertation, Georgia Tech. Accessed February 16, 2019. http://hdl.handle.net/1853/59134.

MLA Handbook (7^th Edition):

Kooreman, Gabriel. “Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport.” 2016. Web. 16 Feb 2019.

Vancouver:

Kooreman G. Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport. [Internet] [Doctoral dissertation]. Georgia Tech; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1853/59134.

Council of Science Editors:

Kooreman G. Iterative homogenization method for improving computational efficiency in solving eigenvalue problems in neutron transport. [Doctoral Dissertation]. Georgia Tech; 2016. Available from: http://hdl.handle.net/1853/59134

University of Illinois – Urbana-Champaign

20. McKenzie, George. Modern Rossi alpha measurements.

Degree: MS, 5183, 2015, University of Illinois – Urbana-Champaign

URL: http://hdl.handle.net/2142/72834

► The Rossi-α method determines the prompt neutron decay constant in a nuclear fissioning system at or near delayed critical. Knowledge of the prompt neutron decay… (more)

▼ The Rossi-α method determines the prompt neutron decay constant in a nuclear fissioning system at or near delayed critical. Knowledge of the prompt neutron decay constant is important for a critical system as it is a major contributor to the dynamic system behavior. The classical method for the Rossi experiment used gated circuitry to track the time when a neutron was incident upon the detector. The downside of this method is that the circuitry was complex and only one single fission chain could be measured at a time. The modern method allows many chains to be measured simultaneously by a pulse time tagging system such as the LANL custom designed List-mode module. This thesis examines the implementation of the modern Rossi-α method on the all highly enriched uranium, HEU, Zeus experiment. Measurements are taken at several subcritical configurations, at critical in the presence of a source, and at one supercritical point. During the experiment, the List-mode module generates time tags of incoming neutron pulses. After the experiment, this list of neutron pulses is complied using custom software into a histogram. This histogram is fit using off the shelf graphing software to determine the value of α. The subcritical measurements of α are used to extrapolate α at delayed critical. The extrapolation determined the value of α at delayed critical to be α = 89910 s-1. This value is compared to the measured value of α at delayed critical which is determined to be α = 90408.4 s-1. These values differ by 0.55% which is remarkably good agreement. This thesis also examines the expected value of α using a Monte Carlo transport code, MCNP. MCNP determined the value of α at delayed critical to be 100048 ± 0.584 s-1. This result differs by 11.3% from the extrapolated value of α determined experimentally. When compared to systems with similar neutron spectra, the measured value of α fits well in comparison to historical measurements. Advisors/Committee Members: Kozlowski, Tomasz (advisor).

Subjects/Keywords: Reactor Physics; Rossi Alpha; Zeus Experiment; Neutron Measurements

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APA (6^th Edition):

McKenzie, G. (2015). Modern Rossi alpha measurements. (Thesis). University of Illinois – Urbana-Champaign. Retrieved from http://hdl.handle.net/2142/72834

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

McKenzie, George. “Modern Rossi alpha measurements.” 2015. Thesis, University of Illinois – Urbana-Champaign. Accessed February 16, 2019. http://hdl.handle.net/2142/72834.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

McKenzie, George. “Modern Rossi alpha measurements.” 2015. Web. 16 Feb 2019.

Vancouver:

McKenzie G. Modern Rossi alpha measurements. [Internet] [Thesis]. University of Illinois – Urbana-Champaign; 2015. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/2142/72834.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

McKenzie G. Modern Rossi alpha measurements. [Thesis]. University of Illinois – Urbana-Champaign; 2015. Available from: http://hdl.handle.net/2142/72834

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

North Carolina State University

21. Hiruta, Hikaru. Advanced Computational Methodology for Full-Core Neutronics Calculations.

Degree: PhD, Nuclear Engineering, 2004, North Carolina State University

URL: http://www.lib.ncsu.edu/resolver/1840.16/5463

► The modern computational methodology for reactor physics calculations is based on single–assembly transport calculations with reflective boundary conditions that generate homogenized few–group data, and core–level… (more)

▼ The modern computational methodology for reactor physics calculations is based on single–assembly transport calculations with reflective boundary conditions that generate homogenized few–group data, and core–level coarse-mesh diffusion calculations that evaluate a large-scale behavior of the scalar flux. Recently, an alternative approach has been developed. It is based on the low-order equations of the quasidiffusion (QD) method in order to account accurately for complicated transport effects in full–core calculations. The LOQD equations can capture transport effects to an arbitrary degree of accuracy. This approach is combined with single–assembly transport calculations that use special albedo boundary conditions which enable one to simulate efficiently effects of an unlike neighboring assembly on assembly's group data. In this dissertation, we develop homogenization methodology based on the LOQD equations and spatially consistent coarse–mesh finite element discretization methods for the 2D low–order quasidiffusion equations for the full–core calculations. The coarse–mesh solution generated by this method preserves a number of spatial polynomial moments of the fine–mesh transport solution over coarse cells. The proposed method reproduces accurately the complicated large–scale behavior of the transport solution within assemblies. To demonstrate accuracy of the developed method, we present numerical results of calculations of test problems that simulate interaction of MOX and uranium assemblies. We also develop a splitting method that can efficiently solve coarse-mesh discretized low-order quasidiffusion (LOQD) equations. The presented method splits the LOQD problem into two parts: (i) the D-problem that captures a significant part of transport solution in the central parts of assemblies and can be reduced to a diffusion-type equation, and (ii) the Q-problem that accounts for the complicated behavior of the transport solution near assembly boundaries. Independent coarse-mesh discretizations are applied: the D-problem equations are approximated by means of a finite-element method, whereas the Q-problem equations are discretized using a finite-volume method. Numerical results demonstrate the efficiency of the presented methodology. This methodology can be used to modify existing diffusion codes for full-core calculations (which already solve a version of the D-problem) to account for transport effects. Advisors/Committee Members: Dmitriy Y. Anistratov, Committee Chair (advisor), Paul J. Turinsky, Committee Member (advisor), Robin P. Gardner, Committee Member (advisor), Zhilin Li, Committee Member (advisor).

Subjects/Keywords: Reactor Physics; Neutron Transport

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APA (6^th Edition):

Hiruta, H. (2004). Advanced Computational Methodology for Full-Core Neutronics Calculations. (Doctoral Dissertation). North Carolina State University. Retrieved from http://www.lib.ncsu.edu/resolver/1840.16/5463

Chicago Manual of Style (16^th Edition):

Hiruta, Hikaru. “Advanced Computational Methodology for Full-Core Neutronics Calculations.” 2004. Doctoral Dissertation, North Carolina State University. Accessed February 16, 2019. http://www.lib.ncsu.edu/resolver/1840.16/5463.

MLA Handbook (7^th Edition):

Hiruta, Hikaru. “Advanced Computational Methodology for Full-Core Neutronics Calculations.” 2004. Web. 16 Feb 2019.

Vancouver:

Hiruta H. Advanced Computational Methodology for Full-Core Neutronics Calculations. [Internet] [Doctoral dissertation]. North Carolina State University; 2004. [cited 2019 Feb 16]. Available from: http://www.lib.ncsu.edu/resolver/1840.16/5463.

Council of Science Editors:

Hiruta H. Advanced Computational Methodology for Full-Core Neutronics Calculations. [Doctoral Dissertation]. North Carolina State University; 2004. Available from: http://www.lib.ncsu.edu/resolver/1840.16/5463

22. Talley, Kemper Dyar. Beta-Delayed Neutron Data and Models for SCALE.

Degree: 2016, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_graddiss/4170

► Recent advancements in experimental and theoretical nuclear physics have yielded new data and models that more accurately describe the decay of fission products compared to… (more)

Subjects/Keywords: Neutron Emission; Beta Decay; Nuclear Physics; Reactor Physics; Nuclear Data; SCALE; Nuclear; Nuclear Engineering

8 more images…

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APA (6^th Edition):

Talley, K. D. (2016). Beta-Delayed Neutron Data and Models for SCALE. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/4170

Chicago Manual of Style (16^th Edition):

Talley, Kemper Dyar. “Beta-Delayed Neutron Data and Models for SCALE.” 2016. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed February 16, 2019. https://trace.tennessee.edu/utk_graddiss/4170.

MLA Handbook (7^th Edition):

Talley, Kemper Dyar. “Beta-Delayed Neutron Data and Models for SCALE.” 2016. Web. 16 Feb 2019.

Vancouver:

Talley KD. Beta-Delayed Neutron Data and Models for SCALE. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2016. [cited 2019 Feb 16]. Available from: https://trace.tennessee.edu/utk_graddiss/4170.

Council of Science Editors:

Talley KD. Beta-Delayed Neutron Data and Models for SCALE. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2016. Available from: https://trace.tennessee.edu/utk_graddiss/4170

23. Serra, André da Silva. Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem.

Degree: PhD, Física, 2012, University of São Paulo

URL: http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/ ;

►

O presente trabalho visa contribuir com o desenvolvimento sistemático de novas metodologias experimentais da medida da reatividade de arranjos físseis subcríticos, utilizando: estatísticas de alta… (more)

▼

O presente trabalho visa contribuir com o desenvolvimento sistemático de novas metodologias experimentais da medida da reatividade de arranjos físseis subcríticos, utilizando: estatísticas de alta ordem das contagens de nêutrons com detectores no modo pulso, o recente conceito de reatividade generalizada, e as instalações do reator IPEN/MB-01. Este trabalho reuniu em um só texto diversos aspectos da implementação destes tipos de medidas. Diferentemente das demais técnicas utilizadas nas medidas da reatividade subcrítica, as metodologias apresentadas neste trabalho tem o potencial para permitir a medida experimental da reatividade subcrítica sem a necessidade da estimativa prévia de quaisquer outros parâmetros cinéticos, obtidos de forma teórica ou experimental, calibração de fontes externas ou detectores.A princípio, os métodos estatísticos de alta ordem das contagens de nêutrons permitem obter diretamente o valor da subcriticalidade (ou o fator de multiplicação) de um arranjo físsil, independentemente do modelo de física subcrítica utilizado, sem a utilização de infra-estrutura diferenciada (como uma fonte pulsada de nêutrons), sendo uma extensão natural das metodologias que utilizam estatísticas de ordens inferiores - por exemplo, Feymann-. E este conteúdo estatístico diferenciado dos momentos de altas ordens das contagens de nêutrons, o principal motivador da implementação deste trabalho. Apesar de suas potencialidades, a implementação experimental do método esbarra no tempo e taxa de aquisição de dados; ou seja, na quantidade de conteúdo estatístico necessária para a obtenção de medida útil. Exatamente esta dificuldade impediu a obtenção de uma medida útil/prática nas instalações do reator IPEN/MB-01. Existem, entretanto, outras formas de explorar estatísticas ordem superior. Por exemplo, uma extensão do método de Rossi- sugerida neste trabalho pode utilizar auto bi-correlações (coincidências triplas não acidentais de contagens). A despeito do alto valor das incertezas, os aspectos estatísticos fundamentais de uma medida foram preservados nos métodos empregados neste trabalho. O método das auto bicorrelações é conceitualmente mais robusto contra as influências do tempo morto do sistema de aquisição de dados. Ao longo de sua execução, o presente trabalho visou preencher algumas lacunas de procedimentos experimentais aparentemente pouco abordadas por outros autores, permitindo estabelecer métodos estatisticamente mais rigorosos. Entre as contribuições neste sentido destacam-se, entre outras, as correções por tempo morto ou as geradas pela correlação entre os parâmetros estatísticos em tela. Do ponto de vista teórico, este trabalho sugere duas maneiras originais de abordar o mesmo problema da utilização de estatísticas de altas ordens: (a) auto bicorrelações; e (2) os biespectros de densidade de espectral de potência própria, sendo o primeiro explorado experimentalmente/estatisticamente em detalhes.

The present work aims to contribute to the systematic development of new experimental methods of measuring the…

Advisors/Committee Members: Pascholati, Paulo Reginaldo, Santos, Adimir dos.

Subjects/Keywords: cinética de reatores; estatísticas; física de reatores; nuclear reactors; reactivity; reactor kinetics; reactor physics; reatividade; reatores nucleares; statistics

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Serra, A. d. S. (2012). Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem. (Doctoral Dissertation). University of São Paulo. Retrieved from http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/ ;

Chicago Manual of Style (16^th Edition):

Serra, André da Silva. “Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem.” 2012. Doctoral Dissertation, University of São Paulo. Accessed February 16, 2019. http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/ ;.

MLA Handbook (7^th Edition):

Serra, André da Silva. “Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem.” 2012. Web. 16 Feb 2019.

Vancouver:

Serra AdS. Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem. [Internet] [Doctoral dissertation]. University of São Paulo; 2012. [cited 2019 Feb 16]. Available from: http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/ ;.

Council of Science Editors:

Serra AdS. Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem. [Doctoral Dissertation]. University of São Paulo; 2012. Available from: http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/ ;

Texas A&M University

24. Sternat, Matthew Ryan 1982-. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel.

Degree: 2012, Texas A&M University

URL: http://hdl.handle.net/1969.1/148202

► Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis… (more)

▼ Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample?s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope… Advisors/Committee Members: Charlton, William S (advisor), Boyle, David R (committee member), Adams, Marvin L (committee member), Folden, Charles M (committee member).

Subjects/Keywords: Attribution; Spent Fuel; Research Reactor; Savannah River National Laboratory; Radiochemical Analysis; Inverse Reactor Physics; Inverse Problems; Nuclear Forensics

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APA (6^th Edition):

Sternat, M. R. 1. (2012). Development of Technical Nuclear Forensics for Spent Research Reactor Fuel. (Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/148202

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Sternat, Matthew Ryan 1982-. “Development of Technical Nuclear Forensics for Spent Research Reactor Fuel.” 2012. Thesis, Texas A&M University. Accessed February 16, 2019. http://hdl.handle.net/1969.1/148202.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Sternat, Matthew Ryan 1982-. “Development of Technical Nuclear Forensics for Spent Research Reactor Fuel.” 2012. Web. 16 Feb 2019.

Vancouver:

Sternat MR1. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel. [Internet] [Thesis]. Texas A&M University; 2012. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/1969.1/148202.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Sternat MR1. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel. [Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/148202

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

KTH

25. Kotchoubey, Jurij. POLCA-T Neutron Kinetics Model Benchmarking.

Degree: Reactor Technology, 2015, KTH

URL: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096

► The demand for computational tools that are capable to reliably predict the behavior of a nuclear reactor core in a variety of static and… (more)

Subjects/Keywords: POLCA-T; Neutronics; Neutron Kinetics; Benchmarking; Reactor Kinetics; Space-Time Kinetics; Reactor Physics; POLCA-T; Kinetik; Neutronkinetik; Benchmarking; Reaktorfysik

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APA (6^th Edition):

Kotchoubey, J. (2015). POLCA-T Neutron Kinetics Model Benchmarking. (Thesis). KTH. Retrieved from http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Kotchoubey, Jurij. “POLCA-T Neutron Kinetics Model Benchmarking.” 2015. Thesis, KTH. Accessed February 16, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Kotchoubey, Jurij. “POLCA-T Neutron Kinetics Model Benchmarking.” 2015. Web. 16 Feb 2019.

Vancouver:

Kotchoubey J. POLCA-T Neutron Kinetics Model Benchmarking. [Internet] [Thesis]. KTH; 2015. [cited 2019 Feb 16]. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Kotchoubey J. POLCA-T Neutron Kinetics Model Benchmarking. [Thesis]. KTH; 2015. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Missouri University of Science and Technology

26. Compton, William Kirby. Impact of configuration variations on small modular reactor core performance.

Degree: M.S. in Nuclear Engineering, Nuclear Engineering, Missouri University of Science and Technology

URL: http://scholarsmine.mst.edu/masters_theses/7391

► "One of the most promising new reactor designs is the Small Modular Reactor (SMR). These reactors, which operate under 300 MWe, will help bring… (more)

Subjects/Keywords: MCNP; Reactor Physics; Small Modular Reactor; Nuclear Engineering

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Compton, W. K. (n.d.). Impact of configuration variations on small modular reactor core performance. (Masters Thesis). Missouri University of Science and Technology. Retrieved from http://scholarsmine.mst.edu/masters_theses/7391

Note: this citation may be lacking information needed for this citation format:
No year of publication.

Chicago Manual of Style (16^th Edition):

Compton, William Kirby. “Impact of configuration variations on small modular reactor core performance.” Masters Thesis, Missouri University of Science and Technology. Accessed February 16, 2019. http://scholarsmine.mst.edu/masters_theses/7391.

Note: this citation may be lacking information needed for this citation format:
No year of publication.

MLA Handbook (7^th Edition):

Compton, William Kirby. “Impact of configuration variations on small modular reactor core performance.” Web. 16 Feb 2019.

Note: this citation may be lacking information needed for this citation format:
No year of publication.

Vancouver:

Compton WK. Impact of configuration variations on small modular reactor core performance. [Internet] [Masters thesis]. Missouri University of Science and Technology; [cited 2019 Feb 16]. Available from: http://scholarsmine.mst.edu/masters_theses/7391.

Note: this citation may be lacking information needed for this citation format:
No year of publication.

Council of Science Editors:

Compton WK. Impact of configuration variations on small modular reactor core performance. [Masters Thesis]. Missouri University of Science and Technology; Available from: http://scholarsmine.mst.edu/masters_theses/7391

Note: this citation may be lacking information needed for this citation format:
No year of publication.

McMaster University

27. Tan, Andrew. Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code.

Degree: MASc, 2015, McMaster University

URL: http://hdl.handle.net/11375/18666

►

The goal of this thesis is to study the transient behaviour of the SLOWPOKE-2 reactor using Monte-Carlo simulations with the G4-STORK code. G4-STORK is a… (more)

Subjects/Keywords: Nuclear Physics; Reactor Physics; Nuclear Engineering; Simulations; Transient; Transient Simulations; GEANT4; G4-STORK; Computational Physics; SLOWPOKE-2

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Tan, A. (2015). Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/18666

Chicago Manual of Style (16^th Edition):

Tan, Andrew. “Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code.” 2015. Masters Thesis, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/18666.

MLA Handbook (7^th Edition):

Tan, Andrew. “Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code.” 2015. Web. 16 Feb 2019.

Vancouver:

Tan A. Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code. [Internet] [Masters thesis]. McMaster University; 2015. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/18666.

Council of Science Editors:

Tan A. Transient Simulations of the SLOWPOKE-2 Reactor Using the G4-STORK Code. [Masters Thesis]. McMaster University; 2015. Available from: http://hdl.handle.net/11375/18666

McMaster University

28. Russell, Liam F. Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit.

Degree: MASc, 2012, McMaster University

URL: http://hdl.handle.net/11375/12657

►

When the material or geometry of a reactor varies with time, the neutron flux will respond in the form of a reactor transient. These… (more)

▼

When the material or geometry of a reactor varies with time, the neutron flux will respond in the form of a reactor transient. These transients can occur during normal operations when control rods are moved or the reactor is refuelled (CANDU). During a reactor accident, the transient response is especially important because the reactor properties vary quickly with large amplitudes. Therefore, better understanding these conditions allows for improved identification, prevention and mitigation of reactor transients. However, current nuclear simulation codes are generally limited in their ability to model transient behaviour. The NStable code was created to model time-dependent neutron populations in multiplying mediums using the Geant4 Monte Carlo toolkit. The neutron population is allowed to evolve in time, but is periodically renormalized so that the total number of neutrons is constrained within a manageable range. This ensures that the simulation is viable even in highly sub- or supercritical environments. Since Geant4 was not intrinsically designed to track a neutron population over "long" time periods (up to 10 s), the population renormalization mechanisms needed to be created and integrated with Geant4. Additionally, nuclear reactor analysis functionality was added to calculate important quantities such as k_eff. The NStable code was validated using three established nuclear simulation codes: MCNP 5, DRAGON 3.06J, and TART 2005. The validation cases compared spatial distributions and criticality estimates for either homogeneous spheres (uranium-235 or a uranium-heavy water mixture) or the standard CANDU 6 lattice cell. For all three systems, the criticality estimates in NStable agreed with the appropriate validation code within 10 mk (TART for the spheres and DRAGON for the CANDU 6 lattice). Finally, the NStable code was also used to simulate a temperature transient in a UHW sphere where the temperature linear increased by 700 K over 50 ms. In response to the increasing temperature, k_eff decreased by 100 mk over the same period. In the future, transient modelling in NStable should be investigated further to reproduce actual experimental results, and to couple NStable with a thermohydraulics code to simulate a full transient response.

Master of Applied Science (MASc)

Advisors/Committee Members: Buijs, Adriaan, Jonkmans, Guy, Engineering Physics and Nuclear Engineering.

Subjects/Keywords: Monte Carlo; reactor physics; nuclear engineering; stochastic; time-dependent; Geant4; Engineering Physics; Nuclear; Nuclear Engineering; Engineering Physics

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Russell, L. F. (2012). Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/12657

Chicago Manual of Style (16^th Edition):

Russell, Liam F. “Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit.” 2012. Masters Thesis, McMaster University. Accessed February 16, 2019. http://hdl.handle.net/11375/12657.

MLA Handbook (7^th Edition):

Russell, Liam F. “Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit.” 2012. Web. 16 Feb 2019.

Vancouver:

Russell LF. Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit. [Internet] [Masters thesis]. McMaster University; 2012. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/11375/12657.

Council of Science Editors:

Russell LF. Simulation of Time-Dependent Neutron Populations for Reactor Physics Applications Using the Geant4 Monte Carlo Toolkit. [Masters Thesis]. McMaster University; 2012. Available from: http://hdl.handle.net/11375/12657

University of Ontario Institute of Technology

29. Colton, Ashlea V. K. Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients.

Degree: 2016, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/755

► To model the neutronic physics behavior of the core in CANDU pressure tube type heavy water reactors with natural-uranium fuel, two levels of calculations are… (more)

Subjects/Keywords: Reactor physics; Diffusion theory; Heavy water reactors; Natural uranium; Core reflector interface

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Colton, A. V. K. (2016). Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/755

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Colton, Ashlea V K. “Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients.” 2016. Thesis, University of Ontario Institute of Technology. Accessed February 16, 2019. http://hdl.handle.net/10155/755.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Colton, Ashlea V K. “Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients.” 2016. Web. 16 Feb 2019.

Vancouver:

Colton AVK. Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients. [Internet] [Thesis]. University of Ontario Institute of Technology; 2016. [cited 2019 Feb 16]. Available from: http://hdl.handle.net/10155/755.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Colton AVK. Improved 3D modelling of CANDU reactors using transport-fitted diffusion coefficients. [Thesis]. University of Ontario Institute of Technology; 2016. Available from: http://hdl.handle.net/10155/755

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

30. Rossi, Lubianka Ferrari Russo. Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear.

Degree: PhD, Tecnologia Nuclear - Reatores, 2014, University of São Paulo

URL: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-12022015-154545/ ;

►

Este trabalho apresenta um novo método para o cálculo dos coecientes de sensibilidade, através da união do metodo diferencial e da teoria da perturbação generalizada,… (more)

▼

Este trabalho apresenta um novo método para o cálculo dos coecientes de sensibilidade, através da união do metodo diferencial e da teoria da perturbação generalizada, que são os dois métodos tradicionalmente utilizados em física de reatores para a obtenção de tais grandezas. Esses dois métodos apresentam algumas deciências tornando os cálculos dos coeficientes de sensibilidade lentos ou computacionalmente exaustivos, mas unindo-os e possível eliminar as deciências apresentadas por ambos e obter uma nova equação para o coe- ciente de sensibilidade. O método proposto neste trabalho foi aplicado em um reator do tipo PWR , onde foi feita análise de sensibilidade da produção e da razão de conversão do 239Pu, para um ciclo de 120 dias de queima. O código utilizado para a análise de queima e análise de sensibilidade, o CINEW, foi desenvolvido durante este trabalho e os resultados obtidos foram comparados com os códigos amplamente utilizados em física de reatores, como o CINDER e o SERPENT. As conclusões obtidas foram que o novo método matemático para a obtenção dos coeficientes de sensibilidade e o CINEW, além de fornecer agilidade numérica também presentam eciência e segurança. Pois o novo método matemático para a obtenção dos coeficientes quando comparados com os métodos tradicionais utilizados para a análise de sensibilidade, mostram resultados satisfatórios, mesmo quando o método utiliza aproximações matemáticas que diferem do método proposto, e com a vantagem de não apresentar as deciências apresentadas pelos métodos diferencial e da teoria da perturbação generalizada. As análises de queima obtidas pelo CINEW foram comparadas com o CINDER, que mostraram uma diferença aceitável, apesar do CINDER apresentar alguns problemas computacionais que advém da época em que foi feito. A originalidade deste trabalho e a aplicação do método proposto em problemas que envolvem dependência temporal e a elaboração do primerio código nacional que faz análise de queima e análise de sensibilidade.

The main target of this study is to introduce a new method for calculating the coefficients of sensibility through the union of differential method and generalized perturbation theory, which are the two methods generally used in reactor physics to obtain such variables. These two methods, separated, have some issues turning the sensibility coefficients calculation slower or computationally exhaustive. However, putting them together, it\'s possible to repair these issues and build a new equation for the coecient of sensibility. The method introduced in this study was applied in a PWR reactor, where it was performed the sensibility analysis for the production and 239Pu conversion rate during 120 days (1 cycle) of burnup. The computational code used for both burnup and sensibility analysis, the CINEW, was developed in this study and all the results were compared with codes widely used in reactor physics, such as CINDER and SERPENT. The new mathematical method for calculating the sensibility coefficients and the code CINEW provide good numerical…

Advisors/Committee Members: Santos, Adimir dos.

Subjects/Keywords: analise de sensibilidade; coeficientes de sensibilidade; física de reatores; reactor physics; sensibility analysis; sensitivity coecients

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Rossi, L. F. R. (2014). Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear. (Doctoral Dissertation). University of São Paulo. Retrieved from http://www.teses.usp.br/teses/disponiveis/85/85133/tde-12022015-154545/ ;

Chicago Manual of Style (16^th Edition):

Rossi, Lubianka Ferrari Russo. “Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear.” 2014. Doctoral Dissertation, University of São Paulo. Accessed February 16, 2019. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-12022015-154545/ ;.

MLA Handbook (7^th Edition):

Rossi, Lubianka Ferrari Russo. “Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear.” 2014. Web. 16 Feb 2019.

Vancouver:

Rossi LFR. Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear. [Internet] [Doctoral dissertation]. University of São Paulo; 2014. [cited 2019 Feb 16]. Available from: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-12022015-154545/ ;.

Council of Science Editors:

Rossi LFR. Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear. [Doctoral Dissertation]. University of São Paulo; 2014. Available from: http://www.teses.usp.br/teses/disponiveis/85/85133/tde-12022015-154545/ ;

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