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Texas A&M University
1.
Kelly, Ryan Patrick.
Uncertainty Quantification of Concrete Utilized in Dry Cask Storage.
Degree: MS, Nuclear Engineering, 2013, Texas A&M University
URL: http://hdl.handle.net/1969.1/158896 
► The objective of this thesis is to quantify the uncertainty in radiation dose rate estimates outside of a used fuel dry cask storage unit due…
(more)
▼ The objective of this thesis is to quantify the uncertainty in radiation dose rate estimates outside of a used fuel
dry cask storage unit due to the parametric variability of concrete composition and density. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference
dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The model was developed using the Monte Carlo N-Particle code, with reference data from a used fuel assembly source provided by operators at Comanche Peak Nuclear Power Plant and geometry based on the Holtec HI-STORM 100S used fuel
dry cask storage system. The majority of the model was then fixed and the dose rates were compared when different concrete compositions at their nominal densities were substituted. Additional cases compared results for different concrete compositions at a fixed density, different densities for a fixed composition, and a different gamma energy source term for fixed compositions and densities.
The comparison of model results confirmed that the parametric variability of concrete composition is a major source of uncertainty for
dry cask dose rates. While precise results depend on the compositions compared, general trends cam be identified. The majority of the dose in all cases, typically 70%, depended on gamma rays produced by the fission products. Density was the dominant factor in determining the dose rate, as expected. Composition variation while density was held fixed, however, indicated that the precise composition has a large effect on the dose rates produced by neutrons and gamma rays produced by neutron capture, on the order of 70% or higher for test cases, with only a moderate impact on the dose rate from gamma rays produced by other sources, on the order of 20% for test cases. Alterations to the gamma energy spectra produced by additional decay uniformly lowered the dose, and did not significantly change comparative concrete performance. Overall results indicate that, due to the lack of standardization of concrete poured on site, additional safety factors may be necessary due to variation of shielding effectiveness.
Advisors/Committee Members: Tsvetkov, Pavel V (advisor), Chirayath, Sunil S (advisor), Poston, John W (committee member), Pate, Michael B (committee member).
Subjects/Keywords: concrete; shielding; composition; dry cask storage
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APA (6th Edition):
Kelly, R. P. (2013). Uncertainty Quantification of Concrete Utilized in Dry Cask Storage. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/158896
Chicago Manual of Style (16th Edition):
Kelly, Ryan Patrick. “Uncertainty Quantification of Concrete Utilized in Dry Cask Storage.” 2013. Masters Thesis, Texas A&M University. Accessed March 08, 2021.
http://hdl.handle.net/1969.1/158896.
MLA Handbook (7th Edition):
Kelly, Ryan Patrick. “Uncertainty Quantification of Concrete Utilized in Dry Cask Storage.” 2013. Web. 08 Mar 2021.
Vancouver:
Kelly RP. Uncertainty Quantification of Concrete Utilized in Dry Cask Storage. [Internet] [Masters thesis]. Texas A&M University; 2013. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/1969.1/158896.
Council of Science Editors:
Kelly RP. Uncertainty Quantification of Concrete Utilized in Dry Cask Storage. [Masters Thesis]. Texas A&M University; 2013. Available from: http://hdl.handle.net/1969.1/158896
Texas A&M University
2.
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 (6th 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 (16th 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 08, 2021.
http://hdl.handle.net/1969.1/158895.
MLA Handbook (7th 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. 08 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 08].
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
University of Tennessee – Knoxville
3.
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 (6th 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 (16th 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 08, 2021.
https://trace.tennessee.edu/utk_graddiss/5033.
MLA Handbook (7th Edition):
Liu, Zhengzhi. “METHODOLOGIES FOR IMAGING A USED NUCLEAR FUEL DRY STORAGE CASK WITH COSMIC RAY MUON COMPUTED TOMOGRAPHY.” 2018. Web. 08 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 08].
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 Cape Town
4.
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 (6th 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 (16th 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 08, 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 (7th Edition):
Khoza, Best. “Physics and engineering aspects of South Africa's proposed dry storage facility for spent nuclear fuel.” 2019. Web. 08 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 08].
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
Penn State University
5.
Mcnelly, Brendan P.
Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints.
Degree: 2015, Penn State University
URL: https://submit-etda.libraries.psu.edu/catalog/27409
► The goal of this thesis is to design a structurally- and environmentally-constrained, tethered and semi-autonomous robot that is able to inspect the interior crawl space…
(more)
▼ The goal of this thesis is to design a structurally- and environmentally-constrained, tethered and semi-autonomous robot that is able to inspect the interior crawl space around a nuclear
dry-
storage canister. The specific design aspects covered in this work include detailed analysis of the geometric, delivery, drivetrain, and thermal constraints on the robot that will host the sensing aspects of the inspection system.
The multisensory robot will have complex functionality during
cask inspections. This functionality includes insertion, motion, spatial sensing, extraction and reuse. The robotic system must be able to enter the
cask, determine spatial localization, navigate to the desired inspection location, and descend down the guide channel while recording measurements. During inspection a variety of measurements will be taken, including: temperature, material build-up, surface degradation and material composition. Measuring
cask surfaces for cracks and salt build-up, using a variety of sensor systems, will provide insight into the presence and progression of degradation within the
cask environment. Once the required measurements have been taken, the system will inspect other channels in a similar manner until the entirety of the
cask has been evaluated. Finally upon the completion of
cask inspection the robotic system will be extracted.
These requirements are constrained and limited by the geometry of the
cask, conditions of the environment, force, motion and power demands. The system will be geometrically designed to ensure safe delivery, integration, packaging, and fault-tolerant recovery of all sensor systems while maintaining full-functionality throughout inspections.
Advisors/Committee Members: Sean Brennan, Thesis Advisor/Co-Advisor, Karl Martin Reichard, Thesis Advisor/Co-Advisor, Henry Joseph Sommer Iii, Thesis Advisor/Co-Advisor.
Subjects/Keywords: robotic design; nuclear inspection; dry storage cask; wedging and jamming; peg insertion; thermal constraints
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Manager
APA (6th Edition):
Mcnelly, B. P. (2015). Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/27409
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Mcnelly, Brendan P. “Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints.” 2015. Thesis, Penn State University. Accessed March 08, 2021.
https://submit-etda.libraries.psu.edu/catalog/27409.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Mcnelly, Brendan P. “Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints.” 2015. Web. 08 Mar 2021.
Vancouver:
Mcnelly BP. Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints. [Internet] [Thesis]. Penn State University; 2015. [cited 2021 Mar 08].
Available from: https://submit-etda.libraries.psu.edu/catalog/27409.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Mcnelly BP. Design of a Nuclear Inspection Robot: Considering Dimensional, Force, Sensor Packaging, and Thermal Constraints. [Thesis]. Penn State University; 2015. Available from: https://submit-etda.libraries.psu.edu/catalog/27409
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
University of Houston
6.
-3679-2535.
Crash Analysis of Degraded Concrete Containment Structures.
Degree: PhD, Civil Engineering, 2017, University of Houston
URL: http://hdl.handle.net/10657/4822
► Dry storage casks are composite structures made of an inner steel liner and a concrete outerpack that houses the basket of the spent nuclear fuel…
(more)
▼ Dry storage casks are composite structures made of an inner steel liner and a concrete outerpack that houses the basket of the spent nuclear fuel bundle. The concrete of
dry cask structures will degrade due to the natural aging and the environmental effect shrinkage, creep, heat transfer, moisture diffusion, and Alkali-Silica Reaction (ASR). Long-term creep and shrinkage of vertical
dry cask are modeled using the B4 creep formulation for concrete. The results show that creep and shrinkage does not significantly affect the performance of the
dry cask structure. A fully coupled thermo-hygro-chemo-mechanical (THCM) framework is developed to consider the effect of degradation on the concrete outerpack. The coupling between heat transfer and moisture diffusion is verified through an experimental study on concrete and then implemented in the
dry cask structure. Also the coupling between the chemical (in form of ASR) and the mechanical processes is verified using experimental studies on concrete cylinders. First order kinetic approach is applied to model ASR for the concrete of
dry cask structure. Environmental aging is considered in the form of deleterious mismatch effects of steel and concrete deformation during the
cask performance. Environmental and mechanical degradation of concrete are determined for this
dry cask storage structure. A de-convolution strategy is developed to extract the material properties from the structural stiffness properties of the
cask. Using this strategy, the performance of continuous surface cap material (MAT-159) model is investigated, which is available in the material library of LS-DYNA. The behaviors of the intact and aged
dry cask structure and the model and prototype
cask are compared in the form of stresses, strains and failure modes. The 1:3-scale model
cask and the prototype
cask show similar behavior by applying the added mass to the 1:3-scale model
cask calculated based on the dynamic similitude theory. Also, implementing the result of mechanical and environmental degradation due to ASR to the
dry cask structure, introduces widespread damages to the
cask during the tip-over scenario, while the damage is still localized in case of the initial reference concrete. Two main failure modes during tip-over event are concrete crushing and shear banding.
Advisors/Committee Members: Willam, Kaspar J. (advisor), Gencturk, Bora E. (committee member), Belarbi, Abdeldjelil (committee member), Vipulanandan, Cumaraswamy (committee member), White, Kenneth W. (committee member).
Subjects/Keywords: Dry cask storage; Creep; Shrinkage; Heat transfer; Moisture transfer; Alkali-Silica Reaction (ASR); Tip-over
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APA ·
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Export
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APA (6th Edition):
-3679-2535. (2017). Crash Analysis of Degraded Concrete Containment Structures. (Doctoral Dissertation). University of Houston. Retrieved from http://hdl.handle.net/10657/4822
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Chicago Manual of Style (16th Edition):
-3679-2535. “Crash Analysis of Degraded Concrete Containment Structures.” 2017. Doctoral Dissertation, University of Houston. Accessed March 08, 2021.
http://hdl.handle.net/10657/4822.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
MLA Handbook (7th Edition):
-3679-2535. “Crash Analysis of Degraded Concrete Containment Structures.” 2017. Web. 08 Mar 2021.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Vancouver:
-3679-2535. Crash Analysis of Degraded Concrete Containment Structures. [Internet] [Doctoral dissertation]. University of Houston; 2017. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/10657/4822.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Council of Science Editors:
-3679-2535. Crash Analysis of Degraded Concrete Containment Structures. [Doctoral Dissertation]. University of Houston; 2017. Available from: http://hdl.handle.net/10657/4822
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
University of Houston
7.
-5414-1461.
Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks.
Degree: PhD, Civil Engineering, 2016, University of Houston
URL: http://hdl.handle.net/10657/5383
► The durability is one of the main reasons making concrete the most abundantly consumed material in the world after water. However, severe exposure conditions and…
(more)
▼ The durability is one of the main reasons making concrete the most abundantly consumed material in the world after water. However, severe exposure conditions and improper or low quality mixture designs can cause physical and chemical changes in concrete which lead to deterioration and premature failure of the structure. This process is called aging. Two of the main aging mechanisms in reinforced concrete structures are corrosion and alkali-silica reaction (ASR). These two processes occur very slowly and to assess them in a short period of time, they need to be accelerated. Furthermore, these mechanisms are dependent on the geometry and boundary conditions of the structure. Therefore, research on these aging mechanisms is structure specific and depends on the scale of observation. This study proposes a method through addition of chemicals to the concrete mixture for accelerating aging that is applicable to both small- and large- specimens. The aging is accelerated through addition of sodium hydroxide to accelerate ASR and calcium chloride to the mixture to accelerate corrosion. The effect of the addition of these chemicals on both physical and mechanical properties of concrete was investigated through a series of destructive and non-destructive testing at the materials and structural levels. The results indicate that the addition of sodium hydroxide to the concrete mix, combined with the use of reactive aggregate and no fly ash, considerably accelerates ASR and crack propagation on the surface of the specimens. Similarly, the addition of calcium chloride effectively accelerates corrosion. To investigate the validity of the proposed approach at the structural level, a vertical concrete
dry cask for
storage of nuclear waste was chosen as the case study. In this case study, three 1/3-scale
dry casks were subjected to accelerated ASR and corrosion with the addition of chemicals to the mixture and their effect was measured over eighteen months. Furthermore, the service life of the scaled down casks was estimated considering the real life conditions. This approach is expected to help researchers to better understand the long term behavior of reinforced concrete
dry casks.
Advisors/Committee Members: Gencturk, Bora E. (advisor), Willam, Kaspar J. (committee member), Vipulanandan, Cumaraswamy (committee member), Meen, James K. (committee member), Rodrigues, Debora F. (committee member).
Subjects/Keywords: ASR; Corrosion; Concrete; Aging; Dry cask storage; Service Life; Destructive Testing; Non-Destructive Testing
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APA ·
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APA (6th Edition):
-5414-1461. (2016). Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks. (Doctoral Dissertation). University of Houston. Retrieved from http://hdl.handle.net/10657/5383
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Chicago Manual of Style (16th Edition):
-5414-1461. “Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks.” 2016. Doctoral Dissertation, University of Houston. Accessed March 08, 2021.
http://hdl.handle.net/10657/5383.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
MLA Handbook (7th Edition):
-5414-1461. “Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks.” 2016. Web. 08 Mar 2021.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Vancouver:
-5414-1461. Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks. [Internet] [Doctoral dissertation]. University of Houston; 2016. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/10657/5383.
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Council of Science Editors:
-5414-1461. Accelerated Alkali-Silica Reaction and Corrosion of Reinforcing Steel in Concrete: Application on Dry Casks. [Doctoral Dissertation]. University of Houston; 2016. Available from: http://hdl.handle.net/10657/5383
Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete
Penn State University
8.
Cinbiz, Mahmut Nedim.
THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION.
Degree: 2015, Penn State University
URL: https://submit-etda.libraries.psu.edu/catalog/27328
► Prior to storage in a dry-cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up…
(more)
▼ Prior to
storage in a
dry-
cask facility, spent nuclear fuel must undergo a vacuum drying cycle during which the spent fuel rods are heated up to elevated temperatures of ≤ 400°C to remove moisture the canisters within the
cask. As temperature increases during heating, some of the hydride particles within the cladding dissolve while the internal gas pressure in fuel rods increases generating multi-axial hoop and axial stresses in the closed-end thin-walled cladding tubes. As cool-down starts, the hydrogen in solid solution precipitates as hydride platelets, and if the multi-axial stresses are sufficiently large, the precipitating hydrides reorient from their initial circumferential orientation to radial orientation. Radial hydrides can severely embrittle the spent nuclear fuel cladding at low temperature in response to hoop stress loading.
Because the cladding can experience a range of stress states during the thermo-mechanical treatment induced during vacuum drying, this study has investigated the effect of stress state on the process of hydride reorientation during controlled thermo-mechanical treatments utilizing the combination of in situ X-ray diffraction and novel mechanical testing analyzed by the combination of metallography and finite element analysis. The study used cold worked and stress relieved Zircaloy-4 sheet containing approx. 180 wt. ppm hydrogen as its material basis. The failure behavior of this material containing radial hydrides was also studied over a range of temperatures. Finally, samples from reactor-irradiated cladding tubes were examined by X-ray diffraction using synchrotron radiation.
To reveal the stress state effect on hydride reorientation, the critical threshold stress to reorient hydrides was determined by designing novel mechanical test samples which produce a range of stress states from uniaxial to “near-equibiaxial” tension when a load is applied. The threshold stress was determined after thermo-mechanical treatments by correlating the finite element stress-state results with the spatial distribution of hydride microstructures observed within the optical micrographs for each sample. Experiments showed that the hydride reorientation was enhanced as the stress biaxiality increased. The threshold stress decreased from 150 MPa to 80 MPa when stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension. This behavior was also predicted by classical nucleation theory based on the Gibbs free energy of transformation being assisted by the far-field stress.
An analysis of in situ X-ray diffraction data obtained during a thermo-mechanical cycle typical of vacuum drying showed a complex lattice-spacing behavior of the hydride phase during the dissolution and precipitation. The in-plane hydrides showed bilinear lattice expansion during heating with the intrinsic thermal expansion rate of the hydrides being observed only at elevated temperatures as they dissolve. For radial hydrides that precipitate during cooling under stress, the spacing of the close-packed {111}…
Advisors/Committee Members: Arthur Thompson Motta, Dissertation Advisor/Co-Advisor, Arthur Thompson Motta, Committee Chair/Co-Chair, Donald Albert Koss, Committee Member, Kenan Unlu, Committee Member, Clifford Jesse Lissenden Iii, Committee Member, Michael Billone, Special Member.
Subjects/Keywords: Zirconium; hydride; reorientation; dry cask storage; radial hydrides; stress; stress state; in situ; X-ray diffraction; biaxiality
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APA (6th Edition):
Cinbiz, M. N. (2015). THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/27328
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Cinbiz, Mahmut Nedim. “THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION.” 2015. Thesis, Penn State University. Accessed March 08, 2021.
https://submit-etda.libraries.psu.edu/catalog/27328.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Cinbiz, Mahmut Nedim. “THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION.” 2015. Web. 08 Mar 2021.
Vancouver:
Cinbiz MN. THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION. [Internet] [Thesis]. Penn State University; 2015. [cited 2021 Mar 08].
Available from: https://submit-etda.libraries.psu.edu/catalog/27328.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Cinbiz MN. THE EFFECT OF STRESS STATE ON ZIRCONIUM HYDRIDE REORIENTATION. [Thesis]. Penn State University; 2015. Available from: https://submit-etda.libraries.psu.edu/catalog/27328
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
University of Tennessee – Knoxville
9.
DeVoe, Remy Russell.
Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions.
Degree: MS, Nuclear Engineering, 2015, University of Tennessee – Knoxville
URL: https://trace.tennessee.edu/utk_gradthes/3469
► Dry cask safety analyses and material degradation studies depend on reliable best- estimate thermal models. Gaps in design data and storage site conditions require…
(more)
▼ Dry cask safety analyses and material degradation studies depend on reliable best- estimate thermal models. Gaps in design data and
storage site conditions require analysts to frequently rely on bounding assumptions in thermal models that introduce bias to the results. Efforts to gather physical attributes of
cask designs from vendors or experimental measurements can be costly and to determine all the uncertain parameters for a number of different
cask designs is impractical. Measuring the sensitivity of peak cladding temperature to
cask and fuel depletion modeling parameters was done to inform decision makers of which parameters that have the most impact on temperature predictions and should be further investigated. This study is applicable to vertical multi-purpose canister systems for long-term
storage evaluations such as those done for an interim
storage facility. The most sensitive
cask parameters are those that affect convective heat transfer in the
cask annulus and within the canister basket. These parameters are the ambient air temperature, canister pressure, and assembly decay heat profile with measured sensitivity coefficients of 0.50, -0.2841, and 0.0767, respectively. The sensitivity of peak cladding temperature to reactor cycle history was measured, and the most sensitive parameter was burnup, but other reactor operating history variations had little impact on temperature predictions.
Advisors/Committee Members: Steven E. Skutnik, G. Ivan Maldonado, Arthur E. Ruggles.
Subjects/Keywords: Thermal Modeling; Used Nuclear Fuel; Sensitivity Analysis; Nuclear Reactors; Dry Cask Storage; MAGNASTOR; Cladding Temperature; Nuclear Engineering
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APA (6th Edition):
DeVoe, R. R. (2015). Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions. (Thesis). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_gradthes/3469
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
DeVoe, Remy Russell. “Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions.” 2015. Thesis, University of Tennessee – Knoxville. Accessed March 08, 2021.
https://trace.tennessee.edu/utk_gradthes/3469.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
DeVoe, Remy Russell. “Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions.” 2015. Web. 08 Mar 2021.
Vancouver:
DeVoe RR. Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions. [Internet] [Thesis]. University of Tennessee – Knoxville; 2015. [cited 2021 Mar 08].
Available from: https://trace.tennessee.edu/utk_gradthes/3469.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
DeVoe RR. Sensitivity Analysis of Fuel Cladding Temperature to Dry Cask Loading and Storage Conditions. [Thesis]. University of Tennessee – Knoxville; 2015. Available from: https://trace.tennessee.edu/utk_gradthes/3469
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Virginia Tech
10.
Wang, Meng Jen.
Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking.
Degree: PhD, Nuclear Engineering, 2019, Virginia Tech
URL: http://hdl.handle.net/10919/97247
► Since the beginning of nuclear era, enormous amount of radiation applications have been proposed, developed, and applied in our daily life. The radiation is useful…
(more)
▼ Since the beginning of nuclear era, enormous amount of radiation applications have been proposed, developed, and applied in our daily life. The radiation is useful and beneficial when they are under control. However, there will be some "unwanted radiation" from these applications, which have to be shielded. For this, radiation shielding has become a very important task. To effectively shield the unwanted radiations, studying the thickness and design of the shields is important. Instead of directly performing experiments, computation is a more affordable and safer approach. The radiation shielding computation is typically an extremely difffficult task due to very limited "communication" between the radiation within the shield and detector outside the shield. In general, it is impractical to simulate the radiation shielding problems directly because the extremely expensive computation resources. Most of interactions of radiation are within the shield while we are only interested in how many of them penetrate through the shield. This is typically called "deep penetration" problems in the radiation transport community.
Advisors/Committee Members: Haghighat, Alireza (committeechair), Liu, Yang (committee member), Mahajan, Roop L. (committee member), Sjoden, Glenn Eric (committee member), Hin, Celine (committee member).
Subjects/Keywords: Particle Transport Theory; Real-time; Detector Response Function Methodology; Pressure Vessel Fluence Calculations; Dosimeter/Detector Response; Dry Storage Cask External Dosimetry
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Manager
APA (6th Edition):
Wang, M. J. (2019). Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking. (Doctoral Dissertation). Virginia Tech. Retrieved from http://hdl.handle.net/10919/97247
Chicago Manual of Style (16th Edition):
Wang, Meng Jen. “Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking.” 2019. Doctoral Dissertation, Virginia Tech. Accessed March 08, 2021.
http://hdl.handle.net/10919/97247.
MLA Handbook (7th Edition):
Wang, Meng Jen. “Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking.” 2019. Web. 08 Mar 2021.
Vancouver:
Wang MJ. Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking. [Internet] [Doctoral dissertation]. Virginia Tech; 2019. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/10919/97247.
Council of Science Editors:
Wang MJ. Development of a Novel Detector Response Formulation and Algorithm in RAPID and its Benchmarking. [Doctoral Dissertation]. Virginia Tech; 2019. Available from: http://hdl.handle.net/10919/97247
Virginia Tech
11.
Klein, Jackson Alexander.
Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes.
Degree: MS, Mechanical Engineering, 2017, Virginia Tech
URL: http://hdl.handle.net/10919/78031
► Dedicated sensors are widely used throughout many industries to monitor everyday operations, maintain safety, and report performance characteristics. In order to adopt a more sustainable…
(more)
▼ Dedicated sensors are widely used throughout many industries to monitor everyday operations, maintain safety, and report performance characteristics. In order to adopt a more sustainable solution, much research is being applied to self-powered sensing, implementing solutions which harvest wasted ambient energy sources to power these dedicated sensors. The adoption of not only wireless sensor nodes, but also self-powered capabilities in the nuclear energy process is critical as it can address issues in the overall safety and longevity of nuclear power. The removal of wires for data and power transmission can greatly reduce the cost of both installation and upkeep of power plants, while self-powered capabilities can further reduce effort and money spent in replacing batteries, and importantly may enable sensors to work even in losses to power across the plant, increasing plant safety. This thesis outlines three harvesting opportunities in the nuclear energy process from: thermal, vibration, and radiation sources in the main structure of the power plant, and from thermal and radiation energy from spent fuel in
dry cask storage. Thermal energy harvesters for the primary and secondary coolant loops are outlined, and experimental analysis done on their longevity in high-radiation environments is discussed. A vibrational energy harvester for large rotating plant machine vibration is designed, prototyped, and tested, and a model is produced to describe its motion and energy output. Finally, an introduction to the design of a gamma radiation and thermal energy harvester for spent nuclear fuel canisters is discussed, and further research steps are suggested.
Advisors/Committee Members: Zuo, Lei (committeechair), Huxtable, Scott T. (committee member), Pierson, Mark Alan (committee member).
Subjects/Keywords: Energy Harvesting; Nuclear Power Plant; Wireless Sensor Node; Electromagnetic Generator; Compliant Mechanism; Thermoelectric Generator; Gamma Radiation; MCNP Simulation; Gamma Heating; Dry Cask Storage
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APA (6th Edition):
Klein, J. A. (2017). Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes. (Masters Thesis). Virginia Tech. Retrieved from http://hdl.handle.net/10919/78031
Chicago Manual of Style (16th Edition):
Klein, Jackson Alexander. “Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes.” 2017. Masters Thesis, Virginia Tech. Accessed March 08, 2021.
http://hdl.handle.net/10919/78031.
MLA Handbook (7th Edition):
Klein, Jackson Alexander. “Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes.” 2017. Web. 08 Mar 2021.
Vancouver:
Klein JA. Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes. [Internet] [Masters thesis]. Virginia Tech; 2017. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/10919/78031.
Council of Science Editors:
Klein JA. Energy Harvesting Opportunities Throughout the Nuclear Power Cycle for Self-Powered Wireless Sensor Nodes. [Masters Thesis]. Virginia Tech; 2017. Available from: http://hdl.handle.net/10919/78031
Brno University of Technology
12.
Jestřáb, Tomáš.
Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties.
Degree: 2019, Brno University of Technology
URL: http://hdl.handle.net/11012/39700
► The main goal of this Bachelor’s thesis is to develop fuel depletion models of a nuclear fuel used at Nuclear Power Plant Dukovany – originally…
(more)
▼ The main goal of this Bachelor’s thesis is to develop fuel depletion models of a nuclear fuel used at Nuclear Power Plant Dukovany – originally designed nonprofiled fuel used in a three-year cycle, inovated profiled fuel used in a four-year cycle and currently used fuel Gd-2M+ used in a five year fuel cycle. Models are created by computational code UWB1 providing fast calculation of nuclear fuel depletion. Outputs of these models are in thesis graphically displayed and analyzed from the point of view of fuel depletion, enrichment or irradiation time spent in reactor core. These analyses are based on theory including information about the nuclear fuel cycle focusing on its middle part and the back end of the nuclear fuel cycle.
Advisors/Committee Members: Katovský, Karel (advisor), Řež,, Martina Malá, Centrum výzkumu (referee).
Subjects/Keywords: vyhořelé jaderné palivo; jaderný palivový cyklus; bazén skladování; suchý mezisklad; mokrý mezisklad; obalový soubor; Dukovany; UWB1; spent nuclear fuel; nuclear fuel cycle; spent fuel pool; dry storage; wet storage; nuclear waste cask; Dukovany; UWB1
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APA ·
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APA (6th Edition):
Jestřáb, T. (2019). Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties. (Thesis). Brno University of Technology. Retrieved from http://hdl.handle.net/11012/39700
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Jestřáb, Tomáš. “Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties.” 2019. Thesis, Brno University of Technology. Accessed March 08, 2021.
http://hdl.handle.net/11012/39700.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Jestřáb, Tomáš. “Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties.” 2019. Web. 08 Mar 2021.
Vancouver:
Jestřáb T. Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties. [Internet] [Thesis]. Brno University of Technology; 2019. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/11012/39700.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Jestřáb T. Použité jaderné palivo a jeho vlastnosti: Spent nuclear fuel and its properties. [Thesis]. Brno University of Technology; 2019. Available from: http://hdl.handle.net/11012/39700
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
13.
Wierschke, Jonathan Brett.
Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel.
Degree: PhD, Nuclear Engineering and Radiological Sciences, 2015, University of Michigan
URL: http://hdl.handle.net/2027.42/111438
► The objective of this work was to understand the long-term corrosion behavior of Boral® and Bortec® neutron absorbers during deployment in a used nuclear fuel…
(more)
▼ The objective of this work was to understand the long-term corrosion behavior of Boral® and Bortec® neutron absorbers during deployment in a used nuclear fuel
dry cask storage environment for several hundred years. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures between 300 and 570°C. Humidity levels ranged from 0.028 to 0.58 mass fraction with flow rates ranging from approximately 0.2 scfh to 1.5 scfh.
Results from corrosion testing at temperatures between 300 and 570°C with varying humidity have shown that both Boral® and Bortec® develop new aluminum-boron-carbon phases. The phases formed are consistent at 300 and 400°C. Different formations were observed at 570°C. The samples also showed boron depletion at all temperatures.
It is predicted that two mechanisms control the changes in Boral® and Bortec®. The phase changes observed result from the interaction of boron carbide with aluminum. These interactions result in boron and carbon diffusing into the aluminum matrix. The other series of interactions occurring between the sample and the water in the humidified argon. Boron on the surface rapidly reacts with the water to form B2O3 which can be volatilized. The loss of boron at the surface creates a concentration profile that can result in the continued diffusion of boron to the surface. The water will also react with the aluminum to form Al2O3. Aluminum oxide formed in humid conditions has some porosity and can degrade away and allow for continued oxidation of aluminum.
Additional research is required to determine the suitability of these materials for use in
dry cask storage. The new phase formation may affect mechanical properties and adversely affect the fuel baskets structural integrity. Boron redistribution may cause localized areas of boron depletion, additional testing needs to be conducted to determine boron diffusion in these materials in the absence of humidity. In the presence of humidity the boron will leach from the samples.
Cask humidity levels need to be determined to be able to predict how much boron may be lost.
Advisors/Committee Members: Wang, Lumin (committee member), Daly, Samantha Hayes (committee member), Gao, Fei (committee member), Hanson, Brady D. (committee member).
Subjects/Keywords: used nuclear fuel; dry cask storage; Boral; Bortec; aluminum boron carbide; Materials Science and Engineering; Nuclear Engineering and Radiological Sciences; Engineering; Science
…deployment in a used nuclear fuel dry cask storage
environment for several hundred years. Corrosion… …for use in
dry cask storage. The new phase formation may affect mechanical properties and… …an initiative to move more of the used fuel inventory into dry cask storage.
1.1.1 Used… …the deployment of
dry cask storage of used fuel assemblies that have cooled. Even though… …the pool and cooled using natural
convection in dry cask storage.
Figure 3 Decay heat in…
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APA ·
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CSE |
Export
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APA (6th Edition):
Wierschke, J. B. (2015). Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel. (Doctoral Dissertation). University of Michigan. Retrieved from http://hdl.handle.net/2027.42/111438
Chicago Manual of Style (16th Edition):
Wierschke, Jonathan Brett. “Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel.” 2015. Doctoral Dissertation, University of Michigan. Accessed March 08, 2021.
http://hdl.handle.net/2027.42/111438.
MLA Handbook (7th Edition):
Wierschke, Jonathan Brett. “Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel.” 2015. Web. 08 Mar 2021.
Vancouver:
Wierschke JB. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel. [Internet] [Doctoral dissertation]. University of Michigan; 2015. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/2027.42/111438.
Council of Science Editors:
Wierschke JB. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel. [Doctoral Dissertation]. University of Michigan; 2015. Available from: http://hdl.handle.net/2027.42/111438
14.
Spencer, Kristina Yancey.
Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns.
Degree: PhD, Nuclear Engineering, 2017, Texas A&M University
URL: http://hdl.handle.net/1969.1/173097
► To address the evolving needs of dry storage, this research developed an optimization methodology to identify loading configurations to minimize the number of casks, their…
(more)
▼ To address the evolving needs of
dry storage, this research developed an optimization
methodology to identify loading configurations to minimize the number of
casks, their heat load, and the time when they meet transportation requirements.
The motivation was to investigate strategies that balance and reduce risk over the
lifetime of a site's reactor(s).
The
dry cask loading problem was formulated as an adaptable dynamic bin
packing problem, accommodating different site and
cask limits in broadly-defined
constraints. A new method was developed to address its complexities, named the
GRASP-enabled adaptive multiobjective memetic algorithm with partial clustering
(GAMMA-PC). This method embeds greedy randomized adaptive search procedures
in a multiobjective evolutionary algorithm with local search techniques and partial
decomposition of the objective space during crossover.
GAMMA-PC was demonstrated through integration with the unified database
from the Used Fuel Systems group at Oak Ridge National Laboratory to optimize
simulated loading configurations for Vermont Yankee, Comanche Peak, and Zion
Nuclear Power Stations. Its performance was evaluated through comparisons to test
solutions and to the real Zion loading configuration. GAMMA-PC produced diverse
solutions that dominated the testing sets. The improvement was concentrated in
the average heat load, and the third objective function was shown to be sensitive to
individual assembly characteristics.
The results suggested the usefulness of GAMMA-PC for utilities considering long-term
goals. They showed that more diverse
cask loadings and strategic placements
of empty positions can be used to reduce initial heat loads. Moving to a higher capacity
cask increases loading flexibility but can result in transportation delays.
Long-term planning enables a more thorough consideration of the trade-offs involved
in any decision.
This research contributes one of the first in-depth studies of the
dry cask loading
problem. It expands the current treatment of assembly selection over longer timeframes
and meets user-defined requirements. It is also one of the first tri-objective
dynamic bin packing problems, and the first to pack items with time-dependent
characteristics. Future work should focus on refining the objectives and incorporating
uncertainty. With its adaptable structure, GAMMA-PC is a promising new
metaheuristic for this task and for dynamic bin packing problems in general.
Advisors/Committee Members: Tsvetkov, Pavel V (advisor), Butenko, Sergiy (committee member), Jarrell, Joshua J (committee member), McDeavitt, Sean M (committee member), Peddicord, Kenneth L (committee member).
Subjects/Keywords: nuclear waste management; used nuclear fuel; dry cask storage; decay heat; multiobjective optimization; combinatorial optimization; bin packing problems; memetic algorithms; GRASP; GAMMA-PC
…Storage
The U.S. does not have a standardized dry cask design. Instead, a few companies
offer… …HI-STORM UMAX canister storage system (CoC 1040),
10. the TN-32 dry storage cask… …x28;CoC 1021),
11. the TN-68 dry storage cask (CoC 1027),
12. the NAC-UMS… …Projected
Dry Casks
to Be Filled
Already in
Dry Cask
Storage
Figure 1.2: Illustration of dry… …patterns of used fuel assemblies in dry cask storage, accounting for
current and future inventory…
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APA ·
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Export
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Manager
APA (6th Edition):
Spencer, K. Y. (2017). Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/173097
Chicago Manual of Style (16th Edition):
Spencer, Kristina Yancey. “Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns.” 2017. Doctoral Dissertation, Texas A&M University. Accessed March 08, 2021.
http://hdl.handle.net/1969.1/173097.
MLA Handbook (7th Edition):
Spencer, Kristina Yancey. “Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns.” 2017. Web. 08 Mar 2021.
Vancouver:
Spencer KY. Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns. [Internet] [Doctoral dissertation]. Texas A&M University; 2017. [cited 2021 Mar 08].
Available from: http://hdl.handle.net/1969.1/173097.
Council of Science Editors:
Spencer KY. Adaptable Long-Term Optimization of Dry Cask Storage Loading Patterns. [Doctoral Dissertation]. Texas A&M University; 2017. Available from: http://hdl.handle.net/1969.1/173097
Penn State University
15.
Kern, Ludwig August.
FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK.
Degree: 2009, Penn State University
URL: https://submit-etda.libraries.psu.edu/catalog/9685
► A FLUENT Computational Fluid Dynamics (CFD) model of a spent nuclear fuel dry storage cask was created to predict several critical design temperatures in order…
(more)
▼ A FLUENT Computational Fluid Dynamics (CFD) model of a spent nuclear fuel
dry storage cask was created to predict several critical design temperatures in order to aid in the design of a
dry storage cask. Thermal tests were conducted to determine the ability of the FLUENT CFD code to accurately predict solid structure and fluid temperatures in simple and complex flow geometries. The thermal tests included natural convective flow in a single heated annular tube and realistic heating of a full-scale prototype
dry storage cask using heating configurations which bounded the heat source distributions of all possible fuel loading patterns. The FLUENT CFD model of the prototype
dry storage cask was successful in demonstrating reasonable accuracy (20% agreement) in predicting the majority of temperatures measured during the thermal tests. The model showed even better accuracy in predicting the hottest rod temperature, the hottest temperature within the
cask, which is a critical design temperature, and conservatively bounded all rod temperatures if allowance is given for the measurement uncertainties in the experimental data. However, there was no validation during this set of thermal experiments of the FLUENT CFD model’s ability to accurately model heat transfer between adjacent rods in individual fuel assemblies, which will be necessary before the model is ready for use as part of the thermal design basis for spent nuclear fuel
dry storage cask design. A general discussion of computational methods, modeling methodology, FLUENT and other CFD codes, spent nuclear fuel
storage, the design of nuclear fuel
dry storage casks, and the heat transfer mechanisms modeled for the
dry storage cask is also included in addition to the information specific to the
dry storage cask FLUENT CFD model and the model validation experiments conducted with the full-scale prototype
cask.
Advisors/Committee Members: Fan Bill B Cheung, Thesis Advisor/Co-Advisor, Fan Bill B Cheung, Thesis Advisor/Co-Advisor, John Michael Cimbala, Thesis Advisor/Co-Advisor.
Subjects/Keywords: Fluent; fluent modeling; cfd; cfd modeling; fluent model; cfd model; computational fluid dynamics; nuclear; nuclear engineering; dry storage; dry storage cask; spent fuel; spent nuclear fuel; nuclear fuel; rbmk; chernobyl; holtec; thermal; heat transfer; natural convection; convection; natural convective flow; annular; prototype; full-scale; validation; cfd code; navier-stokes; navier-stokes equation
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APA (6th Edition):
Kern, L. A. (2009). FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/9685
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Kern, Ludwig August. “FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK.” 2009. Thesis, Penn State University. Accessed March 08, 2021.
https://submit-etda.libraries.psu.edu/catalog/9685.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Kern, Ludwig August. “FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK.” 2009. Web. 08 Mar 2021.
Vancouver:
Kern LA. FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK. [Internet] [Thesis]. Penn State University; 2009. [cited 2021 Mar 08].
Available from: https://submit-etda.libraries.psu.edu/catalog/9685.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Kern LA. FLUENT CFD MODELING IN DESIGN OF SPENT NUCLEAR FUEL DRY STORAGE CASK. [Thesis]. Penn State University; 2009. Available from: https://submit-etda.libraries.psu.edu/catalog/9685
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
. 
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