subject:(spent fuel)

University of Illinois – Urbana-Champaign

1. Lartonoix, David. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.

Degree: MS, 5183, 2012, University of Illinois – Urbana-Champaign

URL: http://hdl.handle.net/2142/31990

► While nuclear reactors in the United States have produced economy-driving power for several decades, they have also left behind a significant amount of spent nuclear… (more)

▼ While nuclear reactors in the United States have produced economy-driving power for several decades, they have also left behind a significant amount of spent nuclear fuel. The federal government, ultimately responsible for this spent fuel, has a history just as long in attempting to effectively bury, dispose, reprocess, or otherwise deal with this waste. As no attempts to date have been entirely successful, work continues to find an effective waste management solution. To aid planners, policymakers, and scientists in this endeavor, tools are currently needed to model the radioisotope inventory of all spent nuclear requiring disposal or other forms of remediation to accurately frame the scope of the issue. This project describes a simple method of calculating radioisotope concentrations in spent fuel by utilizing a unique approach to solving the diffusion equation eigenvalue problem. Herein, the dissolved boron concentration, essentially a chemical shim, is adjusted over an operational time period to maintain criticality in the reactor, compensating for fuel burnup, burnable poison burnout, and actinide and fission product buildup. It is shown, as an example, that the fractional reduction in boron concentration after a month of reactor operation is 2.3%. The normalized neutron flux in the example scenario is calculated and confirmed to be relatively flat radially and vertically. Similarly, the normalized thermal energy production rate is also shown to be relatively flat, as expected. Radionuclides of interest are tracked and isotopic concentrations are shown at various vertical heights within the core. Upon further refining, these concentrations can be taken to represent the radioisotope inventory of spent nuclear fuel under various burnup scenarios. Ultimately, characterizing the spent fuel requiring disposal will aid in developing an efficient waste management strategy. Even while several shortfalls are noted and described, tools such as this computer code can play a useful role in addressing the nation's nuclear waste disposal dilemma. Advisors/Committee Members: Singer, Clifford E. (advisor).

Subjects/Keywords: spent nuclear fuel; radioisotope inventory

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APA (6^th Edition):

Lartonoix, D. (2012). Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. (Thesis). University of Illinois – Urbana-Champaign. Retrieved from http://hdl.handle.net/2142/31990

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):

Lartonoix, David. “Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.” 2012. Thesis, University of Illinois – Urbana-Champaign. Accessed March 02, 2021. http://hdl.handle.net/2142/31990.

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

MLA Handbook (7^th Edition):

Lartonoix, David. “Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.” 2012. Web. 02 Mar 2021.

Vancouver:

Lartonoix D. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. [Internet] [Thesis]. University of Illinois – Urbana-Champaign; 2012. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/2142/31990.

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

Council of Science Editors:

Lartonoix D. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. [Thesis]. University of Illinois – Urbana-Champaign; 2012. Available from: http://hdl.handle.net/2142/31990

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

University of Cape Town

2. Khoza, Best. Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel.

Degree: MPhil, Electrical Engineering, 2019, University of Cape Town

URL: http://hdl.handle.net/11427/31697

► The continual increase in electricity dependence for the advancement of society has led to increased demand in electricity globally. This increased demand, among other things… (more)

▼ The continual increase in electricity dependence for the advancement of society has led to increased demand in electricity globally. This increased demand, among other things such as global warming interventions and energy security have encouraged the need to diversify electricity generation sources. Civilian use of nuclear power dates back to the 1950s. The United States of America and France are currently leading with the highest nuclear power generation in the world, generating 101 GWe and 63 GWe, respectively. Several countries such as China and the United Arab Emirates have committed to new nuclear build in order to increase their nuclear power generation capacities. Standing against the prospects of growth of the nuclear power industry are technical and nontechnical challenges. These include proliferation risk, safety, high capital costs and high-level waste management. Most spent nuclear fuel from power reactors is currently stored in the spent fuel pools on reactor sites, and some have been reprocessed. It is estimated that about 32% (370 000 tons of Heavy Metal) of the total spent fuel generated from power reactors have been reprocessed up to date. With most of the spent fuel pools filling up, alternative interim and long term disposal of spent nuclear fuel solutions have been under investigation from as early as the 1970s. South Africa has planned an interim dry storage facility for the spent nuclear fuel to be established at the existing Koeberg power station. The interim dry storage facility will make use of HI-STAR 100 multi-purpose casks to store spent nuclear fuel until the country decides on final disposal solution. There are many aspects that are critical to safe, efficient and cost-effective long term storage of spent nuclear fuel. Some of the physics and engineering aspects concerning dry storage facilities are briefly discussed. The aspects presented here are: radiation containment, spent fuel, sub-criticality, decay heat removal, site location aspects, response to seismic events, cask corrosion, transportation infrastructure, operability and monitoring. The study of the three existing dry cask storages from the USA, Hungary and Belgium gives an overview of the dry cask technology in use today. These presentations are based on publicly available reliable information. The proposed dry storage facility at Koeberg will be in the existing power station footprint using the HI-STAR 100 casks. The decision to have the proposed dry storage facility at Koeberg will minimise related licence applications and part of security installations as the site already has some security. The location of the facility in the power station’s footprint also allows for cost-effective and safe transportation of casks from the reactor building to the proposed facility. The modularity aspect of the dry cask storage facility at MV Paks in Hungary should also be employed at Koeberg to allow for more storage. This will cater for additional casks that may need to be stored if more nuclear power plants are procured in the future.… Advisors/Committee Members: Aschman, David (advisor).

Subjects/Keywords: spent fuel; dry cask storage; spent fuel pool

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APA (6^th Edition):

Khoza, B. (2019). Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel. (Thesis). University of Cape Town. Retrieved from http://hdl.handle.net/11427/31697

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):

Khoza, Best. “Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel.” 2019. Thesis, University of Cape Town. Accessed March 02, 2021. http://hdl.handle.net/11427/31697.

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

MLA Handbook (7^th Edition):

Khoza, Best. “Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel.” 2019. Web. 02 Mar 2021.

Vancouver:

Khoza B. Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel. [Internet] [Thesis]. University of Cape Town; 2019. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/11427/31697.

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

Council of Science Editors:

Khoza B. Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel. [Thesis]. University of Cape Town; 2019. Available from: http://hdl.handle.net/11427/31697

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

Georgia Tech

3. Conant, Andrew. Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod.

Degree: MS, Mechanical Engineering, 2015, Georgia Tech

URL: http://hdl.handle.net/1853/56227

► The purpose of this research is to examine the effects of systematic uncertainty of reactor operating parameters on isotope ratios in spent fuel rods, specifically… (more)

Subjects/Keywords: Uncertainty; Plutonium; Cesium; Reactor; Spent fuel

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APA (6^th Edition):

Conant, A. (2015). Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/56227

Chicago Manual of Style (16^th Edition):

Conant, Andrew. “Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod.” 2015. Masters Thesis, Georgia Tech. Accessed March 02, 2021. http://hdl.handle.net/1853/56227.

MLA Handbook (7^th Edition):

Conant, Andrew. “Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod.” 2015. Web. 02 Mar 2021.

Vancouver:

Conant A. Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod. [Internet] [Masters thesis]. Georgia Tech; 2015. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1853/56227.

Council of Science Editors:

Conant A. Sensitivity and uncertainty analysis of plutonium and cesium isotope ratios in BR3 core 4A/B fuel rod. [Masters Thesis]. Georgia Tech; 2015. Available from: http://hdl.handle.net/1853/56227

University of Illinois – Urbana-Champaign

4. Jarrah, Ibrahim. Risk of misloading spent nuclear fuel cask for light water reactors.

Degree: MS, Nuclear, Plasma, Radiolgc Engr, 2018, University of Illinois – Urbana-Champaign

URL: http://hdl.handle.net/2142/102866

► The spent fuel dry cask should remain subcritical under normal, abnormal, and accident conditions. The cask may become susceptible to criticality if it is misloaded… (more)

▼ The spent fuel dry cask should remain subcritical under normal, abnormal, and accident conditions. The cask may become susceptible to criticality if it is misloaded with assemblies that do not conform with the Certificate of Compliance (CoC). Assessment of probability of criticality for a misloaded cask that subsequently experiences an accident during the transportation process is also of interest. To avoid misloading, the cask loading process involves several verification steps to make sure that all of the loaded assemblies satisfy the CoC requirements. However, most of the loading and verification steps are carried out by humans with finite probabilities for errors, which need to be quantified. Quantification of the risk of having a misloaded cask may reduce the conservatism in the cask designs and eliminate unnecessary steps in the spent fuel handling and loading procedure. In the first part of this study, the probability of misloading a cask with at least one light water reactor, pressurized water reactor (PWR) and boiling water reactor (BWR), fuel assembly is quantified first using the event tree method. An event tree and associated fault trees are developed for the cask loading procedures. Probability distribution functions (PDFs) for all of the human errors are obtained using the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) human reliability analysis method. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software is used to quantify the event tree and to calculate the probability of misload. The probability of misloading a cask with at least one fuel assembly is first determined for PWR fuel and it was found to be 5.56E-06, which agrees well with that reported in the literature. The probability of misloading a cask with at least one fuel assembly for the BWR fuel is found to be 2.95E-05. The impact of the cask capacity on the probability of misload is quantified and discussed. The Fussell-Vesely (FV) importance measure is performed to determine the tasks that contribute the most to having a misloaded cask. The effects of the available time to perform a task and the stress level of the operator on the final probability of misload are studied. The available time and stress are found to have a significant impact on the final misload probability. Based on the neutronic calculations, the cask needs to be misloaded with more than one fuel assemblies in order to become susceptible to criticality. In the second part of this study, an event tree is built to predict the multiple misloads scenarios. Six multiple misloads scenarios are identified from the tree. The probabilities of the six scenarios and the total probability are calculated for casks for both reactor types. The probabilities calculated using this method are found to be 6.73E-07 and 7.55E-06 for PWR and BWR fuels, respectively. In addition, the probability of multiple misloads is calculated as a function of the cask capacity. Advisors/Committee Members: Uddin, Rizwan (advisor), Mohaghegh, Zahra (committee member).

Subjects/Keywords: Spent fuel cask; misload; PRA; Risk

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APA (6^th Edition):

Jarrah, I. (2018). Risk of misloading spent nuclear fuel cask for light water reactors. (Thesis). University of Illinois – Urbana-Champaign. Retrieved from http://hdl.handle.net/2142/102866

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):

Jarrah, Ibrahim. “Risk of misloading spent nuclear fuel cask for light water reactors.” 2018. Thesis, University of Illinois – Urbana-Champaign. Accessed March 02, 2021. http://hdl.handle.net/2142/102866.

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

MLA Handbook (7^th Edition):

Jarrah, Ibrahim. “Risk of misloading spent nuclear fuel cask for light water reactors.” 2018. Web. 02 Mar 2021.

Vancouver:

Jarrah I. Risk of misloading spent nuclear fuel cask for light water reactors. [Internet] [Thesis]. University of Illinois – Urbana-Champaign; 2018. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/2142/102866.

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

Council of Science Editors:

Jarrah I. Risk of misloading spent nuclear fuel cask for light water reactors. [Thesis]. University of Illinois – Urbana-Champaign; 2018. Available from: http://hdl.handle.net/2142/102866

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

Penn State University

5. Bratton, Ryan Nathaniel. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.

Degree: 2015, Penn State University

URL: https://submit-etda.libraries.psu.edu/catalog/27289

► The bias and uncertainty in fuel isotopic calculations for a well-defined radiochemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in… (more)

▼ The bias and uncertainty in fuel isotopic calculations for a well-defined radiochemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by comparing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investigation into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. Moreover, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layers is derived and applied to FRAPCON output data to quantify the RIP and CHS for these fuel rods. SCALE/Polaris is used to quantify fuel rod-specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel blankets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be… Advisors/Committee Members: Kostadin Nikolov Ivanov, Dissertation Advisor/Co-Advisor, Igor Jovanovic, Committee Chair/Co-Chair, Monique Yvonne Yaari, Committee Member, Maria Nikolova Avramova, Committee Member, Matthew A Jessee, Special Member, William A Wieselquist, Special Member.

Subjects/Keywords: SCALE; FRAPCON; rod internal pressure; fuel isotopics; spent nuclear fuel

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APA (6^th Edition):

Bratton, R. N. (2015). Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/27289

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):

Bratton, Ryan Nathaniel. “Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.” 2015. Thesis, Penn State University. Accessed March 02, 2021. https://submit-etda.libraries.psu.edu/catalog/27289.

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

MLA Handbook (7^th Edition):

Bratton, Ryan Nathaniel. “Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.” 2015. Web. 02 Mar 2021.

Vancouver:

Bratton RN. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. [Internet] [Thesis]. Penn State University; 2015. [cited 2021 Mar 02]. Available from: https://submit-etda.libraries.psu.edu/catalog/27289.

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

Council of Science Editors:

Bratton RN. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. [Thesis]. Penn State University; 2015. Available from: https://submit-etda.libraries.psu.edu/catalog/27289

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

Texas A&M University

6. Khudoleeva, Alexandra P. Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations.

Degree: MS, Nuclear Engineering, 2013, Texas A&M University

URL: http://hdl.handle.net/1969.1/158895

► The International Atomic Energy Agency (IAEA) needs to enhance its capabilities for safeguarding spent nuclear fuel (SNF) stored in dry cask storage facilities and for… (more)

▼ The International Atomic Energy Agency (IAEA) needs to enhance its capabilities for safeguarding spent nuclear fuel (SNF) stored in dry cask storage facilities and for maintaining persistent continuity of knowledge (CoK) about it. The current safeguards approach relies heavily upon containment and surveillance measures, where seals are placed inside and outside the dry cask. The disadvantage of this approach is that, if a seal is broken, no method currently exists to verify the dry cask content other than opening it and checking the internal seal and the SNF inside. This is a costly and difficult activity. Thus other measures need to be developed. This study focused on the development of a remote monitoring system (RMS) for dry cask storage which is capable of detecting neutron and gamma radiation emitted by the SNF and the signal thus generated can then be continually transmitted to the IAEA to maintain the CoK about the dry cask content. The remote option was chosen after reviewing the current IAEA needs. A computational approach was used to develop the proposed RMS. Monte Carlo N-Particle transport code (MCNP) was employed to develop a dry cask model with 32 SNF assemblies inside. The ORIGEN-ARP, fuel burn-up and depletion code, was used to generate a radiation source-term. A series of MCNP simulations were performed to investigate the neutron and gamma flux behavior inside the dry cask. The results of these simulations aided the design of the RMS and determination of the optimal location for its components. The RMS was placed inside the dry cask on the top of the multi-purpose canister (MPC). The final conceptual design of the RMS included two fission chambers (to detect neutrons) and one ionization chamber (to detect gamma radiation) enclosed in a polyethylene box with a thin cadmium plate inside, so the sequence of layers starting from the MPC lid was: polyethylene bottom layer, cadmium plate, chambers enclosed in polyethylene and polyethylene layer on top. Such configuration provided a suppression effect for thermal neutron flux coming from the bottom SNF assemblies and made system more sensitive to the opening of the dry cask lid and removal of SNF assemblies from the peripheral MPC cells. The proposed RMS design was tested through diversion analysis. The fission chamber unit design was successfully able to detect all the SNF diversion scenarios studied. The ionization chambers were able to detect only removal of SNF assemblies located just below it. However, the ionization chamber was found to be able to identify the opening of the dry cask lid through reduction in signal whenever the lid was opened. Therefore, the ionization chamber was kept in the RMS design to provide secondary confirmation for the detection of dry cask lid opening. Advisors/Committee Members: Chirayath, Sunil S. (advisor), Charlton, William S. (advisor), Folden III, Charles M. (committee member).

Subjects/Keywords: spent nuclear fuel; dry cask storage; safeguards; remote monitoring system

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APA (6^th Edition):

Khudoleeva, A. P. (2013). Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/158895

Chicago Manual of Style (16^th Edition):

Khudoleeva, Alexandra P. “Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations.” 2013. Masters Thesis, Texas A&M University. Accessed March 02, 2021. http://hdl.handle.net/1969.1/158895.

MLA Handbook (7^th Edition):

Khudoleeva, Alexandra P. “Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations.” 2013. Web. 02 Mar 2021.

Vancouver:

Khudoleeva AP. Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations. [Internet] [Masters thesis]. Texas A&M University; 2013. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1969.1/158895.

Council of Science Editors:

Khudoleeva AP. Conceptual Development of Remote Monitoring System for Pressurized Water Reactor Spent Fuel Dry Cask Storage Through Neutron and Gamma Transport Simulations. [Masters Thesis]. Texas A&M University; 2013. Available from: http://hdl.handle.net/1969.1/158895

Uppsala University

7. Lundkvist, Niklas. AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements.

Degree: Applied Nuclear Physics, 2010, Uppsala University

URL: http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011

►

This report is concerned with the question whether Accelerator mass spectrometry (AMS) is asuitable technique for measuring actinide inventory in spent nuclear fuel, and… (more)

▼

This report is concerned with the question whether Accelerator mass spectrometry (AMS) is asuitable technique for measuring actinide inventory in spent nuclear fuel, and if it is better thanpresent techniques for these measurements. AMS is a kind of Mass spectrometry (MS) and has alot of applications where radio carbon dating is one of the most common. AMS has been used formaking measurements on actinides before but mostly from traces in bioassay that could have beenin contact with weapon plutonium, and in bioassay near enrichment plants and reprocessingplants. It is shown in this report that AMS is more sensitive in low level measurements than thecurrent technique for spent nuclear fuel. ICP-MS is the current technique in use for inventorymeasurements on nuclear fuel at Swedish Nuclear Fuel and Waste Management Company (SKB).ICP-MS is also a kind of MS technique which is well-tried for inventory measurements on spentnuclear fuel. The difference in sensitivity ranges in levels of magnitude depending on whichisotope that is interesting for measurements. The lower detection limits for AMS is about 105-107atoms which makes it possible to use samples from nuclear fuel that is in the order of 10-10-10-16gto achieve the lower detection limit. The recommendation from this report is to make studies ifAMS also is an economical and efficiently suitable technique for future use on the actinideinventory in spent nuclear fuel.

Denna rapport handlar om huruvida Accelerator masspektrometri (AMS) är en lämplig teknik förmätning av aktinidinventeriet i använt kärnbränsle. Rapporten går också igenom om AMS är bättreän nuvarande tekniker för dessa mätningar. AMS är en typ av masspektrometri (MS) och har enmängd användningsområden, kol-14 metoden är en av de vanligaste. AMS har också ofta använtsför att göra mätningar på aktinidinnehåll i biomassa som kan ha varit i kontakt medvapenplutonium, och i närheten av anrikningsanläggningar och upparbetningsanläggningar. Detvisas i rapporten att AMS är en mer känslig metod än de nuvarande teknikerna som används förmätningar på aktinidinventariet i använt kärnbränsle. ICP-MS är den aktuella teknik som användsför mätningar på aktinidinventariet i använt kärnbränsle vid Svenska Kärnbränslehantering AB(SKB). ICP-MS är också en typ av MS teknik. MS är väl beprövad för mätningar av inventariet påanvänt kärnbränsle. Skillnaden i känslighet varierar i flera storleksordningar beroende på vilkenisotop som är intressant för mätningarna. Den lägre detektionsgränsen för AMS är cirka 105-107atomer, vilket gör det möjligt att använda prover från kärnbränsle som är i storleksordningen 10-10-10-16g för att uppnå den lägre detektionsgränsen. Rekommendationen från denna rapport är attgöra undersökningar om AMS också är ekonomiskt lönsam och tillräckligt effektiv teknik förframtida bruk inom mätningar av aktinidinventariet i använt kärnbränsle.

Subjects/Keywords: AMS; actinides; spent; nuclear; fuel

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APA (6^th Edition):

Lundkvist, N. (2010). AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements. (Thesis). Uppsala University. Retrieved from http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011

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):

Lundkvist, Niklas. “AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements.” 2010. Thesis, Uppsala University. Accessed March 02, 2021. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011.

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

MLA Handbook (7^th Edition):

Lundkvist, Niklas. “AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements.” 2010. Web. 02 Mar 2021.

Vancouver:

Lundkvist N. AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements. [Internet] [Thesis]. Uppsala University; 2010. [cited 2021 Mar 02]. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011.

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

Council of Science Editors:

Lundkvist N. AMS on the actinides in spent nuclear fuel : a study on a technique for inventory measurements. [Thesis]. Uppsala University; 2010. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-130011

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

University of Tennessee – Knoxville

8. Liu, Zhengzhi. METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY.

Degree: 2018, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_graddiss/5033

► It's important to the International Atomic Energy Agency (IAEA) to develop a nondestructive assay technique that may that be used to verify the presence of… (more)

▼ It's important to the International Atomic Energy Agency (IAEA) to develop a nondestructive assay technique that may that be used to verify the presence of the used nuclear fuel stored in a dry storage cask once continuity of knowledge has been lost. X-rays and neutrons are not good candidates for assay because they do not penetrate dry storage casks with high probability, and gammas and neutrons are also emitted by the used nuclear fuel. In contrast, cosmic ray muons are naturally occurring highly penetrating particles. Muons interact with matter via two major interaction mechanisms: ionization and radioactive process, and multiple Coulomb scattering leading to energy loss and trajectory deflection, respectively. For a monoenergetic muon beam crossing an object, the scattering angle follows a Gaussian distribution with a zero mean value and a variance that depends on the atomic number of the material object it traversed. Thus, the measured scattering angle may be used to reconstruct the geometrical and material information of the contents inside the dry storage cask.In traditional X-ray computed tomography, the projection information used to reconstruct the attenuation map of the imaged objects is the negative natural logarithm of the transmission rate of the X-rays, which is equal to the linear summation of the X-ray attenuation coefficients along the incident path. Similarly, the variance of the muon scattering angle is also the linear integral of the scattering density of the objects crossed by the muons. Thus, a muon CT image can be built by equating scattering density with attenuation coefficient. However, muon CT faces some unique challenges including: 1) long measurement times due to low cosmic muon flux, 2) insufficiently accurate muon path models, and 3) the inability to precisely measuring muon momentum.In this work, three different muon path models, two different projection methods, and two different reconstruction methods were investigated for use in muon CT of dry storage casks. The investigation was conducted in a validated Geant4 workspace, both in an ideal case and with relevant engineering restrictions considered. The results of these investigations and the expected benefits for fuel cask monitoring are reported herein.

Subjects/Keywords: Spent nuclear fuel dry storage cask; Muon tomography

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APA (6^th Edition):

Liu, Z. (2018). METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/5033

Chicago Manual of Style (16^th Edition):

Liu, Zhengzhi. “METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY.” 2018. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed March 02, 2021. https://trace.tennessee.edu/utk_graddiss/5033.

MLA Handbook (7^th Edition):

Liu, Zhengzhi. “METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY.” 2018. Web. 02 Mar 2021.

Vancouver:

Liu Z. METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2018. [cited 2021 Mar 02]. Available from: https://trace.tennessee.edu/utk_graddiss/5033.

Council of Science Editors:

Liu Z. METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2018. Available from: https://trace.tennessee.edu/utk_graddiss/5033

University of Tennessee – Knoxville

9. Iyengar, Anagha Srikanth. The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography.

Degree: 2019, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_graddiss/5343

► Safeguarding spent fuel in spent fuel pools, during transportation, and at dry cask storage sites has been a continuing priority for the International Atomic Energy… (more)

Subjects/Keywords: International Safeguards; Spent Fuel; Fast Neutrons; Imaging; Tomography

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APA (6^th Edition):

Iyengar, A. S. (2019). The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/5343

Chicago Manual of Style (16^th Edition):

Iyengar, Anagha Srikanth. “The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography.” 2019. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed March 02, 2021. https://trace.tennessee.edu/utk_graddiss/5343.

MLA Handbook (7^th Edition):

Iyengar, Anagha Srikanth. “The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography.” 2019. Web. 02 Mar 2021.

Vancouver:

Iyengar AS. The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2019. [cited 2021 Mar 02]. Available from: https://trace.tennessee.edu/utk_graddiss/5343.

Council of Science Editors:

Iyengar AS. The Design of an Imager to Safeguard Spent Fuel Using Passive Fast Neutron Emission Tomography. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2019. Available from: https://trace.tennessee.edu/utk_graddiss/5343

10. Wikman, Tom. Analysis of Loss of Cooling in Spent Nuclear Fuel Pools .

Degree: Chalmers tekniska högskola / Institutionen för fysik, 2019, Chalmers University of Technology

URL: http://hdl.handle.net/20.500.12380/300378

► This master thesis presents an improved estimate of the course of events in the loss of cooling of the spent fuel pools on Ringhals 3… (more)

Subjects/Keywords: Ringhals; loss of cooling; spent fuel pool; evaporation; Comsol Multiphysics

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APA (6^th Edition):

Wikman, T. (2019). Analysis of Loss of Cooling in Spent Nuclear Fuel Pools . (Thesis). Chalmers University of Technology. Retrieved from http://hdl.handle.net/20.500.12380/300378

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):

Wikman, Tom. “Analysis of Loss of Cooling in Spent Nuclear Fuel Pools .” 2019. Thesis, Chalmers University of Technology. Accessed March 02, 2021. http://hdl.handle.net/20.500.12380/300378.

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

MLA Handbook (7^th Edition):

Wikman, Tom. “Analysis of Loss of Cooling in Spent Nuclear Fuel Pools .” 2019. Web. 02 Mar 2021.

Vancouver:

Wikman T. Analysis of Loss of Cooling in Spent Nuclear Fuel Pools . [Internet] [Thesis]. Chalmers University of Technology; 2019. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/20.500.12380/300378.

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

Council of Science Editors:

Wikman T. Analysis of Loss of Cooling in Spent Nuclear Fuel Pools . [Thesis]. Chalmers University of Technology; 2019. Available from: http://hdl.handle.net/20.500.12380/300378

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

Virginia Tech

11. Roskoff, Nathan. Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation.

Degree: PhD, Nuclear Engineering, 2018, Virginia Tech

URL: http://hdl.handle.net/10919/84487

► Fuel burnup calculations provide material concentrations and intrinsic neutron and gamma source strengths as a function of irradiation and cooling time. Detailed, full-core 3D burnup… (more)

▼ Fuel burnup calculations provide material concentrations and intrinsic neutron and gamma source strengths as a function of irradiation and cooling time. Detailed, full-core 3D burnup calculations are critical for nuclear fuel management studies, including core design and spent fuel storage safety and safeguards analysis. For core design, specifically during refueling, full- core pin-wise, axially-dependent burnup distributions are necessary to determine assembly positioning to efficiently utilize fuel resources. In spent fuel storage criticality safety analysis, detailed burnup distributions enable best-estimate analysis which allows for more effective utilization of storage space. Additionally, detailed knowledge of neutron and gamma source distributions provide the ability to ensure nuclear material safeguards. The need for accurate and efficient burnup calculations has become more urgent for the simulation of advanced reactors and monitoring and safeguards of spent fuel pools. To this end, the Virginia Tech Transport Theory Group (VT3G) has been working on advanced computational tools for accurate modeling and simulation of nuclear systems in real-time. These tools are based on the Multi-stage Response-function Transport (MRT) methodology. For monitoring and safety evaluation of spent fuel pools and casks, the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system has been developed. This dissertation presents a novel methodology and algorithm for performing 3D fuel bur- nup calculations, referred to as bRAPID- Burnup with RAPID . bRAPID utilizes the existing RAPID code system for accurate calculation of 3D fission source distributions as the trans- port calculation tool to drive the 3D burnup calculation. bRAPID is capable of accurately and efficiently calculating assembly-wise axially-dependent fission source and burnup dis- tributions, and irradiated-fuel properties including material compositions, neutron source, gamma source, spontaneous fission source, and activities. bRAPID performs 3D burnup calculations in a fraction of the time required by state-of-the-art methodologies because it utilizes a pre-calculated database of response functions. The bRAPID database pre-calculation procedure, and its automation, is presented. The ex- isting RAPID code is then benchmarked against the MCNP and Serpent Monte Carlo codes for a spent fuel pool and the U.S. Naval Academy Subcritical Reactor facility. RAPID is shown to accurately calculate eigenvalue, subcritical multiplication, and 3D fission source dis- tributions. Finally, bRAPID is compared to traditional, state-of-the art Serpent Monte Carlo burnup calculations and its performance will be evaluated. It is important to note that the automated pre-calculation proceedure is required for evaluating the performance of bRAPID. Additionally, benchmarking of the RAPID code is necessary to understand RAPID's ability to solve problems with variable burnups distributions and to asses its accuracy. Advisors/Committee Members: Haghighat, Alireza (committeechair), Liu, Yang (committee member), Pierson, Mark Alan (committee member), Tafti, Danesh K. (committee member), Sjoden, Glenn Eric (committee member).

Subjects/Keywords: burnup; depletion; neutron transport; spent nuclear fuel; fission matrix; RAPID

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APA (6^th Edition):

Roskoff, N. (2018). Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation. (Doctoral Dissertation). Virginia Tech. Retrieved from http://hdl.handle.net/10919/84487

Chicago Manual of Style (16^th Edition):

Roskoff, Nathan. “Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation.” 2018. Doctoral Dissertation, Virginia Tech. Accessed March 02, 2021. http://hdl.handle.net/10919/84487.

MLA Handbook (7^th Edition):

Roskoff, Nathan. “Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation.” 2018. Web. 02 Mar 2021.

Vancouver:

Roskoff N. Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation. [Internet] [Doctoral dissertation]. Virginia Tech; 2018. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/10919/84487.

Council of Science Editors:

Roskoff N. Development of a Novel Fuel Burnup Methodology and Algorithm in RAPID and its Benchmarking and Automation. [Doctoral Dissertation]. Virginia Tech; 2018. Available from: http://hdl.handle.net/10919/84487

Vanderbilt University

12. Favret, Derek Joe. Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities.

Degree: MS, Physics, 2006, Vanderbilt University

URL: http://hdl.handle.net/1803/13360

► Since September 11, 2001, the safety and security of the U.S. nuclear reactor complex has become a topic of controversy. Due to the safety features… (more)

Subjects/Keywords: terrorist; spent; nuclear; fuel; storage

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APA (6^th Edition):

Favret, D. J. (2006). Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities. (Thesis). Vanderbilt University. Retrieved from http://hdl.handle.net/1803/13360

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):

Favret, Derek Joe. “Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities.” 2006. Thesis, Vanderbilt University. Accessed March 02, 2021. http://hdl.handle.net/1803/13360.

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

MLA Handbook (7^th Edition):

Favret, Derek Joe. “Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities.” 2006. Web. 02 Mar 2021.

Vancouver:

Favret DJ. Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities. [Internet] [Thesis]. Vanderbilt University; 2006. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1803/13360.

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

Council of Science Editors:

Favret DJ. Analysis on the Potential Implications of a Terrorist Attack at U.S. Spent Nuclear Fuel Storage Facilities. [Thesis]. Vanderbilt University; 2006. Available from: http://hdl.handle.net/1803/13360

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

Purdue University

13. Chatzidakis, Stylianos. Cosmic ray muons for spent nuclear fuel monitoring.

Degree: PhD, Nuclear Engineering, 2016, Purdue University

URL: https://docs.lib.purdue.edu/open_access_dissertations/631

► There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option.… (more)

▼ There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks from muon measurements. A combination of muon scattering and muon transmission imaging can improve resolution and thus a missing fuel assembly can be identified for vertical and horizontal dry casks. The apparent separation of the images reveals that the muon scattering and transmission can be used for discrimination between casks, satisfying the diversion criteria set by IAEA. Advisors/Committee Members: Lefteri H. Tsoukalas, Lefteri H. Tsoukalas, Chan K. Choi, Mary L. Comer, Mamoru Ishii, Won Sik Yang.

Subjects/Keywords: Applied sciences; Cosmic ray muons; Spent nuclear fuel; Nuclear Engineering

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APA (6^th Edition):

Chatzidakis, S. (2016). Cosmic ray muons for spent nuclear fuel monitoring. (Doctoral Dissertation). Purdue University. Retrieved from https://docs.lib.purdue.edu/open_access_dissertations/631

Chicago Manual of Style (16^th Edition):

Chatzidakis, Stylianos. “Cosmic ray muons for spent nuclear fuel monitoring.” 2016. Doctoral Dissertation, Purdue University. Accessed March 02, 2021. https://docs.lib.purdue.edu/open_access_dissertations/631.

MLA Handbook (7^th Edition):

Chatzidakis, Stylianos. “Cosmic ray muons for spent nuclear fuel monitoring.” 2016. Web. 02 Mar 2021.

Vancouver:

Chatzidakis S. Cosmic ray muons for spent nuclear fuel monitoring. [Internet] [Doctoral dissertation]. Purdue University; 2016. [cited 2021 Mar 02]. Available from: https://docs.lib.purdue.edu/open_access_dissertations/631.

Council of Science Editors:

Chatzidakis S. Cosmic ray muons for spent nuclear fuel monitoring. [Doctoral Dissertation]. Purdue University; 2016. Available from: https://docs.lib.purdue.edu/open_access_dissertations/631

14. Alam, Md Hasibul. ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process.

Degree: 2016, University of Nevada – Reno

URL: http://hdl.handle.net/11714/2117

► This thesis describes the development and design of simulation model in “ANSYS FLUENT” for Spent Nuclear Fuel dehydration process by “Forced Helium Dehydration” [13] method… (more)

Subjects/Keywords: Forced Helium Dehydration; Nuclear Waste Management; Spent Nuclear Fuel Drying

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APA (6^th Edition):

Alam, M. H. (2016). ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process. (Thesis). University of Nevada – Reno. Retrieved from http://hdl.handle.net/11714/2117

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):

Alam, Md Hasibul. “ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process.” 2016. Thesis, University of Nevada – Reno. Accessed March 02, 2021. http://hdl.handle.net/11714/2117.

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

MLA Handbook (7^th Edition):

Alam, Md Hasibul. “ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process.” 2016. Web. 02 Mar 2021.

Vancouver:

Alam MH. ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process. [Internet] [Thesis]. University of Nevada – Reno; 2016. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/11714/2117.

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

Council of Science Editors:

Alam MH. ANSYS/Fluent Simulation Model Development for Forced Helium Dehydration Process. [Thesis]. University of Nevada – Reno; 2016. Available from: http://hdl.handle.net/11714/2117

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

Texas A&M University

15. Yancey, Kristina. Nationwide Used Fuel Inventory Analysis.

Degree: MS, Nuclear Engineering, 2013, Texas A&M University

URL: http://hdl.handle.net/1969.1/151866

► The goal of this research was to develop a methodology to collect inventory estimates for the analysis and characterization of used fuel in the United… (more)

▼ The goal of this research was to develop a methodology to collect inventory estimates for the analysis and characterization of used fuel in the United States. To accomplish this, the Spent Fuel Database (SFD) was created. Data was collected for the database from publicly available information on the 103 operating reactors in January 2012. Using this data, plant models were developed using ORIGEN-ARP, a point-depletion tool. The output for each reactor model included current inventory estimates for used fuel taken out of the reactor 0, 1, 3, 5, 10, and 20 years ago. To determine the applicability of the database, a methodology was developed to analyze and compare the SFD with mass values produced using knowledge of past fuel assembly designs for general reactor classes. The methodology was centered around the idea of the “applicability range” (AR) of the database, which was defined as the degree to which a correct estimate can be made quantitatively. Pressurized Water Reactors (PWRs) were shown to have a much higher AR than Boiling Water Reactors (BWRs), and older assembly classes were shown to have a lower AR than newer classes. The fission products in the database were shown to consistently have a high AR. Berkelium and californium had low AR for all of the assembly classes, curium had low AR for BWR classes and mixed AR for PWR classes, and americium and some plutonium isotopes had low AR for BWR classes. An assessment of the inventory estimates considered the potential radiotoxicity and heat load from these masses. The radiotoxicity by ingestion decreased by about a factor of 10 from the newest used fuel to the oldest, and the radiotoxicity by inhalation decreased by a factor of 2. While one person could never eat or inhale a spent fuel assembly, radiotoxicity was used as a metric for the upper limit of possible harm. The heat load decreased by more than a factor of 100 over the same range of fuel assemblies. On a per assembly basis, the radiotoxicity and heat load showed similar trends, with newer PWR assemblies being the highest and BWR assemblies being the lowest in both categories. Considering these results, at a potential interim storage facility, priority should be given to the oldest BWR assemblies to reduce the radiotoxic risk and heating requirements. Also, reprocessing and transmuting is highly encouraged to reduce the radiotoxicity and heat of the waste entering storage. Finally, to continue improving the SFD, future work should seek to quantify the magnitude of the impact of variations in AR for curium and for BWR classes. Moreover, future work should incorporate the used fuel from all the shutdown reactors into the database. Even in its current form, though, the SFD is a useful reference tool. Advisors/Committee Members: Tsvetkov, Pavel V (advisor), Marianno, Craig (committee member), Kattawar, George (committee member).

Subjects/Keywords: nuclear; energy; spent fuel; used fuel; nuclide inventories; PWR; BWR; applicability range

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APA (6^th Edition):

Yancey, K. (2013). Nationwide Used Fuel Inventory Analysis. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/151866

Chicago Manual of Style (16^th Edition):

Yancey, Kristina. “Nationwide Used Fuel Inventory Analysis.” 2013. Masters Thesis, Texas A&M University. Accessed March 02, 2021. http://hdl.handle.net/1969.1/151866.

MLA Handbook (7^th Edition):

Yancey, Kristina. “Nationwide Used Fuel Inventory Analysis.” 2013. Web. 02 Mar 2021.

Vancouver:

Yancey K. Nationwide Used Fuel Inventory Analysis. [Internet] [Masters thesis]. Texas A&M University; 2013. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1969.1/151866.

Council of Science Editors:

Yancey K. Nationwide Used Fuel Inventory Analysis. [Masters Thesis]. Texas A&M University; 2013. Available from: http://hdl.handle.net/1969.1/151866

16. Gullekson, Brian J. Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases.

Degree: MS, Nuclear Engineering, 2013, Oregon State University

URL: http://hdl.handle.net/1957/37706

► Spent nuclear fuel (SNF) resultant from the generation of nuclear power is a chemically and radiologically diverse system which is advantageous to chemically process prior… (more)

▼ Spent nuclear fuel (SNF) resultant from the generation of nuclear power is a chemically and radiologically diverse system which is advantageous to chemically process prior to geologic disposal. Hydrometallurgy is the primary technology for chemical processing for light water reactor spent fuels, where spent fuel is dissolved in an acid for liquid based separations. The primary means for recovery of desired metals from the SNF solution is liquid-liquid extraction which is based on distribution (partitioning) of the metal ions between two immiscible phases based on thermodynamic favorability. One of the means of increasing this favorability is by designing extractant molecules to be either "harder" or "softer" bases, which will more preferentially extract harder or softer metal ions respectively. This technique is used in designing extractant molecules for targeted extraction as actinides are slightly softer than lanthanides, and precious metals produced in significant quantities from the fission process are especially soft metals. The work performed in this thesis involved the synthesis of a novel soft electron donor organic extractant molecule for testing of targeted soft metal extraction. The molecule synthesized was bis-dibutanethiolthiophosphinato-methane, or S6, a bidentate neutral extractant molecule with significant thiolysis for a softer electron environment. The synthesis technique was refined and the molecule composition and structure was confirmed by ¹H NMR, ³¹P NMR, and elemental analysis. Two metal groups, f-elements (actinides and lanthanides) and soft transition metals were tested for their extractability from nitric acid solutions into an S6 solution in n-dodecane. Aqueous solutions of nitric acid and n-dodecane as an organic diluent are typical liquid-liquid extraction conditions in spent nuclear fuel reprocessing. As extraction experiments were performed with radiotracers, for the soft metal extraction experiment, a mixture of the selected metals was neutron-activated in the OSU TRIGA reactor, as was europium to create a lanthanide radiotracer. Actinides and lanthanides were not seen to effectively extract into the organic or form a precipitate at all, making their partitioning with this extractant seemingly ineffective. Through gamma spectroscopy of an irradiated metal solution post-extraction, it is seen that only silver and palladium preferentially complex in the mixed metal samples into an insoluble organic ligand, dropping out of solution. This effect was more pronounced at higher acid concentrations, but silver was seen to slightly extract to the organic phase at all acid concentrations as well. This testing has shown that the S6 extractant can be used to recover silver and palladium from a mixed metal aqueous solution, such as one resultant from advanced spent nuclear fuel reprocessing operations. This result shows promise for future development of sulfur based organophosphate ligands for targeted extraction of precious metals from solutions. Advisors/Committee Members: Paulenova, Alena (advisor), Marcum, Wade (committee member).

Subjects/Keywords: Spent Nuclear Fuel Reprocessing; Spent reactor fuels

…1 Spent Nuclear Fuel Characterization �� ….. �� ….1 I.2 Liquid-Liquid Extraction… …Principles �� ….3 I.3 Spent Nuclear Fuel Reprocessing �� .6 I.4 Hard-Soft… …81 I. Problem Background and Experimental Approach I.1 Spent Nuclear Fuel… …precious metals from spent nuclear fuel reprocessing waste streams. Nuclear fission is the only… …MWd/MT at 150 days after discharge, the spent fuel contains 95.9% uranium, 1.0% minor…

10 more images…

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APA (6^th Edition):

Gullekson, B. J. (2013). Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/37706

Chicago Manual of Style (16^th Edition):

Gullekson, Brian J. “Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases.” 2013. Masters Thesis, Oregon State University. Accessed March 02, 2021. http://hdl.handle.net/1957/37706.

MLA Handbook (7^th Edition):

Gullekson, Brian J. “Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases.” 2013. Web. 02 Mar 2021.

Vancouver:

Gullekson BJ. Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases. [Internet] [Masters thesis]. Oregon State University; 2013. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1957/37706.

Council of Science Editors:

Gullekson BJ. Synthesis and testing of a novel soft donor organic extractant molecule for targeted soft metal extraction from aqueous phases. [Masters Thesis]. Oregon State University; 2013. Available from: http://hdl.handle.net/1957/37706

University of Tennessee – Knoxville

17. Petersen, Gordon Matthew. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.

Degree: 2016, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_graddiss/4156

► Commercial nuclear power plants produce long-lasting nuclear waste, primarily in the form of spent nuclear fuel (SNF) assemblies. Spent fuel pools (SFP) and canisters or… (more)

Subjects/Keywords: Optimization; Spent Nuclear Fuel; Allocation Strategy; Fuel Removal; Used Nuclear Fuel; Tractable Validation Model; Nuclear Engineering

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APA (6^th Edition):

Petersen, G. M. (2016). Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/4156

Chicago Manual of Style (16^th Edition):

Petersen, Gordon Matthew. “Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.” 2016. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed March 02, 2021. https://trace.tennessee.edu/utk_graddiss/4156.

MLA Handbook (7^th Edition):

Petersen, Gordon Matthew. “Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.” 2016. Web. 02 Mar 2021.

Vancouver:

Petersen GM. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2016. [cited 2021 Mar 02]. Available from: https://trace.tennessee.edu/utk_graddiss/4156.

Council of Science Editors:

Petersen GM. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2016. Available from: https://trace.tennessee.edu/utk_graddiss/4156

Brno University of Technology

18. Marcell, Jan. Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment.

Degree: 2019, Brno University of Technology

URL: http://hdl.handle.net/11012/12194

► The main object of this diploma thesis is solving problems concerning heat transfer in disposal cannister for spent nuclear fuel. In forepart possibilities of conceptual… (more)

Subjects/Keywords: úložný obalový soubor; hlubinné úložiště; vyhořelé jaderné palivo; přestup tepla; analytická metoda; numerická metoda; spent fuel cask; deep repository; spent fuel; heat transfer; analytical method; numerical method

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APA (6^th Edition):

Marcell, J. (2019). Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment. (Thesis). Brno University of Technology. Retrieved from http://hdl.handle.net/11012/12194

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):

Marcell, Jan. “Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment.” 2019. Thesis, Brno University of Technology. Accessed March 02, 2021. http://hdl.handle.net/11012/12194.

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

MLA Handbook (7^th Edition):

Marcell, Jan. “Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment.” 2019. Web. 02 Mar 2021.

Vancouver:

Marcell J. Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment. [Internet] [Thesis]. Brno University of Technology; 2019. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/11012/12194.

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

Council of Science Editors:

Marcell J. Přenos tepla v úložném obalovém souboru a jeho vliv na okolí: Heat transfer in the storage cask and its impact on the environment. [Thesis]. Brno University of Technology; 2019. Available from: http://hdl.handle.net/11012/12194

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

University of Guelph

19. AL-Kaabi, Zainab. Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor.

Degree: PhD, 2020, University of Guelph

URL: https://atrium.lib.uoguelph.ca/xmlui/handle/10214/17722

► One of the strategies of the circular economy is the utilization of organic byproducts by converting them to renewable bio-fuel and bio-materials. Adoption of this… (more)

▼ One of the strategies of the circular economy is the utilization of organic byproducts by converting them to renewable bio-fuel and bio-materials. Adoption of this concept is also an important focus of the current environmental bio-technology research. Black liquor, widely known as spent liquor, is a renewable byproduct from pulp and paper industry. This thesis aimed to develop an environment friendly and cost-effective method to produce bio-carbon from recycled paper neutral sulphite semi- chemical spent liquor (RPNSSCSL) as a potential source for bio-fuel and bio-material applications. The major barrier towards realizing this is understanding the related impurities and developing innovative procedures for converting them to products with acceptable profiles as per desired applications. Therefore, three objectives were recognized, 1) characterization of RPNSSCSL for the above potential applications. RPNSSCSL revealed unique chemical compositions profile and the major findings include the presence of chemical compounds from a non-plant origin, such as significant quantities of hexanedioic acid derivatives. The favorable bio-fuel characteristics of this spent liquor were: neutral pH of 7.1±0.5%, high volatile content of 66.19±0.3%, and higher heating value (HHV) of 15.71 MJ/kg. However, its ash content of 23.27±0.8% did not meet the criteria for bio-fuel applications although it was less than that of regular black liquors, 2) assessment of identified methods for production of cleaner bio-carbon from RPNSSCSL by a) a novel oxidation procedure using H2O2 b) hydrothermal carbonization (HTC) of this spent liquor. The produced bio-carbon ranged from 1.25 ± 0.05% to 1.48 ± 0.05% using the oxidation method, and 1.11 ± 0.03% to 1.45 ± 0.04% of ash from the HTC method. HHV ranged from 25.32 to 26.11 MJ/kg and from 28.68 to 29.34 MJ/kg respectively, from both methods, which are suitable for target applications, 3) evaluation for the production of bio-carbon from other industrial spent liquors (a) pine and spruce kraft spent liquors (PSKSL) (b) aspen and balsam poplar spent liquor (APKSL). Biocarbon was refined to reduce the ash content from 35.93±0.50% to 1.82±0.05% and from 37.04±0.5% to 0.71±0.07%, respectively for both spent liquors, with improved HHV of 24.26 MJ/kg and 26.89 MJ/kg for PSKSL and APKSL respectively. Advisors/Committee Members: Dutta, Animesh (advisor).

Subjects/Keywords: bio-carbon; bio-fuel; hydrogen peroxide; oxidation; recycled paper; spent liquor; bio-carbon; bio-fuel; hydrogen peroxide; oxidation; recycled paper; spent liquor

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APA (6^th Edition):

AL-Kaabi, Z. (2020). Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor. (Doctoral Dissertation). University of Guelph. Retrieved from https://atrium.lib.uoguelph.ca/xmlui/handle/10214/17722

Chicago Manual of Style (16^th Edition):

AL-Kaabi, Zainab. “Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor.” 2020. Doctoral Dissertation, University of Guelph. Accessed March 02, 2021. https://atrium.lib.uoguelph.ca/xmlui/handle/10214/17722.

MLA Handbook (7^th Edition):

AL-Kaabi, Zainab. “Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor.” 2020. Web. 02 Mar 2021.

Vancouver:

AL-Kaabi Z. Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor. [Internet] [Doctoral dissertation]. University of Guelph; 2020. [cited 2021 Mar 02]. Available from: https://atrium.lib.uoguelph.ca/xmlui/handle/10214/17722.

Council of Science Editors:

AL-Kaabi Z. Bio-carbon Production from Recycled Paper Neutral Sulphite Semi-chemical Spent Liquor. [Doctoral Dissertation]. University of Guelph; 2020. Available from: https://atrium.lib.uoguelph.ca/xmlui/handle/10214/17722

University of California – Berkeley

20. Mozin, Vladimir. Delayed Gamma-Ray Assay for Nuclear Safeguards.

Degree: Nuclear Engineering, 2011, University of California – Berkeley

URL: http://www.escholarship.org/uc/item/5mm1v58k

► This dissertation addresses the need for new non-destructive assay instruments capable of quantifying the fissile isotopic composition of spent nuclear fuel and of independently verifying… (more)

▼ This dissertation addresses the need for new non-destructive assay instruments capable of quantifying the fissile isotopic composition of spent nuclear fuel and of independently verifying the declared amounts of special nuclear materials at various stages of the nuclear fuel cycle. High-energy delayed gamma-ray spectroscopy can provide the ability to directly assay fissile and fertile isotopes in the highly radioactive environment of the spent fuel assemblies and to achieve the safeguards goal of measuring nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, and final disposal and repository sites. The delayed gamma-ray assay concept is investigated within this context with the objective of assessing whether the delayed gamma-ray assay instrument can provide sufficient sensitivity, isotope specificity and accuracy as required in nuclear material safeguards applications. Preliminary system design analysis indicates that the delayed gamma-ray response is affected by multiple parameters: type and intensity of the interrogating source, the configuration of the interrogation setup, the time pattern of the interrogation, and the resolution and count rate limit of the gamma-ray detection system.In order to handle the variety of factors associated with the delayed gamma-ray assay of spent nuclear fuel, a high-fidelity response modeling technique is introduced. The new algorithm seamlessly combines transport calculations with analytical decay/depletion, and discrete gamma-ray source reconstruction codes. Its performance was benchmarked in the dedicated experimental campaign involving accelerator-driven photo-neutron sources and samples containing fissile and fertile isotopes.Analytical estimations of the intensity of the delayed gamma-ray response and the passive background rate are utilized to develop a concept of the non-destructive instrument for the assay of spent nuclear fuel. The modeling technique is then applied to more detailed parametric study. These simulations included extensive spent fuel inventories, and accounted for realistic assay configurations and instrumentation. The results of this preliminary analysis indicate that the delayed gamma-ray assay of spent nuclear fuel assemblies can be performed with available neutron generator and detection technology.The sensitivity of the delayed gamma-ray spectra to the actinide content of the spent nuclear fuel is investigated. The simplest analysis of the delayed gamma-ray response is based on the analysis of integrated count rates and peak ratios. More powerful analytical and numerical methods are likely needed for determining the relative concentrations of fissile and fertile isotopes in samples with complex compositions.

Subjects/Keywords: Nuclear engineering; delayed gamma-ray assay; non-destructive assay; nuclear safeguards; spent nuclear fuel

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APA (6^th Edition):

Mozin, V. (2011). Delayed Gamma-Ray Assay for Nuclear Safeguards. (Thesis). University of California – Berkeley. Retrieved from http://www.escholarship.org/uc/item/5mm1v58k

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):

Mozin, Vladimir. “Delayed Gamma-Ray Assay for Nuclear Safeguards.” 2011. Thesis, University of California – Berkeley. Accessed March 02, 2021. http://www.escholarship.org/uc/item/5mm1v58k.

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

MLA Handbook (7^th Edition):

Mozin, Vladimir. “Delayed Gamma-Ray Assay for Nuclear Safeguards.” 2011. Web. 02 Mar 2021.

Vancouver:

Mozin V. Delayed Gamma-Ray Assay for Nuclear Safeguards. [Internet] [Thesis]. University of California – Berkeley; 2011. [cited 2021 Mar 02]. Available from: http://www.escholarship.org/uc/item/5mm1v58k.

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

Council of Science Editors:

Mozin V. Delayed Gamma-Ray Assay for Nuclear Safeguards. [Thesis]. University of California – Berkeley; 2011. Available from: http://www.escholarship.org/uc/item/5mm1v58k

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

Texas A&M University

21. Goodsell, Alison. Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications.

Degree: MS, Nuclear Engineering, 2012, Texas A&M University

URL: http://hdl.handle.net/1969.1/ETD-TAMU-2012-08-11627

► The ability to quickly and accurately quantify the plutonium (Pu) content in pressurized water reactor (PWR) spent nuclear fuel (SNF) is critical for nuclear forensics… (more)

▼ The ability to quickly and accurately quantify the plutonium (Pu) content in pressurized water reactor (PWR) spent nuclear fuel (SNF) is critical for nuclear forensics purposes. One non-destructive assay (NDA) technique being investigated to detect bulk Pu in SNF is measuring the self-induced x-ray fluorescence (XRF). Previous XRF measurements of Three Mile Island (TMI) PWR SNF taken in July 2008 and January 2009 at Oak Ridge National Laboratory (ORNL) successfully illustrated the ability to detect the 103.7 keV x ray from Pu using a planar high-purity germanium (HPGe) detector. This allows for a direct measurement of Pu in SNF. Additional gamma ray and XRF measurements were performed on TMI SNF at ORNL in October 2011 to measure the signal-to-noise ratio for the 103.7 keV peak. Previous work had shown that the Pu/U peak ratio was directly proportional to the Pu/U content and increased linearly with burnup. However, the underlying Compton background significantly reduced the signal-to-noise ratio for the x-ray peaks of interest thereby requiring a prolonged count time. Comprehensive SNF simulations by Stafford et al showed the contributions to the Compton continuum were due to high-energy gamma rays scattering in the fuel, shipping tube, cladding, collimator and detector1. The background radiation was primarily due to the incoherent scattering of the 137Cs 661.7 keV gamma. In this work methods to reduce the Compton background and thereby increase the signal-to-noise ratio were investigated. To reduce the debilitating effects of the Compton background, a crystal x-ray spectrometer system was designed. This wavelength-dispersive spectroscopy technique isolated the Pu and U x rays according to Bragg's law by x-ray diffraction through a crystal structure. The higher energy background radiation was blocked from reaching the detector using a customized collimator and shielding system. A flat quartz-crystal x-ray spectrometer system was designed specifically to fit the constraints and requirements of detecting XRF from SNF. Simulations were performed to design and optimize the collimator design and to quantify the improved signal-to-noise ratio of the Pu and U x-ray peaks. The proposed crystal spectrometer system successfully diffracted the photon energies of interest while blocking the high-energy radiation from reaching the detector and contributing to background counts. The spectrometer system provided a higher signal-to-noise ratio and lower percent error for the XRF peaks of interest from Pu and U. Using the flat quartz-crystal x-ray spectrometer and customized collimation system, the Monte Carlo N-Particle (MCNP) simulations showed the 103.7 keV Pu x-ray peak signal-to-noise ratio improved by a factor of 13 and decreased the percent error by a factor of 3.3. Advisors/Committee Members: Charlton, William S. (advisor), Poston, John W. (committee member), Vedlitz, Arnold (committee member).

Subjects/Keywords: x-ray fluorescence; Pu; U; nuclear forensics; spent nuclear fuel; crystal x-ray spectrometer

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APA (6^th Edition):

Goodsell, A. (2012). Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2012-08-11627

Chicago Manual of Style (16^th Edition):

Goodsell, Alison. “Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications.” 2012. Masters Thesis, Texas A&M University. Accessed March 02, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2012-08-11627.

MLA Handbook (7^th Edition):

Goodsell, Alison. “Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications.” 2012. Web. 02 Mar 2021.

Vancouver:

Goodsell A. Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2012-08-11627.

Council of Science Editors:

Goodsell A. Flat Quartz-Crystal X-ray Spectrometer for Nuclear Forensics Applications. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2012-08-11627

Texas A&M University

22. Gerhart, Jeremy Jens. Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel.

Degree: MS, Nuclear Engineering, 2015, Texas A&M University

URL: http://hdl.handle.net/1969.1/156269

► There is a need for a technique that is able to quickly and accurately quantify the amount of ^239Pu in spent nuclear fuel. With the… (more)

Subjects/Keywords: nuclear; resonance; fluorescence; nrf; cog; plutonium; mega-ray; spent fuel; pressurized water reactor; pwr

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APA (6^th Edition):

Gerhart, J. J. (2015). Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/156269

Chicago Manual of Style (16^th Edition):

Gerhart, Jeremy Jens. “Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel.” 2015. Masters Thesis, Texas A&M University. Accessed March 02, 2021. http://hdl.handle.net/1969.1/156269.

MLA Handbook (7^th Edition):

Gerhart, Jeremy Jens. “Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel.” 2015. Web. 02 Mar 2021.

Vancouver:

Gerhart JJ. Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel. [Internet] [Masters thesis]. Texas A&M University; 2015. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1969.1/156269.

Council of Science Editors:

Gerhart JJ. Simulation of Nuclear Resonance Fluorescence for the Quantification of Plutonium-239 in Nuclear Fuel. [Masters Thesis]. Texas A&M University; 2015. Available from: http://hdl.handle.net/1969.1/156269

Texas A&M University

23. Eigenbrodt, Julia M. Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures.

Degree: PhD, Nuclear Engineering, 2016, Texas A&M University

URL: http://hdl.handle.net/1969.1/156845

► The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful… (more)

▼ The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improve the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements. PNAR was used for measurements of SNF in Japan as well as fresh fuel pins at Los Alamos National Laboratory (LANL). The measured PNAR signal was shown to trend well with neutron multiplication and fissile content of the SNF. The PNAR measurements showed a 73% change in signal for a fuel burnup range of 7.1 to 19.2 GWd/MTHM of spent mixed-oxide (MOX) fuel and a 40% change in signal over a range of initial ^235U enrichment from 0.2% to 3.2% in UO2 fuel. Photon measurements were performed on a wide range of SNF pins to determine which photon signatures are visible in different sets of fuels. These signatures were then investigated and tested using a sensitivity analysis to determine the spent fuel parameters to which each signal is most sensitive. These photon signatures can be used to determine SNF parameters that can support PNAR determination of SNF fissile content. Based on the results from these measurements, we have concluded that spectral photon measurements can determine operating parameters to improve the implementation of PNAR. We also concluded that PNAR can accurately measure multiplication and fissile content in SNF with standard deviations of 1% and 4%, respectively. The PNAR and photon measurements can be used together as a powerful tool to support the IAEA safeguards technical objective. Advisors/Committee Members: Charlton, William S (advisor), Chirayath, Sunil (committee member), Tsvetkov, Pavel (committee member), Bettati, Riccardo (committee member).

Subjects/Keywords: PNAR; NDA; safeguards; nuclear engineering; spent nuclear fuel; photon measurements; neutron measurements

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APA (6^th Edition):

Eigenbrodt, J. M. (2016). Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/156845

Chicago Manual of Style (16^th Edition):

Eigenbrodt, Julia M. “Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures.” 2016. Doctoral Dissertation, Texas A&M University. Accessed March 02, 2021. http://hdl.handle.net/1969.1/156845.

MLA Handbook (7^th Edition):

Eigenbrodt, Julia M. “Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures.” 2016. Web. 02 Mar 2021.

Vancouver:

Eigenbrodt JM. Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures. [Internet] [Doctoral dissertation]. Texas A&M University; 2016. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/1969.1/156845.

Council of Science Editors:

Eigenbrodt JM. Spent Fuel Measurements: Passive Neutron Albedo Reactivity (PNAR) and Photon Signatures. [Doctoral Dissertation]. Texas A&M University; 2016. Available from: http://hdl.handle.net/1969.1/156845

Penn State University

24. Bender, Sarah Elizabeth. Study of Compton Suppression For Use in Spent Nuclear Fuel Assay.

Degree: 2014, Penn State University

URL: https://submit-etda.libraries.psu.edu/catalog/21088

► Nuclear material accountancy is of continuous concern for the regulatory, safeguards, and verification communities. In particular, spent nuclear fuel reprocessing facilities pose one of the… (more)

▼ Nuclear material accountancy is of continuous concern for the regulatory, safeguards, and verification communities. In particular, spent nuclear fuel reprocessing facilities pose one of the most difficult accountancy challenges: continuously monitoring large volumes of highly radioactive, fluid sample streams. Current accountancy methods for nuclear fuel reprocessing facilities are resource intensive and time-consuming. The adaptation of passive gamma-ray detection coupled with multivariate analysis techniques could reduce the man-power requirements and processing time of samples. However, in measured gamma-ray spectra from spent nuclear fuel, the Compton continuum from the dominant 661.7 keV 137Cs fission product gamma-ray photon peak obscures lower energy lines. The application of Compton suppression to gamma-ray measurements of spent fuel will reduce the high continuum from nuclides like 137Cs and may allow other less intense, lower energy peaks to be detected, potentially improving the accuracy of multivariate analysis algorithms. There has been previous investigation into the use of room temperature detectors for gamma-ray spectroscopy of spent nuclear fuel. Interest has also been expressed in the application of Compton suppression to room temperature detectors for similar applications. Therefore, the focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. This allowed for the quantification of the number of additionally resolvable peaks through the application of Compton suppression and enabled the analysis of the effect of Compton suppressed gamma-ray detection in the presence high radiation field. A novel Compton suppressed detector model for the simulation of spent fuel measurements was developed. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any… Advisors/Committee Members: Kenan Unlu, Dissertation Advisor/Co-Advisor, Jack Brenizer Jr., Committee Chair/Co-Chair, Igor Jovanovic, Committee Member, Joshua Alexander Robinson, Committee Member, Christopher Orton, Special Member, Jon Schwantes, Special Member.

Subjects/Keywords: Spent fuel; Compton suppression; gamma spectroscopy; lanthanum bromide; burnup; Monte Carlo; Geant4

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APA (6^th Edition):

Bender, S. E. (2014). Study of Compton Suppression For Use in Spent Nuclear Fuel Assay. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/21088

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):

Bender, Sarah Elizabeth. “Study of Compton Suppression For Use in Spent Nuclear Fuel Assay.” 2014. Thesis, Penn State University. Accessed March 02, 2021. https://submit-etda.libraries.psu.edu/catalog/21088.

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

MLA Handbook (7^th Edition):

Bender, Sarah Elizabeth. “Study of Compton Suppression For Use in Spent Nuclear Fuel Assay.” 2014. Web. 02 Mar 2021.

Vancouver:

Bender SE. Study of Compton Suppression For Use in Spent Nuclear Fuel Assay. [Internet] [Thesis]. Penn State University; 2014. [cited 2021 Mar 02]. Available from: https://submit-etda.libraries.psu.edu/catalog/21088.

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

Council of Science Editors:

Bender SE. Study of Compton Suppression For Use in Spent Nuclear Fuel Assay. [Thesis]. Penn State University; 2014. Available from: https://submit-etda.libraries.psu.edu/catalog/21088

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

North Carolina State University

25. Bobolea, Ruxandra. A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel.

Degree: MS, Nuclear Engineering, 2009, North Carolina State University

URL: http://www.lib.ncsu.edu/resolver/1840.16/738

► Several methods of reprocessing are currently available to separate recyclable materials from spent nuclear fuel. Electrochemical processing, also known as pyroprocessing, represents a non-aqueous method… (more)

▼ Several methods of reprocessing are currently available to separate recyclable materials from spent nuclear fuel. Electrochemical processing, also known as pyroprocessing, represents a non-aqueous method of reprocessing that uses high temperature molten-salt based electrochemical technology. This method provides several advantages over conventional aqueous processing with respect to proliferation resistance. With electrochemical processing there is no pure plutonium separation and the presence of large decay heat and high radiation barriers dissuades diversion attempts. As the current electrochemical processing relies on a batch operation, the total throughput of the system is inherently limited and nuclear materials accounting is difficult due to the nonhomogeneous nature of the process. This results in much larger uncertainties in the total amount of material processed compared to the aqueous UREX+ or PUREX processes. Continuous electrochemical processing was considered as a way to address these concerns. The objective of this research was to investigate the feasibility of a continuous electrochemical processing operation to achieve the desired separation performance by using computer based simulation. The conceptual design of the continuous electrochemical processing includes two separate stages in a molten salt medium. First, a pure uranium deposit is collected at a solid cathode during the uranium extraction stage. When the amount of plutonium in electrorefiner becomes comparable or higher than the amount of uranium in the electrorefiner, a liquid cathode is employed to extract both uranium and plutonium in the second stage. In this approach, molten salt, as the material carrier, flows through the electrorefiner while chopped spent fuel is continuously fed into the system. Simulations of electrochemical reactions at the electrode surfaces were based on the kinetic modeling capability of a time-dependent code, REFIN. Based on a screening study performed for the most significant process parameters over a broad range of values, a functional combination of initial uranium and plutonium concentrations at the anode and in the molten salt was determined for continuous operation. This dictated the use of a higher concentration of uranium than plutonium at the anode and a lower concentration of uranium than plutonium in the molten salt. Furthermore, using design of experiment technique for computers, a refinement of initial concentrations was performed to maximize the total throughput and minimize the operational time. The flow velocity profiles and chemical concentration distributions of elements in molten salt have been determined through three dimensional Computational Fluid Dynamics simulations using ANSYS CFX. This approach resulted in the need to evaluate the diffusion layer thickness at the cathode Ã¢â‚¬â€œ molten salt interface, an important parameter for the electrochemical process. Computer based simulations of the continuous electrochemical processing concept presented in this study have provided an indication… Advisors/Committee Members: David N. McNelis, Committee Member (advisor), Man-Sung Yim, Committee Chair (advisor), Jeff Thompson, Committee Member (advisor).

Subjects/Keywords: diffusion layer thickness; continuous reprocessing; electrochemical processing; spent nuclear fuel; pyroprocessing; electrorefiner

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APA (6^th Edition):

Bobolea, R. (2009). A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel. (Thesis). North Carolina State University. Retrieved from http://www.lib.ncsu.edu/resolver/1840.16/738

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):

Bobolea, Ruxandra. “A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel.” 2009. Thesis, North Carolina State University. Accessed March 02, 2021. http://www.lib.ncsu.edu/resolver/1840.16/738.

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

MLA Handbook (7^th Edition):

Bobolea, Ruxandra. “A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel.” 2009. Web. 02 Mar 2021.

Vancouver:

Bobolea R. A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel. [Internet] [Thesis]. North Carolina State University; 2009. [cited 2021 Mar 02]. Available from: http://www.lib.ncsu.edu/resolver/1840.16/738.

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

Council of Science Editors:

Bobolea R. A Study of Continuous Electrochemical Processing Operation Feasibility for Spent Nuclear Fuel. [Thesis]. North Carolina State University; 2009. Available from: http://www.lib.ncsu.edu/resolver/1840.16/738

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

University of Manchester

26. Foster, Lynn. Understanding the microbial ecology of highly radioactive nuclear storage facilities.

Degree: 2019, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:318101

► The First Generation Magnox Storage Pond (FGMSP) is situated on the Sellafield Ltd site in the UK, and is an extremely inhospitable environment comprising of… (more)

▼ The First Generation Magnox Storage Pond (FGMSP) is situated on the Sellafield Ltd site in the UK, and is an extremely inhospitable environment comprising of significant levels of radioactivity coupled with a high pH of pH 11.4. Despite such extreme conditions, microorganisms are known to colonise the pond and can form dense microbial blooms in the summer months. The blooms can restrict the visibility within the pond which hinders plant operations. The FGMSP is currently undergoing decommissioning and waste retrieval operations, as a priority on site, therefore any plant downtime increases both the cost and timeframe for decommissioning. Here we describe the microbial community that colonises the FGSMP, including during two bloom periods. In addition efforts to determine the adaptive mechanisms that key microorganisms use to colonise the pond and their interactions with Sr are described. Over the course of the sampling period Proteobacteria were the dominant phylum in the pond, with variations seen at the lower taxonomic levels. In addition, a single cyanobacterium, affiliated with a Pseudanabaena species, was identified as the dominant photosynthetic microorganism from samples taken from two bloom periods, comprising up to 30% of the phylotypes detected. While the FGMSP was dominated by prokaryotes, a hydraulically linked auxiliary pond was more abundant in eukaryotic organisms. Comparisons between the two pond communities suggested that the auxiliary pond was not seeding the FGMSP, as the elevated pH and radiation levels inhibited such colonisation. Data supplied by Sellafield Ltd. showed that the onset of the bloom periods coincided with increases in the residence time of the purge water, used to maintain the elevated pH of the pond. Once the residence time of the purge water was reduced the visibility was restored in the pond, indicating that this was an effective means of removing the bloom-forming microorganisms. Laboratory-cultures of Pseudanabaena catenata were used to investigate the adaptive responses to ionizing radiation. The culture was found to consist of 9 other operational taxonomic units, 5 of which were affiliated with genera identified in the FGMSP. Detailed investigations indicated that X-irradiation treatment (95Gy) had no significant impact on the growth rate of the culture, however there was an increase in polysaccharide production and a reduction in protein and chlorophyll-a production. Increases in polysaccharides could be of importance in the FGMSP as this could influence the fate of radionuclides present in the water. Sr was used to determine whether P. catenata could influence the fate of radionuclides. P. catenata cells could be seen to accumulate Sr associated with polyphosphate bodies, whilst SrPO4 and calcium containing SrCO3 minerals were formed. The colonisation of FGMSP by organisms closely related to those studied here, including the cyanobacterium Pseudanabaena catenata requires careful consideration. The results presented here suggest that elevated levels of polysaccharides could… Advisors/Committee Members: MORRIS, KATHERINE K, PITTMAN, JON J, Lloyd, Jonathan, Morris, Katherine, Pittman, Jon.

Subjects/Keywords: First Generation Magnox Storage Pond; Peudanabaena; Radioactivity; Spent nuclear fuel ponds; Cyanobacteria

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

APA (6^th Edition):

Foster, L. (2019). Understanding the microbial ecology of highly radioactive nuclear storage facilities. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:318101

Chicago Manual of Style (16^th Edition):

Foster, Lynn. “Understanding the microbial ecology of highly radioactive nuclear storage facilities.” 2019. Doctoral Dissertation, University of Manchester. Accessed March 02, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:318101.

MLA Handbook (7^th Edition):

Foster, Lynn. “Understanding the microbial ecology of highly radioactive nuclear storage facilities.” 2019. Web. 02 Mar 2021.

Vancouver:

Foster L. Understanding the microbial ecology of highly radioactive nuclear storage facilities. [Internet] [Doctoral dissertation]. University of Manchester; 2019. [cited 2021 Mar 02]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:318101.

Council of Science Editors:

Foster L. Understanding the microbial ecology of highly radioactive nuclear storage facilities. [Doctoral Dissertation]. University of Manchester; 2019. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:318101

University of Colorado

27. Jing, Yuxiang. Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants.

Degree: PhD, 2018, University of Colorado

URL: https://scholar.colorado.edu/cven_gradetds/360

► To ensure safe operation of nuclear power plants (NPPs) during their service life and enhance the performance of spent nuclear fuel (SNF) storage systems, comprehensive… (more)

▼ To ensure safe operation of nuclear power plants (NPPs) during their service life and enhance the performance of spent nuclear fuel (SNF) storage systems, comprehensive investigation on the behavior of concrete and their components under the long-term nuclear radiation is needed. A theoretical model was developed first to predict the deterioration of concrete under neutron radiation, taking into account both of the effects of neutron radiation and the radiation-induced heating on the mechanical property and volume change of concrete. It was shown that the volume change of concrete is dominated by the expanding characteristic of aggregates. Since neutron radiation can deteriorate mechanical properties of the concrete materials, it’s critical to obtain the accurate neutron radiation levels in concrete structures during their service live. Neutron diffusion equations and heat conduction equation were used for prediction of neutron radiation and thermal field in concrete, respectively. The effects of potential variations of transport properties due to neutron radiation and elevated temperature on neutron diffusion in concrete were estimated. A simplified example of a typical concrete biological shielding wall was analyzed up to 80 years, and the results were discussed. The results show that neutron radiation and elevated temperature can result in considerable increases of neutron flux and fluence in the concrete. In order to understand the current state of knowledge about nuclear irradiated concrete, a collection of articles on neutron and gamma-ray radiation damage to concrete and/or its components was acquired. Information on testing conditions and concrete performance was extracted from the collected literature, and a database was developed. Data analysis of the effect of neutrons levels, water-cement ratio, aggregate fraction, and temperature on various properties of cementitious materials subjected to neutrons irradiation was conducted, and the results were presented. In order to monitor the long-term deterioration process of concrete used in NPPs, the self-sensing capability of carbon fiber reinforced cementitious composites under mechanical loading and elevated temperature was experimentally studied, and the results were described. It has potential to become a sensor and can be used to monitor the long-term variation of strain in concrete of NPPs structures or SNF storage systems. Advisors/Committee Members: Yunping Xi, Franck Vernerey, Jeong-Hoon Song, Mija H. Hubler, Ross B. Corotis.

Subjects/Keywords: spent nuclear fuel; neutron radiation; concrete; nuclear power plants; deterioration; Civil Engineering; Nuclear

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

APA (6^th Edition):

Jing, Y. (2018). Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants. (Doctoral Dissertation). University of Colorado. Retrieved from https://scholar.colorado.edu/cven_gradetds/360

Chicago Manual of Style (16^th Edition):

Jing, Yuxiang. “Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants.” 2018. Doctoral Dissertation, University of Colorado. Accessed March 02, 2021. https://scholar.colorado.edu/cven_gradetds/360.

MLA Handbook (7^th Edition):

Jing, Yuxiang. “Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants.” 2018. Web. 02 Mar 2021.

Vancouver:

Jing Y. Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants. [Internet] [Doctoral dissertation]. University of Colorado; 2018. [cited 2021 Mar 02]. Available from: https://scholar.colorado.edu/cven_gradetds/360.

Council of Science Editors:

Jing Y. Deterioration of Multi-Functional Cementitious Materials in Nuclear Power Plants. [Doctoral Dissertation]. University of Colorado; 2018. Available from: https://scholar.colorado.edu/cven_gradetds/360

Virginia Tech

28. Hor, Yuenkeung. The Daya Bay Reactor Neutrino Experiment.

Degree: PhD, Physics, 2014, Virginia Tech

URL: http://hdl.handle.net/10919/50523

► The Daya Bay experiment has determined the last unknown mixing angle θ₁₃. This thesis describes the layout of the experiment and the detector design. The… (more)

Subjects/Keywords: muon veto; anti-neutrino detector; attenuation length; spent fuel neutrino; reactor spectrum

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

APA (6^th Edition):

Hor, Y. (2014). The Daya Bay Reactor Neutrino Experiment. (Doctoral Dissertation). Virginia Tech. Retrieved from http://hdl.handle.net/10919/50523

Chicago Manual of Style (16^th Edition):

Hor, Yuenkeung. “The Daya Bay Reactor Neutrino Experiment.” 2014. Doctoral Dissertation, Virginia Tech. Accessed March 02, 2021. http://hdl.handle.net/10919/50523.

MLA Handbook (7^th Edition):

Hor, Yuenkeung. “The Daya Bay Reactor Neutrino Experiment.” 2014. Web. 02 Mar 2021.

Vancouver:

Hor Y. The Daya Bay Reactor Neutrino Experiment. [Internet] [Doctoral dissertation]. Virginia Tech; 2014. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/10919/50523.

Council of Science Editors:

Hor Y. The Daya Bay Reactor Neutrino Experiment. [Doctoral Dissertation]. Virginia Tech; 2014. Available from: http://hdl.handle.net/10919/50523

Brno University of Technology

29. Hlatký, Pavel. Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation.

Degree: 2019, Brno University of Technology

URL: http://hdl.handle.net/11012/17616

► This work deals with questions of spent fuel storage casks thermal and physical properties investigation. Foundations of mathematics which are necessary for describing field of… (more)

Subjects/Keywords: Přenos tepla; použití jaderné palivo; skladovací kontejner.; Heat transfer; spent nuclear fuel; storage cask.

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

APA (6^th Edition):

Hlatký, P. (2019). Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation. (Thesis). Brno University of Technology. Retrieved from http://hdl.handle.net/11012/17616

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):

Hlatký, Pavel. “Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation.” 2019. Thesis, Brno University of Technology. Accessed March 02, 2021. http://hdl.handle.net/11012/17616.

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

MLA Handbook (7^th Edition):

Hlatký, Pavel. “Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation.” 2019. Web. 02 Mar 2021.

Vancouver:

Hlatký P. Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation. [Internet] [Thesis]. Brno University of Technology; 2019. [cited 2021 Mar 02]. Available from: http://hdl.handle.net/11012/17616.

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

Council of Science Editors:

Hlatký P. Studium tepelných a fyzikálních vlastností skladovacích kontejnerů pro použité jaderné palivo: Spent fuel storage casks thermal and physical properties investigation. [Thesis]. Brno University of Technology; 2019. Available from: http://hdl.handle.net/11012/17616

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

University of Manchester

30. Foster, Lynn. Understanding the microbial ecology of highly radioactive nuclear storage facilities.

Degree: PhD, 2019, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/understanding-the-microbial-ecology-of-highly-radioactive-nuclear-storage-facilities(fcac8152-dbec-4260-ac8c-08f4580e0736).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.799458

► The First Generation Magnox Storage Pond (FGMSP) is situated on the Sellafield Ltd site in the UK, and is an extremely inhospitable environment comprising of… (more)

▼ The First Generation Magnox Storage Pond (FGMSP) is situated on the Sellafield Ltd site in the UK, and is an extremely inhospitable environment comprising of significant levels of radioactivity coupled with a high pH of pH 11.4. Despite such extreme conditions, microorganisms are known to colonise the pond and can form dense microbial blooms in the summer months. The blooms can restrict the visibility within the pond which hinders plant operations. The FGMSP is currently undergoing decommissioning and waste retrieval operations, as a priority on site, therefore any plant downtime increases both the cost and timeframe for decommissioning. Here we describe the microbial community that colonises the FGSMP, including during two bloom periods. In addition efforts to determine the adaptive mechanisms that key microorganisms use to colonise the pond and their interactions with Sr are described. Over the course of the sampling period Proteobacteria were the dominant phylum in the pond, with variations seen at the lower taxonomic levels. In addition, a single cyanobacterium, affiliated with a Pseudanabaena species, was identified as the dominant photosynthetic microorganism from samples taken from two bloom periods, comprising up to 30% of the phylotypes detected. While the FGMSP was dominated by prokaryotes, a hydraulically linked auxiliary pond was more abundant in eukaryotic organisms. Comparisons between the two pond communities suggested that the auxiliary pond was not seeding the FGMSP, as the elevated pH and radiation levels inhibited such colonisation. Data supplied by Sellafield Ltd. showed that the onset of the bloom periods coincided with increases in the residence time of the purge water, used to maintain the elevated pH of the pond. Once the residence time of the purge water was reduced the visibility was restored in the pond, indicating that this was an effective means of removing the bloom-forming microorganisms. Laboratory-cultures of Pseudanabaena catenata were used to investigate the adaptive responses to ionizing radiation. The culture was found to consist of 9 other operational taxonomic units, 5 of which were affiliated with genera identified in the FGMSP. Detailed investigations indicated that X-irradiation treatment (95Gy) had no significant impact on the growth rate of the culture, however there was an increase in polysaccharide production and a reduction in protein and chlorophyll-a production. Increases in polysaccharides could be of importance in the FGMSP as this could influence the fate of radionuclides present in the water. Sr was used to determine whether P. catenata could influence the fate of radionuclides. P. catenata cells could be seen to accumulate Sr associated with polyphosphate bodies, whilst SrPO4 and calcium containing SrCO3 minerals were formed. The colonisation of FGMSP by organisms closely related to those studied here, including the cyanobacterium Pseudanabaena catenata requires careful consideration. The results presented here suggest that elevated levels of polysaccharides could…

Subjects/Keywords: Cyanobacteria; First Generation Magnox Storage Pond; Peudanabaena; Radioactivity; Spent nuclear fuel ponds

Record Details Similar Records Cite « Share

❌

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

APA (6^th Edition):

Foster, L. (2019). Understanding the microbial ecology of highly radioactive nuclear storage facilities. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/understanding-the-microbial-ecology-of-highly-radioactive-nuclear-storage-facilities(fcac8152-dbec-4260-ac8c-08f4580e0736).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.799458

Chicago Manual of Style (16^th Edition):

Foster, Lynn. “Understanding the microbial ecology of highly radioactive nuclear storage facilities.” 2019. Doctoral Dissertation, University of Manchester. Accessed March 02, 2021. https://www.research.manchester.ac.uk/portal/en/theses/understanding-the-microbial-ecology-of-highly-radioactive-nuclear-storage-facilities(fcac8152-dbec-4260-ac8c-08f4580e0736).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.799458.

MLA Handbook (7^th Edition):

Foster, Lynn. “Understanding the microbial ecology of highly radioactive nuclear storage facilities.” 2019. Web. 02 Mar 2021.

Vancouver:

Foster L. Understanding the microbial ecology of highly radioactive nuclear storage facilities. [Internet] [Doctoral dissertation]. University of Manchester; 2019. [cited 2021 Mar 02]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/understanding-the-microbial-ecology-of-highly-radioactive-nuclear-storage-facilities(fcac8152-dbec-4260-ac8c-08f4580e0736).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.799458.

Council of Science Editors:

Foster L. Understanding the microbial ecology of highly radioactive nuclear storage facilities. [Doctoral Dissertation]. University of Manchester; 2019. Available from: https://www.research.manchester.ac.uk/portal/en/theses/understanding-the-microbial-ecology-of-highly-radioactive-nuclear-storage-facilities(fcac8152-dbec-4260-ac8c-08f4580e0736).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.799458

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