subject:(Nuclear Graphite)

University of Pretoria

1. Murovhi, Phathutshedzo. Low temperature thermal properties of HTR nuclear fuel composite graphite.

Degree: Physics, 2013, University of Pretoria

► Graphite and graphite composite materials are of great importance in various applications; however, they have been widely used in nuclear applications. Primarily in nuclear applications… (more)

▼ Graphite and graphite composite materials are of great importance in various applications; however, they have been widely used in nuclear applications. Primarily in nuclear applications such, as a moderator where its primary aim is to stop the fast neutrons to thermal neutron. The composite graphite (HTR-10) has potential applications as a moderator and other applications including in aerospace field. Structurally the composite shows stable hexagonal form of graphite and no traces of the unstable Rhombohedral patterns. Thermal conductivity indicates the same trends observed and known for nuclear graded graphite. The composite was made as a mixture of 64 wt% of natural graphite, 16 wt% of synthetic graphite binded together by 20 wt% of phenolic resin. The resinated graphite powder was uni-axially pressed by 19.5 MPa to form a disc shaped specimen. The disc was then cut and annealed to 1800 °C. The composite was further cut into two directions (parallel and perpendicular) to the pressing direction. For characterization the samples were cut into 2.5 x 2.5 x 10 mm3. There were exposed to proton irradiation for 3 and 4.5 hrs respectively and characterized both structurally and thermally. Through the study what we have observed was that as the composite is exposed to proton irradiation there is an improvement structurally. Thus, the D peak in the Raman spectroscopy has decreased substantially with the irradiated samples. XRD has indicated that there is no un-stable Rhombohedral phase pattern in both the pristine and the irradiated samples. However this was further confirmed with that thermal conductivity is also increasing with irradiation exposure. This is anomalous to irradiated graphite in which defects are supposedly induced. Looking into the electrical resistivity we have noted that pristine samples have higher resistivity as compared to the irradiated samples. Seebeck coefficient indicates that there is some form of structural perfection and the samples have a phonon drag dip at the known graphite temperature of 35 K. This has shown us there are no impurities induced by irradiation of the samples. Advisors/Committee Members: Manyala, Ncholu I. (advisor).

Subjects/Keywords: Graphite; Graphite composite materials; Nuclear applications; UCTD

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APA (6^th Edition):

Murovhi, P. (2013). Low temperature thermal properties of HTR nuclear fuel composite graphite. (Masters Thesis). University of Pretoria. Retrieved from http://hdl.handle.net/2263/33156

Chicago Manual of Style (16^th Edition):

Murovhi, Phathutshedzo. “Low temperature thermal properties of HTR nuclear fuel composite graphite.” 2013. Masters Thesis, University of Pretoria. Accessed January 18, 2021. http://hdl.handle.net/2263/33156.

MLA Handbook (7^th Edition):

Murovhi, Phathutshedzo. “Low temperature thermal properties of HTR nuclear fuel composite graphite.” 2013. Web. 18 Jan 2021.

Vancouver:

Murovhi P. Low temperature thermal properties of HTR nuclear fuel composite graphite. [Internet] [Masters thesis]. University of Pretoria; 2013. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/2263/33156.

Council of Science Editors:

Murovhi P. Low temperature thermal properties of HTR nuclear fuel composite graphite. [Masters Thesis]. University of Pretoria; 2013. Available from: http://hdl.handle.net/2263/33156

University of Manchester

2. Bodel, William. THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE.

Degree: 2013, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:198592

► In addition to its role as moderator within British nuclear reactors, polycrystalline graphite is also a major structural component of the core, enabling access for… (more)

▼ In addition to its role as moderator within British nuclear reactors, polycrystalline graphite is also a major structural component of the core, enabling access for control rods, coolant gas and fuel. Aging processes, primarily fast neutron irradiation and radiolytic oxidation lead to distortion of the graphite components and property changes which ultimately reduce the material’s effectiveness and can lead to component failure.Despite much research into the material, graphite behaviour under irradiation conditions is not fully understood and has resulted in an overestimation of the extent of component failures in Magnox reactors, and a subsequent underestimation of component failures in the following generation Advanced Gas-cooled Reactors (AGRs). A greater understanding of the material is therefore required in order to make more informed evaluations as part of on-going safety cases.Young’s modulus is one property which varies as a complex function of radiolytic oxidation and fast neutron irradiation dose; this work investigates investigate the Young’s modulus behaviour of nuclear grade graphites through property measurement and microstructural characterisation. Physical properties are dependent on microstructure, which is in turn a result of the manufacturing processes and raw materials used in its fabrication. Because of this, this thesis begins with a microstructural study of AGR graphite artefacts from varying points during the manufacturing process and post-irradiation, utilising X-ray diffraction to observe changes in crystallinity, microscopy to directly observe the microstructure and pycnometry to gauge porosity variations. Increases in crystallinity towards graphitisation are seen, with a subsequent decrease after irradiation; and significant changes are observed from inspection of optical and scanning electron micrographs.Young’s modulus property data are obtained using a combination of static and dynamic techniques to accumulate data from a variety of techniques. An experiment designed to track changes to the speed of sound under compressive load was carried out on Magnox and AGR graphite, showing different behaviour between the grades, and variation with irradiation.A final series of tests combine compressive testing with in-situ microscopy to try and better understand the reasons behind this varied in behaviour and relate microstructural changes to graphite behaviour under compressive loading. Advisors/Committee Members: JONES, ABBIE A, Marsden, Barry, Jones, Abbie.

Subjects/Keywords: Nuclear; Graphite; Young's Modulus

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APA (6^th Edition):

Bodel, W. (2013). THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:198592

Chicago Manual of Style (16^th Edition):

Bodel, William. “THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE.” 2013. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:198592.

MLA Handbook (7^th Edition):

Bodel, William. “THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE.” 2013. Web. 18 Jan 2021.

Vancouver:

Bodel W. THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE. [Internet] [Doctoral dissertation]. University of Manchester; 2013. [cited 2021 Jan 18]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:198592.

Council of Science Editors:

Bodel W. THE RELATIONSHIP BETWEEN MICROSTRUCTURE AND YOUNG’S MODULUS OF NUCLEAR GRAPHITE. [Doctoral Dissertation]. University of Manchester; 2013. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:198592

University of Manchester

3. Morrison, Craig Neil. Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes.

Degree: 2016, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:299976

► The integrity of graphite components is critical for their fitness for purpose. Since graphite is a quasi-brittle material the dominant mechanism for loss of integrity… (more)

Subjects/Keywords: nuclear graphite; lattice modelling

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APA (6^th Edition):

Morrison, C. N. (2016). Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:299976

Chicago Manual of Style (16^th Edition):

Morrison, Craig Neil. “Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes.” 2016. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:299976.

MLA Handbook (7^th Edition):

Morrison, Craig Neil. “Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes.” 2016. Web. 18 Jan 2021.

Vancouver:

Morrison CN. Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes. [Internet] [Doctoral dissertation]. University of Manchester; 2016. [cited 2021 Jan 18]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:299976.

Council of Science Editors:

Morrison CN. Lattice-Modelling of Nuclear Graphite for Improved Understanding of Fracture Processes. [Doctoral Dissertation]. University of Manchester; 2016. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:299976

University of Plymouth

4. Jones, Katie. A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite.

Degree: PhD, 2019, University of Plymouth

URL: http://hdl.handle.net/10026.1/13436

► The pore-level characterisation of nuclear graphite is critical for predicting reactor safety and for assessing the viability of different grades of graphite material. The majority… (more)

▼ The pore-level characterisation of nuclear graphite is critical for predicting reactor safety and for assessing the viability of different grades of graphite material. The majority of studies often focus on the impact of one specific length scale (macroscopic/mesoscopic/ microscopic) but very few studies have attempted to provide void size information which spans multiple length scales. This study, therefore, aimed to advance the knowledge of the entire void range for Gilsocarbon graphite, including any changes to the void structure that occur as a consequence of irradiation damage and radiolytic oxidation, by incorporating a combination of experimental and modelling techniques. Gilsocarbon graphite, which is incorporated in the current UK Advanced gas cooled reactors, is particularly difficult to characterise at a pore level due to its highly complex pore matrix which comprises voidage over several orders of magnitude. Additionally, the radioactive nature of the samples limits the amount of material available for analysis. In order to successfully measure the small sample volumes provided, novel instrumentation and interpretation methods were developed. This included the construction of a micropycnometer, built to obtain values of the accessible and inaccessible pore volumes. In addition, graphite's low surface area demanded the use of krypton as an adsorbative, which required the acquisition of a high performance instrument as well as the development of an interpretive GCMC kernel for obtaining pore size information. These data were used to correct the pore size information obtained at high pressure during mercury porosimetry, as the porosimetry data was suspected to contain inaccuracies due to damage or deformation of the graphite's microstructure caused by the analysis. The bespoke software package PoreXpert, designed at the University of Plymouth, was used to inverse model the experimentally measured percolation characteristics and total accessible porosity to generate simulated void network structures. The improved, quasi-Bayesian, modelling of the combined percolation curves identified differences in the pore size distributions for Gilsocarbon samples during various stages of ageing. The findings from the bespoke models complemented the experimental results, in that the findings supported the idea of uniform evolution for all pore-throat entrance sizes and provided a more robust modelling procedure which together enhanced the understanding of the mechanistic interpretations. Such findings contradict the current weight loss prediction models, but complement the working hypothesis formulated within EDF Energy graphite research group. Therefore, the experimental and modelled results will feature in a supportive document of proposed revisions to the EDF Energy Safety Case submitted to the Office for Nuclear Regulation.

Subjects/Keywords: Nuclear Graphite; Characterisation; Porosity

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

APA (6^th Edition):

Jones, K. (2019). A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite. (Doctoral Dissertation). University of Plymouth. Retrieved from http://hdl.handle.net/10026.1/13436

Chicago Manual of Style (16^th Edition):

Jones, Katie. “A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite.” 2019. Doctoral Dissertation, University of Plymouth. Accessed January 18, 2021. http://hdl.handle.net/10026.1/13436.

MLA Handbook (7^th Edition):

Jones, Katie. “A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite.” 2019. Web. 18 Jan 2021.

Vancouver:

Jones K. A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite. [Internet] [Doctoral dissertation]. University of Plymouth; 2019. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/10026.1/13436.

Council of Science Editors:

Jones K. A study of the effects of irradiation and radiolytic oxidation of the pore-level structure of gilsocarbon nuclear graphite. [Doctoral Dissertation]. University of Plymouth; 2019. Available from: http://hdl.handle.net/10026.1/13436

University of Manchester

5. Bodel, William. The relationship between microstructure and Young's modulus of nuclear graphite.

Degree: Thesis (Eng.D.), 2013, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/the-relationship-between-microstructure-and-youngs-modulus-of-nuclear-graphite(ac5fe868-cefb-4f0c-8b22-a8904bc97da5).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.576873

► In addition to its role as moderator within British nuclear reactors, polycrystalline graphite is also a major structural component of the core, enabling access for… (more)

▼ In addition to its role as moderator within British nuclear reactors, polycrystalline graphite is also a major structural component of the core, enabling access for control rods, coolant gas and fuel. Aging processes, primarily fast neutron irradiation and radiolytic oxidation lead to distortion of the graphite components and property changes which ultimately reduce the material's effectiveness and can lead to component failure.Despite much research into the material, graphite behaviour under irradiation conditions is not fully understood and has resulted in an overestimation of the extent of component failures in Magnox reactors, and a subsequent underestimation of component failures in the following generation Advanced Gas-cooled Reactors (AGRs). A greater understanding of the material is therefore required in order to make more informed evaluations as part of on-going safety cases. Young's modulus is one property which varies as a complex function of radiolytic oxidation and fast neutron irradiation dose; this work investigates investigate the Young's modulus behaviour of nuclear grade graphites through property measurement and microstructural characterisation. Physical properties are dependent on microstructure, which is in turn a result of the manufacturing processes and raw materials used in its fabrication. Because of this, this thesis begins with a microstructural study of AGR graphite artefacts from varying points during the manufacturing process and post-irradiation, utilising X-ray diffraction to observe changes in crystallinity, microscopy to directly observe the microstructure and pycnometry to gauge porosity variations. Increases in crystallinity towards graphitisation are seen, with a subsequent decrease after irradiation; and significant changes are observed from inspection of optical and scanning electron micrographs. Young's modulus property data are obtained using a combination of static and dynamic techniques to accumulate data from a variety of techniques. An experiment designed to track changes to the speed of sound under compressive load was carried out on Magnox and AGR graphite, showing different behaviour between the grades, and variation with irradiation.A final series of tests combine compressive testing with in-situ microscopy to try and better understand the reasons behind this varied in behaviour and relate microstructural changes to graphite behaviour under compressive loading.

Subjects/Keywords: 553.2; Nuclear; Graphite; Young's Modulus

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APA (6^th Edition):

Bodel, W. (2013). The relationship between microstructure and Young's modulus of nuclear graphite. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/the-relationship-between-microstructure-and-youngs-modulus-of-nuclear-graphite(ac5fe868-cefb-4f0c-8b22-a8904bc97da5).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.576873

Chicago Manual of Style (16^th Edition):

Bodel, William. “The relationship between microstructure and Young's modulus of nuclear graphite.” 2013. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/the-relationship-between-microstructure-and-youngs-modulus-of-nuclear-graphite(ac5fe868-cefb-4f0c-8b22-a8904bc97da5).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.576873.

MLA Handbook (7^th Edition):

Bodel, William. “The relationship between microstructure and Young's modulus of nuclear graphite.” 2013. Web. 18 Jan 2021.

Vancouver:

Bodel W. The relationship between microstructure and Young's modulus of nuclear graphite. [Internet] [Doctoral dissertation]. University of Manchester; 2013. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/the-relationship-between-microstructure-and-youngs-modulus-of-nuclear-graphite(ac5fe868-cefb-4f0c-8b22-a8904bc97da5).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.576873.

Council of Science Editors:

Bodel W. The relationship between microstructure and Young's modulus of nuclear graphite. [Doctoral Dissertation]. University of Manchester; 2013. Available from: https://www.research.manchester.ac.uk/portal/en/theses/the-relationship-between-microstructure-and-youngs-modulus-of-nuclear-graphite(ac5fe868-cefb-4f0c-8b22-a8904bc97da5).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.576873

University of Manchester

6. Morrison, Craig Neil. Lattice-modelling of nuclear graphite for improved understanding of fracture processes.

Degree: PhD, 2016, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/latticemodelling-of-nuclear-graphite-for-improved-understanding-of-fracture-processes(10b302d1-88fb-466b-9030-d34b4fc33293).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706240

► The integrity of graphite components is critical for their fitness for purpose. Since graphite is a quasi-brittle material the dominant mechanism for loss of integrity… (more)

Subjects/Keywords: 620.1; nuclear graphite; lattice modelling

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

APA (6^th Edition):

Morrison, C. N. (2016). Lattice-modelling of nuclear graphite for improved understanding of fracture processes. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/latticemodelling-of-nuclear-graphite-for-improved-understanding-of-fracture-processes(10b302d1-88fb-466b-9030-d34b4fc33293).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706240

Chicago Manual of Style (16^th Edition):

Morrison, Craig Neil. “Lattice-modelling of nuclear graphite for improved understanding of fracture processes.” 2016. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/latticemodelling-of-nuclear-graphite-for-improved-understanding-of-fracture-processes(10b302d1-88fb-466b-9030-d34b4fc33293).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706240.

MLA Handbook (7^th Edition):

Morrison, Craig Neil. “Lattice-modelling of nuclear graphite for improved understanding of fracture processes.” 2016. Web. 18 Jan 2021.

Vancouver:

Morrison CN. Lattice-modelling of nuclear graphite for improved understanding of fracture processes. [Internet] [Doctoral dissertation]. University of Manchester; 2016. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/latticemodelling-of-nuclear-graphite-for-improved-understanding-of-fracture-processes(10b302d1-88fb-466b-9030-d34b4fc33293).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706240.

Council of Science Editors:

Morrison CN. Lattice-modelling of nuclear graphite for improved understanding of fracture processes. [Doctoral Dissertation]. University of Manchester; 2016. Available from: https://www.research.manchester.ac.uk/portal/en/theses/latticemodelling-of-nuclear-graphite-for-improved-understanding-of-fracture-processes(10b302d1-88fb-466b-9030-d34b4fc33293).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.706240

University of Manchester

7. Worth, Robert Numa. Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite.

Degree: 2016, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:301365

► THERMAL TREATMENT OF OLDBURY MAGNOX REACTOR IRRADIATED GRAPHITEApproximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current… (more)

▼ THERMAL TREATMENT OF OLDBURY MAGNOX REACTOR IRRADIATED GRAPHITEApproximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current Nuclear Decommissioning Authority (NDA) baseline strategy for irradiated graphite in England and Wales is isolation in a future Geological Disposal Facility, with Scottish policy endorsing an alternative decision of near surface long-term storage. Irradiated graphite disposal routes in the UK remain under review, however, as there are concerns surrounding timing and whether deep geological disposal is the most appropriate course of action for graphite. An alternative waste management solution is treatment prior to disposal to separate mobile radioactive isotopes such as 3H and 14C from the bulk material, allowing for HAW volume reduction and concentration.Optimisation of an existing thermal treatment process at the Nuclear Graphite Research Group (NGRG) of the University of Manchester has been effected and a detailed review of the uncertainties associated with quantitative determination of radioisotope releases during thermal treatment of irradiated graphite samples has been conducted. Thermal treatment experiments in both an inert atmosphere and 1% oxygen in argon atmosphere have been conducted for temperatures ranging from 600 °C to 800 °C, and durations from 4 to 120 hours, to determine the effects of oxidation time and temperature, and the consequent oxidation characteristics on the release rate of prominent radioisotopes, with a focus on the release of 14C. Lower temperature treatments in an oxidising atmosphere have shown that a preferential release of 14C-enriched graphite can be achieved from the bulk material of Oldbury Magnox reactor irradiated graphite, with evidence demonstrating that this liberated 14C-enriched region is located at the graphite surfaces throughout the porous structure. A large proportion of radiocarbon found in this irradiated graphite, however, is uniformly distributed throughout the bulk material and cannot be selectively oxidised. It is found that prominent metallic radioisotopes such as 60Co are not mobile at these temperatures and remain in the bulk graphite material, inclusive of radioactive caesium which the literature suggests will volatilise.The preliminary results were undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE, and this studentship has been funded by the Engineering and Physical Sciences Research Council (EPSRC). Advisors/Committee Members: JONES, ABBIE A, Mummery, Paul, Jones, Abbie.

Subjects/Keywords: carbon-14; nuclear graphite; irradiated graphite; thermal treatment; waste

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

APA (6^th Edition):

Worth, R. N. (2016). Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:301365

Chicago Manual of Style (16^th Edition):

Worth, Robert Numa. “Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite.” 2016. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:301365.

MLA Handbook (7^th Edition):

Worth, Robert Numa. “Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite.” 2016. Web. 18 Jan 2021.

Vancouver:

Worth RN. Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite. [Internet] [Doctoral dissertation]. University of Manchester; 2016. [cited 2021 Jan 18]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:301365.

Council of Science Editors:

Worth RN. Thermal Treatment of Oldbury Magnox Reactor Irradiated Graphite. [Doctoral Dissertation]. University of Manchester; 2016. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:301365

University of Manchester

8. Mcdermott, Lorraine. Characterisation and chemical treatment of irradiated UK graphite waste.

Degree: PhD, 2012, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/characterisation-and-chemical-treatment-of-irradiated-uk-graphite-waste(11852bdd-2c37-4361-ad0f-0554d05782e2).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.553504

► Once current nuclear reactor operation ceases in the U.K. there will be an estimated 99,000 tonnes of irradiated nuclear graphite waste which may account for… (more)

▼ Once current nuclear reactor operation ceases in the U.K. there will be an estimated 99,000 tonnes of irradiated nuclear graphite waste which may account for up to 30% of any future UK geological ILW disposal facility [1]. In order to make informed decisions of how best to dispose of such large volumes of irradiated graphite (I-graphite) within the UK nuclear programme, it is necessary to understand the nature and migration of isotopes present within the graphite structure. I-graphite has a combination of short and long term isotopes such as 14C, 3H and 36Cl, how these behave prior to and during disposal is of great concern to scientific and regulatory bodies when evaluating present decommissioning options. Various proposed decontamination and immobilisation treatments within the EU Euroatom FP7 CARBOWASTE program have been explored [2, 3]. Experiments have been carried out on UK irradiated British Experimental Pile Zero and Magnox Wylfa graphite in order to remove isotopic content prior to long term storage and to assess the long term leachability of isotopes. Several leaching conditions have been developed to remove 3H and 14C from the irradiated graphite using oxidising and various acidic environments and show mobility of 3H and 14C. Leaching analysis obtained from this research and differences observed under varying leaching conditions will be discussed. Thermal analysis of the samples pre and post leaching has been performed to quantify and validate the 14C and 3H inventory. Finally the research objectives address differences in leachability in the graphite to that of structural and operational variation of the material. Techniques including X-ray Tomography, Scanning Electron Microscopy, Autoradiography and Raman spectroscopy have been examined and show a significant differences in microstructure, isotope distribution and location depending of irradiation history, temperature and graphite source. Ultimately the suitability of the developed chemical treatments will be discussed as whether chemical treatment is a viable option prior to irradiated graphite long term disposal.

Subjects/Keywords: 621.4838; Nuclear Graphite; Chemical Treatment; Leaching; Characterisation

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APA (6^th Edition):

Mcdermott, L. (2012). Characterisation and chemical treatment of irradiated UK graphite waste. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/characterisation-and-chemical-treatment-of-irradiated-uk-graphite-waste(11852bdd-2c37-4361-ad0f-0554d05782e2).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.553504

Chicago Manual of Style (16^th Edition):

Mcdermott, Lorraine. “Characterisation and chemical treatment of irradiated UK graphite waste.” 2012. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/characterisation-and-chemical-treatment-of-irradiated-uk-graphite-waste(11852bdd-2c37-4361-ad0f-0554d05782e2).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.553504.

MLA Handbook (7^th Edition):

Mcdermott, Lorraine. “Characterisation and chemical treatment of irradiated UK graphite waste.” 2012. Web. 18 Jan 2021.

Vancouver:

Mcdermott L. Characterisation and chemical treatment of irradiated UK graphite waste. [Internet] [Doctoral dissertation]. University of Manchester; 2012. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/characterisation-and-chemical-treatment-of-irradiated-uk-graphite-waste(11852bdd-2c37-4361-ad0f-0554d05782e2).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.553504.

Council of Science Editors:

Mcdermott L. Characterisation and chemical treatment of irradiated UK graphite waste. [Doctoral Dissertation]. University of Manchester; 2012. Available from: https://www.research.manchester.ac.uk/portal/en/theses/characterisation-and-chemical-treatment-of-irradiated-uk-graphite-waste(11852bdd-2c37-4361-ad0f-0554d05782e2).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.553504

University of Missouri – Columbia

9. Troy, Raymond Steven. Generation of graphite particles by abrasion and their characterization.

Degree: 2015, University of Missouri – Columbia

URL: http://hdl.handle.net/10355/46903

► Characterization of graphite particles (dust) produced by abrasion that would occur in an operating pebble bed reactor is of interest for reasons of safety, operation,… (more)

Subjects/Keywords: Pebble bed reactors; Graphite; Nuclear power plants

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APA (6^th Edition):

Troy, R. S. (2015). Generation of graphite particles by abrasion and their characterization. (Thesis). University of Missouri – Columbia. Retrieved from http://hdl.handle.net/10355/46903

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):

Troy, Raymond Steven. “Generation of graphite particles by abrasion and their characterization.” 2015. Thesis, University of Missouri – Columbia. Accessed January 18, 2021. http://hdl.handle.net/10355/46903.

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

MLA Handbook (7^th Edition):

Troy, Raymond Steven. “Generation of graphite particles by abrasion and their characterization.” 2015. Web. 18 Jan 2021.

Vancouver:

Troy RS. Generation of graphite particles by abrasion and their characterization. [Internet] [Thesis]. University of Missouri – Columbia; 2015. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/10355/46903.

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

Council of Science Editors:

Troy RS. Generation of graphite particles by abrasion and their characterization. [Thesis]. University of Missouri – Columbia; 2015. Available from: http://hdl.handle.net/10355/46903

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

University of Pretoria

10. [No author]. Purifying coal for the production of nuclear graphite .

Degree: 2008, University of Pretoria

URL: http://upetd.up.ac.za/thesis/available/etd-04212008-124149/

► Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials.… (more)

▼ Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials. It is possible to synthesise graphite and carbon materials from coal. Coal is available in large quantities and could be used for the production of high-purity carbon graphite. However, it contains large quantities of impurities that need to be removed prior to graphitisation/carbonisation. The impurity levels of certain elements in this graphite must be kept at very low levels. Boron, which absorbs neutrons strongly, should be below 500 ppb. Europium and gadolinium, which absorb neutrons and are activated to highly radioactive products, as is cobalt, should be as low as 50 ppb. Lithium transforms to tritium, which leads to the circulating helium becoming radioactive. Other elements, such as calcium, sodium, silicon, thorium and uranium, should not be ignored. The purpose of this study was to lower or remove completely the impurities and trace elements in coal that affect the quality of nuclear-grade graphite. The organic part of Tshikondeni coal was dissolved in a solvent, dimethylformamide (DMF), on addition of sodium hydroxide. The first stage of purification is centrifugation and filtration, which removes most of the impurities. The recovered organic material, known as ‘Refcoal’, may be converted to graphitisable coke. Some elements, significantly boron and cobalt, associate with the organic material in solution and are not sufficiently separated by centrifugation and filtration. Further purification was employed during each process step in the conversion of coal solution into graphite. Different methods of purification were employed in this study. They included chlorination, acid treatment and the ion-exchange or complexation method. Chlorine gas and hexachlorocyclohexane (benzene hexachloride) were used in the chlorination method. Acids such as hydrochloric, hydrofluoric and ascorbic were used in acid treatment. In the ion-exchange method, reagents such as methane, starch, potassium cyanide, ethylene-diaminetetraacetic acid, sodium fluoride, sodium sulphate, ice, glycerol and sodium nitrate were used. All the treated Refcoal was coked at 1 000º C. Pyrolysis was applied in other methods with the aim of volatilising elements that form volatile halides at higher temperatures. Analysis was done for elements such as calcium, cobalt, europium, gadolinium, lithium, sodium, silicon, thorium and uranium, and other elements in the periodic table. Inductively coupled plasma mass spectroscopy and inductively coupled plasma optical emission spectroscopy were used to analyse the concentrations of the trace elements in the coal (treated and untreated) and the coked Refcoal. In inductively coupled plasma mass spectroscopy, microwave digestion and fusion were applied as methods of preparation. However, the instrumentation gave different results for the same sample. The results showed that specific methods work for specific elements. The chlorination method and the… Advisors/Committee Members: Dr D L Morgan (advisor).

Subjects/Keywords: Purification; Nuclear graphite; Fusion reactors; Nuclear materials; UCTD

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

APA (6^th Edition):

author], [. (2008). Purifying coal for the production of nuclear graphite . (Masters Thesis). University of Pretoria. Retrieved from http://upetd.up.ac.za/thesis/available/etd-04212008-124149/

Chicago Manual of Style (16^th Edition):

author], [No. “Purifying coal for the production of nuclear graphite .” 2008. Masters Thesis, University of Pretoria. Accessed January 18, 2021. http://upetd.up.ac.za/thesis/available/etd-04212008-124149/.

MLA Handbook (7^th Edition):

author], [No. “Purifying coal for the production of nuclear graphite .” 2008. Web. 18 Jan 2021.

Vancouver:

author] [. Purifying coal for the production of nuclear graphite . [Internet] [Masters thesis]. University of Pretoria; 2008. [cited 2021 Jan 18]. Available from: http://upetd.up.ac.za/thesis/available/etd-04212008-124149/.

Council of Science Editors:

author] [. Purifying coal for the production of nuclear graphite . [Masters Thesis]. University of Pretoria; 2008. Available from: http://upetd.up.ac.za/thesis/available/etd-04212008-124149/

University of Pretoria

11. Phupheli, Milingoni Robert. Purifying coal for the production of nuclear graphite.

Degree: MSc, Chemistry, 2008, University of Pretoria

URL: http://hdl.handle.net/2263/24051

► Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials.… (more)

▼ Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials. It is possible to synthesise graphite and carbon materials from coal. Coal is available in large quantities and could be used for the production of high-purity carbon graphite. However, it contains large quantities of impurities that need to be removed prior to graphitisation/carbonisation. The impurity levels of certain elements in this graphite must be kept at very low levels. Boron, which absorbs neutrons strongly, should be below 500 ppb. Europium and gadolinium, which absorb neutrons and are activated to highly radioactive products, as is cobalt, should be as low as 50 ppb. Lithium transforms to tritium, which leads to the circulating helium becoming radioactive. Other elements, such as calcium, sodium, silicon, thorium and uranium, should not be ignored. The purpose of this study was to lower or remove completely the impurities and trace elements in coal that affect the quality of nuclear-grade graphite. The organic part of Tshikondeni coal was dissolved in a solvent, dimethylformamide (DMF), on addition of sodium hydroxide. The first stage of purification is centrifugation and filtration, which removes most of the impurities. The recovered organic material, known as ‘Refcoal’, may be converted to graphitisable coke. Some elements, significantly boron and cobalt, associate with the organic material in solution and are not sufficiently separated by centrifugation and filtration. Further purification was employed during each process step in the conversion of coal solution into graphite. Different methods of purification were employed in this study. They included chlorination, acid treatment and the ion-exchange or complexation method. Chlorine gas and hexachlorocyclohexane (benzene hexachloride) were used in the chlorination method. Acids such as hydrochloric, hydrofluoric and ascorbic were used in acid treatment. In the ion-exchange method, reagents such as methane, starch, potassium cyanide, ethylene-diaminetetraacetic acid, sodium fluoride, sodium sulphate, ice, glycerol and sodium nitrate were used. All the treated Refcoal was coked at 1 000º C. Pyrolysis was applied in other methods with the aim of volatilising elements that form volatile halides at higher temperatures. Analysis was done for elements such as calcium, cobalt, europium, gadolinium, lithium, sodium, silicon, thorium and uranium, and other elements in the periodic table. Inductively coupled plasma mass spectroscopy and inductively coupled plasma optical emission spectroscopy were used to analyse the concentrations of the trace elements in the coal (treated and untreated) and the coked Refcoal. In inductively coupled plasma mass spectroscopy, microwave digestion and fusion were applied as methods of preparation. However, the instrumentation gave different results for the same sample. The results showed that specific methods work for specific elements. The chlorination method and the… Advisors/Committee Members: Dr D L Morgan (advisor).

Subjects/Keywords: Purification; Nuclear graphite; Fusion reactors; Nuclear materials; UCTD

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

APA (6^th Edition):

Phupheli, M. (2008). Purifying coal for the production of nuclear graphite. (Masters Thesis). University of Pretoria. Retrieved from http://hdl.handle.net/2263/24051

Chicago Manual of Style (16^th Edition):

Phupheli, Milingoni. “Purifying coal for the production of nuclear graphite.” 2008. Masters Thesis, University of Pretoria. Accessed January 18, 2021. http://hdl.handle.net/2263/24051.

MLA Handbook (7^th Edition):

Phupheli, Milingoni. “Purifying coal for the production of nuclear graphite.” 2008. Web. 18 Jan 2021.

Vancouver:

Phupheli M. Purifying coal for the production of nuclear graphite. [Internet] [Masters thesis]. University of Pretoria; 2008. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/2263/24051.

Council of Science Editors:

Phupheli M. Purifying coal for the production of nuclear graphite. [Masters Thesis]. University of Pretoria; 2008. Available from: http://hdl.handle.net/2263/24051

University of Manchester

12. Bakenne, Adetokunboh Temitope. DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING.

Degree: 2013, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:193369

► Graphite is used within a reactor as a moderator and a reflector material. During fast neutron irradiation, the physical properties and dimensions of nuclear graphite… (more)

▼ Graphite is used within a reactor as a moderator and a reflector material. During fast neutron irradiation, the physical properties and dimensions of nuclear graphite are changed significantly. Graphite shrinkage could lead to disengagement of individual component and loss of core geometry; differential shrinkage in the graphite component could lead to the generation of internal stresses and component failure by cracking. The latter behaviour is complicated by the irradiation induced changes in Young’s modulus and strength. These dimensional and modulus change have been associated with the irradiation-induced closure of many thousands of micro-cracks associated with the graphite crystallites due to crystal dimensional change. Closure of microcracks in nuclear graphite was simulated by external pressure (hydrostatic loading, deviatory stress and dynamic loading) and not by irradiation, whilst Young’s modulus was measured to check if there was any correlation between the two mechanisms. A study of the deformation behaviour of polycrystalline graphite hydrostatically loaded up to 200MPa are reported. Gilsocarbon specimens (isotropic) and Pile Grade A (PGA) specimens are (anisotropic in nature) were investigated. Strain measurements were made in the axial and circumferential directions of cylindrical samples by using strain gauges. Dynamic Young’s modulus was also investigated from the propagation velocity of an ultrasonic wave. Porosity measurements are made to determine the change in the porosity before and after deformation and also their contribution towards the compression and dilatation of graphite under pressure. Graphite crystal orientation during loading was also investigated by using XRD (X-ray diffraction) pole figures. Effective medium models were also investigated to describe the effect of porosity on graphite elastic modulus.All the graphite specimens investigated exhibited non-linear pressure- volumetric strain behaviour in both direction (axial and circumferencial). In most of the experiments, the deformation was closing porosity despite new porosity being generated. Under hydrostatic loading, PGA graphite initially stiff then it became less stiff after a few percent of volume strain and then after about ~20% volumetric strain they stiffen up again, whist Gilsocarbon showed similar behaviour at lower volumetric strain (~10-13%). Gilsocarbon was stiff than PGA; this behaviour is due to the fact that Gilsocarbon has higher density and lower porosity than PGA. During unloading, a large hysteresis was formed. The stressed grains are relieved; the initial closed pores began to reopen. It is suggested that during this stage, the volume of pore re-opening superseded the volume of pores closing, the graphite sample volume almost fully recovered.In the axial compression test, PGA perpendicular to the extrusion direction (PGA-AG) was less stiff than PGA parallel to the extrusion direction (PGA-WG); in the hydrostatic compaction test, the PGA-WG sample deformed more because it had to undergo a less complicated shape… Advisors/Committee Members: HALL, GRAHAM GN, MECKLENBURGH, JULIAN J, Hall, Graham, Mecklenburgh, Julian, Marsden, Barry, Jones, Abbie.

Subjects/Keywords: Nuclear graphite; Irradiation; Virgin- unirradiated; Hydrostatic pressure; Young's moduli; Dynamic moduli; Graphite moderator

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

APA (6^th Edition):

Bakenne, A. T. (2013). DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:193369

Chicago Manual of Style (16^th Edition):

Bakenne, Adetokunboh Temitope. “DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING.” 2013. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:193369.

MLA Handbook (7^th Edition):

Bakenne, Adetokunboh Temitope. “DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING.” 2013. Web. 18 Jan 2021.

Vancouver:

Bakenne AT. DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING. [Internet] [Doctoral dissertation]. University of Manchester; 2013. [cited 2021 Jan 18]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:193369.

Council of Science Editors:

Bakenne AT. DEFORMATION AND MODULUS CHANGES OF NUCLEAR GRAPHITE DUE TO HYDROSTATIC PRESSURE LOADING. [Doctoral Dissertation]. University of Manchester; 2013. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:193369

University of Manchester

13. Bakenne, Adetokunboh. Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading.

Degree: PhD, 2013, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/deformation-and-modulus-changes-of-nuclear-graphite-due-to-hydrostatic-pressure-loading(aa6b8fd6-1c9f-4e71-b0dc-b5150b67223d).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.740252

► Graphite is used within a reactor as a moderator and a reflector material. During fast neutron irradiation, the physical properties and dimensions of nuclear graphite… (more)

▼ Graphite is used within a reactor as a moderator and a reflector material. During fast neutron irradiation, the physical properties and dimensions of nuclear graphite are changed significantly. Graphite shrinkage could lead to disengagement of individual component and loss of core geometry; differential shrinkage in the graphite component could lead to the generation of internal stresses and component failure by cracking. The latter behaviour is complicated by the irradiation induced changes in Young's modulus and strength. These dimensional and modulus change have been associated with the irradiation-induced closure of many thousands of micro-cracks associated with the graphite crystallites due to crystal dimensional change. Closure of microcracks in nuclear graphite was simulated by external pressure (hydrostatic loading, deviatory stress and dynamic loading) and not by irradiation, whilst Young's modulus was measured to check if there was any correlation between the two mechanisms. A study of the deformation behaviour of polycrystalline graphite hydrostatically loaded up to 200MPa are reported. Gilsocarbon specimens (isotropic) and Pile Grade A (PGA) specimens are (anisotropic in nature) were investigated. Strain measurements were made in the axial and circumferential directions of cylindrical samples by using strain gauges. Dynamic Young's modulus was also investigated from the propagation velocity of an ultrasonic wave. Porosity measurements are made to determine the change in the porosity before and after deformation and also their contribution towards the compression and dilatation of graphite under pressure. Graphite crystal orientation during loading was also investigated by using XRD (X-ray diffraction) pole figures. Effective medium models were also investigated to describe the effect of porosity on graphite elastic modulus. All the graphite specimens investigated exhibited non-linear pressure- volumetric strain behaviour in both direction (axial and circumferencial). In most of the experiments, the deformation was closing porosity despite new porosity being generated. Under hydrostatic loading, PGA graphite initially stiff then it became less stiff after a few percent of volume strain and then after about ~20% volumetric strain they stiffen up again, whist Gilsocarbon showed similar behaviour at lower volumetric strain (~10-13%). Gilsocarbon was stiff than PGA; this behaviour is due to the fact that Gilsocarbon has higher density and lower porosity than PGA. During unloading, a large hysteresis was formed. The stressed grains are relieved; the initial closed pores began to reopen. It is suggested that during this stage, the volume of pore re-opening superseded the volume of pores closing, the graphite sample volume almost fully recovered. In the axial compression test, PGA perpendicular to the extrusion direction (PGA-AG) was less stiff than PGA parallel to the extrusion direction (PGA-WG); in the hydrostatic compaction test, the PGA-WG sample deformed more because it had to undergo a less complicated…

Subjects/Keywords: 621.48; Graphite moderator; Dynamic moduli; Young's moduli; Nuclear graphite; Virgin- unirradiated; Irradiation; Hydrostatic pressure

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

APA (6^th Edition):

Bakenne, A. (2013). Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/deformation-and-modulus-changes-of-nuclear-graphite-due-to-hydrostatic-pressure-loading(aa6b8fd6-1c9f-4e71-b0dc-b5150b67223d).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.740252

Chicago Manual of Style (16^th Edition):

Bakenne, Adetokunboh. “Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading.” 2013. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/deformation-and-modulus-changes-of-nuclear-graphite-due-to-hydrostatic-pressure-loading(aa6b8fd6-1c9f-4e71-b0dc-b5150b67223d).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.740252.

MLA Handbook (7^th Edition):

Bakenne, Adetokunboh. “Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading.” 2013. Web. 18 Jan 2021.

Vancouver:

Bakenne A. Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading. [Internet] [Doctoral dissertation]. University of Manchester; 2013. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/deformation-and-modulus-changes-of-nuclear-graphite-due-to-hydrostatic-pressure-loading(aa6b8fd6-1c9f-4e71-b0dc-b5150b67223d).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.740252.

Council of Science Editors:

Bakenne A. Deformation and modulus changes of nuclear graphite due to hydrostatic pressure loading. [Doctoral Dissertation]. University of Manchester; 2013. Available from: https://www.research.manchester.ac.uk/portal/en/theses/deformation-and-modulus-changes-of-nuclear-graphite-due-to-hydrostatic-pressure-loading(aa6b8fd6-1c9f-4e71-b0dc-b5150b67223d).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.740252

North-West University

14. Markgraaff, Renier Francois. Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff .

Degree: 2010, North-West University

URL: http://hdl.handle.net/10394/4813

► NBG–18 graphite is developed by SGL Carbon for the Pebble Bed Modular Reactor Company (PBMR), and is used as the preferred material for the internal… (more)

Subjects/Keywords: NGB-18 nuclear graphite; Hammer impulse excitation; Fatigue; Natural frequency; Relationship

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APA (6^th Edition):

Markgraaff, R. F. (2010). Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff . (Thesis). North-West University. Retrieved from http://hdl.handle.net/10394/4813

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):

Markgraaff, Renier Francois. “Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff .” 2010. Thesis, North-West University. Accessed January 18, 2021. http://hdl.handle.net/10394/4813.

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

MLA Handbook (7^th Edition):

Markgraaff, Renier Francois. “Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff .” 2010. Web. 18 Jan 2021.

Vancouver:

Markgraaff RF. Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff . [Internet] [Thesis]. North-West University; 2010. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/10394/4813.

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

Council of Science Editors:

Markgraaff RF. Relationship between the natural frequencies and fatigue life of NGB–18 graphite / Renier Markgraaff . [Thesis]. North-West University; 2010. Available from: http://hdl.handle.net/10394/4813

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

University of Manchester

15. Turner, Joel David. The performance of a nuclear fuel-matrix material in a sealed CO₂ system.

Degree: PhD, 2013, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566529

► An advanced concept high temperature reactor (HTR) design has been proposed - The ‘U-Battery’, which utilises a unique sealed coolant loop, and is intended to… (more)

▼ An advanced concept high temperature reactor (HTR) design has been proposed - The ‘U-Battery’, which utilises a unique sealed coolant loop, and is intended to operate with minimal human oversight. In order to reduce the need for moving parts within the design, CO2 has been selected as a candidate coolant, potentially allowing a naturally circulated system. HTR fuel is held within a semi-graphitic fuel-matrix material, and this has not previously been tested within a CO2 environment. Graphite in CO2 is subject to two oxidation reactions, one thermally driven and one radiolytically. As such, the oxidation performance of fuel-matrix material has been tested within CO2 at both high temperatures and under ionising radiation within a sealed-system. Performance has been compared to that of the Gilsocarbon and NBG-18 nuclear graphite grades. Gilsocarbon is the primary graphite grade used within the currently operating AGR fleet within the UK, and as such is known to have acceptable oxidation performance under reactor conditions. NBG-18 is a modern graphite grade, and is a candidate material for use within the U-Battery. Virgin characterisation of all materials was performed, including measurements of bulk mass and volume, skeletal volumes and surface areas. High-resolution optical microscopy has also been performed and pore size distributions inferred from digital image analysis. All results were seen to agree well with literature values, and the variation between samples has been quanti- fied and found to be < 10% between samples of Gilsocarbon, and < 4% for samples of fuel-matrix and NBG-18. Thermal performance of fuel-matrix material was observed between 600 °C – 1200 °C and seen to be broadly comparable to that of the nuclear graphite grades tested. NBG-18 showed surprisingly poor performance at 600°C, with an oxidation rate of 3×10−4%/min, approximately ten times faster than Gilsocarbon in similar conditions, and three times faster than fuel-matrix material. The radiolytic oxidation performance of fuel-matrix material and NBG-18 has been observed by irradiating sealed quartz ampoules. Ampoules were pressurised with CO2 prior to irradiation, and the pressure after 30 days of irradiation was measured and seen to fall by 50%. Radiolytic oxidation, and the subsequent radiolysis of the reaction product, CO, was seen to cause significant carbonaceous deposition on the internal surfaces of the ampoule and throughout the samples. Due to the short irradiation times available in the present study, an investigation of the microporosity within irradiated samples has been carried out, using nitrogen adsorption and small-angle neutron scattering (SANS). Pore size distributions produced from SANS show the closure of microporosity within NBG-18, most likely as a result of low-temperature neutron irradiation.As a result of this work, CO2 is no longer a candidate coolant for use with the U-Battery design, due to the rapid deposition observed following irradiation.

Subjects/Keywords: 621.48; Nuclear Graphite; Oxidation; High Temperature Reactor; TRISO

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

APA (6^th Edition):

Turner, J. D. (2013). The performance of a nuclear fuel-matrix material in a sealed CO₂ system. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566529

Chicago Manual of Style (16^th Edition):

Turner, Joel David. “The performance of a nuclear fuel-matrix material in a sealed CO₂ system.” 2013. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566529.

MLA Handbook (7^th Edition):

Turner, Joel David. “The performance of a nuclear fuel-matrix material in a sealed CO₂ system.” 2013. Web. 18 Jan 2021.

Vancouver:

Turner JD. The performance of a nuclear fuel-matrix material in a sealed CO₂ system. [Internet] [Doctoral dissertation]. University of Manchester; 2013. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566529.

Council of Science Editors:

Turner JD. The performance of a nuclear fuel-matrix material in a sealed CO₂ system. [Doctoral Dissertation]. University of Manchester; 2013. Available from: https://www.research.manchester.ac.uk/portal/en/theses/the-performance-of-a-nuclear-fuelmatrix-material-in-a-sealed-co2-system(caaeee7f-9551-485b-b3dc-fe14e75bcc5a).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566529

University of Manchester

16. Luyken, Lewis. Using intercalation to simulate irradiation damage of nuclear graphite.

Degree: PhD, 2012, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/using-intercalation-to-simulate-irradiation-damage-of-nuclear-graphite(85ce1ebb-2981-4807-8cf8-4aedc4acea97).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566532

► This thesis investigates the use of bromine intercalation of graphite as a method to simulate and investigate irradiation damage. In particular this study investigates the… (more)

Subjects/Keywords: 620.1; Graphite; Nuclear; Intercalation; Dimensional Change; Young's Modulus

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APA (6^th Edition):

Luyken, L. (2012). Using intercalation to simulate irradiation damage of nuclear graphite. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/using-intercalation-to-simulate-irradiation-damage-of-nuclear-graphite(85ce1ebb-2981-4807-8cf8-4aedc4acea97).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566532

Chicago Manual of Style (16^th Edition):

Luyken, Lewis. “Using intercalation to simulate irradiation damage of nuclear graphite.” 2012. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/using-intercalation-to-simulate-irradiation-damage-of-nuclear-graphite(85ce1ebb-2981-4807-8cf8-4aedc4acea97).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566532.

MLA Handbook (7^th Edition):

Luyken, Lewis. “Using intercalation to simulate irradiation damage of nuclear graphite.” 2012. Web. 18 Jan 2021.

Vancouver:

Luyken L. Using intercalation to simulate irradiation damage of nuclear graphite. [Internet] [Doctoral dissertation]. University of Manchester; 2012. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/using-intercalation-to-simulate-irradiation-damage-of-nuclear-graphite(85ce1ebb-2981-4807-8cf8-4aedc4acea97).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566532.

Council of Science Editors:

Luyken L. Using intercalation to simulate irradiation damage of nuclear graphite. [Doctoral Dissertation]. University of Manchester; 2012. Available from: https://www.research.manchester.ac.uk/portal/en/theses/using-intercalation-to-simulate-irradiation-damage-of-nuclear-graphite(85ce1ebb-2981-4807-8cf8-4aedc4acea97).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.566532

University of Manchester

17. Fletcher, Adam. Non-destructive testing of the graphite core within an advanced gas-cooled reactor.

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

URL: https://www.research.manchester.ac.uk/portal/en/theses/nondestructive-testing-of-the-graphite-core-within-an-advanced-gascooled-reactor(3ca5c904-6860-46b8-8538-4136cb2aedcd).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.634878

► The aim of this work has been to apply the techniques of non-destructive testing and evaluation to the graphite fuel channel bricks which form the… (more)

▼ The aim of this work has been to apply the techniques of non-destructive testing and evaluation to the graphite fuel channel bricks which form the core of an Advanced Gas-Cooled reactor. Two modes of graphite degradation have been studied: subsurface cracks originating from the keyway corners of the bricks and the reduction in material density caused by radiolytic oxidation. This work has focused on electromagnetic inspection techniques. Brick cracking has been studied using a multi-frequency eddy current technique with the aim of determining quantitative information. In order to accurately control the crack dimensions this work has used radially machined slots as an analogue. Two sensor geometries were studied and it was determined that slots of at least 10 mm through-wall extent could be located. A novel, empirical method of determining the slot size is presented using a brick machined with a series of reference slots. Machined slots originating from a keyway could be sized to within 2 mm using this method. A parametric 3D finite element study was also carried out on this problem. These simulations could distinguish the location of the slots and had some sensitivity to their size, however, the model was found to be overly sensitive to the specific mesh used. Two new contributions to the inverse problem are presented. The first is a minor extension to the usual adjoint problem in which one system now contains a gradiometer. The second is a proposed solution to the ambiguous nature of the inner product required by the sensitivity formulation. This solution has been validated with finite element modelling. Density reduction has been studied via its relationship with electrical conductivity using a technique based on impedance spectroscopy. An inverse eddy current problem has been solved using the regularised Gauss-Newton method to determine the conductivity of the brick over its cross section. The associated forward problem has been solved using the finite element method on a simplified geometry. Tikhonov regularisation has been employed to overcome the ill-posed nature of the inverse problem. This method has been applied to a range of sample and sensor geometries and found to produce excellent results from laboratory data provided the finite element model is well calibrated. Bore or surface conductivity values can be reproduced to better than 1% with the accuracy reducing with distance from the sensor. The sensitivity of the algorithm to the regularisation parameter has been studied using the L-curve method and the effect of two regularisation operators has also been examined. A new method of choosing the regularisation parameter a priori is proposed and tested. Data taken during reactor outages produces physically realistic profiles although the results appear off-set from electrical resistivity values measured using the four-point method. The focus of future work should be to remove this effect which will likely require improvements to the forward model.

Subjects/Keywords: 621.382; nuclear; graphite; electromagnetic; ndt; finite element; modelling; eddy current

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APA (6^th Edition):

Fletcher, A. (2014). Non-destructive testing of the graphite core within an advanced gas-cooled reactor. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/nondestructive-testing-of-the-graphite-core-within-an-advanced-gascooled-reactor(3ca5c904-6860-46b8-8538-4136cb2aedcd).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.634878

Chicago Manual of Style (16^th Edition):

Fletcher, Adam. “Non-destructive testing of the graphite core within an advanced gas-cooled reactor.” 2014. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/nondestructive-testing-of-the-graphite-core-within-an-advanced-gascooled-reactor(3ca5c904-6860-46b8-8538-4136cb2aedcd).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.634878.

MLA Handbook (7^th Edition):

Fletcher, Adam. “Non-destructive testing of the graphite core within an advanced gas-cooled reactor.” 2014. Web. 18 Jan 2021.

Vancouver:

Fletcher A. Non-destructive testing of the graphite core within an advanced gas-cooled reactor. [Internet] [Doctoral dissertation]. University of Manchester; 2014. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/nondestructive-testing-of-the-graphite-core-within-an-advanced-gascooled-reactor(3ca5c904-6860-46b8-8538-4136cb2aedcd).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.634878.

Council of Science Editors:

Fletcher A. Non-destructive testing of the graphite core within an advanced gas-cooled reactor. [Doctoral Dissertation]. University of Manchester; 2014. Available from: https://www.research.manchester.ac.uk/portal/en/theses/nondestructive-testing-of-the-graphite-core-within-an-advanced-gascooled-reactor(3ca5c904-6860-46b8-8538-4136cb2aedcd).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.634878

University of Manchester

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

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

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

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

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

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

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

APA (6^th Edition):

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

Chicago Manual of Style (16^th Edition):

Black, Greg. “Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal.” 2014. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547.

MLA Handbook (7^th Edition):

Black, Greg. “Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal.” 2014. Web. 18 Jan 2021.

Vancouver:

Black G. Irradiated graphite waste : analysis and modelling of radionuclide production with a view to long term disposal. [Internet] [Doctoral dissertation]. University of Manchester; 2014. [cited 2021 Jan 18]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/irradiated-graphite-waste-analysis-and-modelling-of-radionuclide-production-with-a-view-to-long-term-disposal(9993a76a-15c6-4cbe-a4a3-4c0bc88c3134).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.664547.

Council of Science Editors:

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

Kansas State University

19. Gould, Daniel W. Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences.

Degree: PhD, Department of Mechanical and Nuclear Engineering, 2018, Kansas State University

URL: http://hdl.handle.net/2097/39294

► Helium-cooled, graphite moderated reactors have been considered for a future fleet of high temperature and high efficiency nuclear power plants. Nuclear-grade graphite is used in… (more)

▼ Helium-cooled, graphite moderated reactors have been considered for a future fleet of high temperature and high efficiency nuclear power plants. Nuclear-grade graphite is used in these reactors for structural strength, neutron moderation, heat transfer and, within a helium environment, has demonstrated stability at temperatures well above HTGR operating conditions. However, in the case of an air ingress accident, the oxygen introduced into the core can affect the integrity of the fuel graphite matrix. In this work a combination of computational models and mixed effects experiments were used to better understand the air ingress process and its potential effects on the heat removal capabilities of an HTGR design following an air-ingress accident. Contributions were made in the understanding of the air-ingress phenomenon, its potential effects on graphite, and in experimental and computational techniques. The first section of this thesis focuses on experimental and computational studies that were undertaken to further the understanding of the Onset of Natural Convection (ONC) phenomenon expected to occur inside of an HTGR following an air ingress accident. The effects of two newly identified factors on ONC – i.e., the existence of the large volume of stagnate helium in a reactor's upper plenum, and the possibility of an upper head leak – were investigated. Mixed-effects experimental studies were performed to determine the changes induced in nuclear grade graphite exposed to high-temperature, oxidizing flow of varying flow rates. Under all scenarios, the thermal diffusivity of the graphite test samples was shown to increase. Thermal conductivity changes due to oxidation were found to be minor in the tested graphite samples – especially compared to the large drop in thermal conductivity the graphite is expected to experience due to irradiation. Oxidation was also found to increase the graphite's surface roughness and create a thin outer layer of decreased density. The effects of thermal contacts on the passive cooling ability of an HTGR were experimentally investigated. Conduction cool down experiments were performed on assemblies consisting of a number of rods packed into a cylindrical tube. Experimental conditions were then modeled using several different methodologies, including a novel graph laplacian approach, and their results compared to the experimentally obtained temperature data. Although the graph laplacian technique shows great promise, the 2–D Finite Element Model (FEM) provided the best results. Finally, a case study was constructed in which a section of a pebble bed reactor consisting of a number of randomly packed, spherical fuel particles was modeled using the validated FEM technique. Using a discrete elements model, a stable, randomly packed geometry was created to represent the pebble bed. A conduction cool down scenario was modeled and the results from the FEM model were compared to best possible results obtainable from a more traditional, homogeneous 1–D approximation. When the graphite in the bed… Advisors/Committee Members: Hitesh Bindra.

Subjects/Keywords: HTGR; Air Ingress; Onset of Natural Convection; graphite; oxidation; nuclear reactor

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

APA (6^th Edition):

Gould, D. W. (2018). Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences. (Doctoral Dissertation). Kansas State University. Retrieved from http://hdl.handle.net/2097/39294

Chicago Manual of Style (16^th Edition):

Gould, Daniel W. “Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences.” 2018. Doctoral Dissertation, Kansas State University. Accessed January 18, 2021. http://hdl.handle.net/2097/39294.

MLA Handbook (7^th Edition):

Gould, Daniel W. “Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences.” 2018. Web. 18 Jan 2021.

Vancouver:

Gould DW. Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences. [Internet] [Doctoral dissertation]. Kansas State University; 2018. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/2097/39294.

Council of Science Editors:

Gould DW. Air Ingress in HTGRs: the process, eﬀects, and experimental methods relating to its investigation and consequences. [Doctoral Dissertation]. Kansas State University; 2018. Available from: http://hdl.handle.net/2097/39294

University of Texas – Austin

20. Hill, Jason Edward, 1978-. One-dimensional electron systems on graphene edges.

Degree: PhD, Physics, 2007, University of Texas – Austin

URL: http://hdl.handle.net/2152/3797

► In this dissertation several aspects on one-dimensional edge states in grapheme are studied. First, a background in the history and development of graphitic forms is… (more)

Subjects/Keywords: Graphite; Nanostructured materials; Nuclear spin

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APA (6^th Edition):

Hill, Jason Edward, 1. (2007). One-dimensional electron systems on graphene edges. (Doctoral Dissertation). University of Texas – Austin. Retrieved from http://hdl.handle.net/2152/3797

Chicago Manual of Style (16^th Edition):

Hill, Jason Edward, 1978-. “One-dimensional electron systems on graphene edges.” 2007. Doctoral Dissertation, University of Texas – Austin. Accessed January 18, 2021. http://hdl.handle.net/2152/3797.

MLA Handbook (7^th Edition):

Hill, Jason Edward, 1978-. “One-dimensional electron systems on graphene edges.” 2007. Web. 18 Jan 2021.

Vancouver:

Hill, Jason Edward 1. One-dimensional electron systems on graphene edges. [Internet] [Doctoral dissertation]. University of Texas – Austin; 2007. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/2152/3797.

Council of Science Editors:

Hill, Jason Edward 1. One-dimensional electron systems on graphene edges. [Doctoral Dissertation]. University of Texas – Austin; 2007. Available from: http://hdl.handle.net/2152/3797

21. Blondel, Antoine. Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites.

Degree: Docteur es, Physique-Chimie, 2013, Université Claude Bernard – Lyon I

URL: http://www.theses.fr/2013LYO10323

►

Ce travail s'inscrit dans le cadre des études sur la gestion des déchets graphités issus du démantèlement des centrales UNGG (Uranium Naturel Graphite Gaz). Les… (more)

▼

Ce travail s'inscrit dans le cadre des études sur la gestion des déchets graphités issus du démantèlement des centrales UNGG (Uranium Naturel Graphite Gaz). Les déchets graphités contiennent, entre autres, du 36Cl, radionucléide de longue période dimensionnant pour le stockage, car très mobile dans les ouvrages cimentaires et formations argileuses. Or, à ce jour, peu de données sont disponibles sur la localisation du 36Cl et sa spéciation dans le graphite nucléaire irradié. Ces informations sont nécessaires pour le dimensionnement d'un site de stockage adapté aux déchets graphités et leur obtention constitue l'objectif de ma thèse, réalisée en partenariat avec EDF, CEA et Andra. Dans ce contexte, nous avons mis en oeuvre des études expérimentales permettant de simuler et d'évaluer l'impact de la température, de l'irradiation et de la radiolyse à l'interface gaz/graphite sur le comportement en réacteur du 36Cl. La présence de 36Cl dans le graphite est simulée par l'implantation ionique de 37Cl dans des échantillons de graphite nucléaire vierge ce qui permet de nous affranchir des contraintes liées à l'utilisation de graphite irradié. En extrapolant nos résultats aux déchets graphités, nos études sur les effets comparatifs de la température et de l'irradiation montrent que la température tend à appauvrir l'inventaire en chlore 36 des graphites irradiés et que l'irradiation en régime mixte, tel qu'elle a lieu dans les réacteurs UNGG, renforce cet effet d'appauvrissement

This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behavior of 36Cl. This radionuclide is mainly issued through the neutron activation of 35Cl by the reaction 35Cl(n, γ)36Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36Cl, we implanted 37Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature…

Advisors/Committee Members: Moncoffre, Nathalie (thesis director), Toulhoat, Nelly (thesis director).

Subjects/Keywords: Graphite; Nucléaire; Irradiation; Température; Chlore; Comportement; UNGG; Implantation; Graphite; Nuclear; Irradiation; Temperature; Chlorine; Behaviour; UNGG; Implantation; 539.7

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

APA (6^th Edition):

Blondel, A. (2013). Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites. (Doctoral Dissertation). Université Claude Bernard – Lyon I. Retrieved from http://www.theses.fr/2013LYO10323

Chicago Manual of Style (16^th Edition):

Blondel, Antoine. “Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites.” 2013. Doctoral Dissertation, Université Claude Bernard – Lyon I. Accessed January 18, 2021. http://www.theses.fr/2013LYO10323.

MLA Handbook (7^th Edition):

Blondel, Antoine. “Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites.” 2013. Web. 18 Jan 2021.

Vancouver:

Blondel A. Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites. [Internet] [Doctoral dissertation]. Université Claude Bernard – Lyon I; 2013. [cited 2021 Jan 18]. Available from: http://www.theses.fr/2013LYO10323.

Council of Science Editors:

Blondel A. Effets de la température et de l’irradiation sur le comportement du chlore 37 dans le graphite nucléaire : conséquences sur la mobilité du chlore 36 dans les graphites irradiés : Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite : consequences on the mobility of chlorine 36 in irradiated graphites. [Doctoral Dissertation]. Université Claude Bernard – Lyon I; 2013. Available from: http://www.theses.fr/2013LYO10323

22. Vaudey, Claire-Émilie. Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors.

Degree: Docteur es, Physico-chimie, 2010, Université Claude Bernard – Lyon I

URL: http://www.theses.fr/2010LYO10177

►

Ce travail se situe dans le cadre des études sur la gestion des déchets graphites des centrales nucléaires Uranium Naturel Graphite Gaz (UNGG) de première… (more)

▼

Ce travail se situe dans le cadre des études sur la gestion des déchets graphites des centrales nucléaires Uranium Naturel Graphite Gaz (UNGG) de première génération. Leur fonctionnement a généré 23000 tonnes de déchets graphites pour lesquels la loi du 28 juin 2006 prévoit un stockage dédié. La gestion à long terme de ces déchets nécessite de prendre en compte deux radionucléides principaux : le ^14C et le ^36Cl, principaux contributeurs de dose sur le long terme. Afin de consolider les données sur l'inventaire de ces radionucléides et de prévoir leur comportement lors de la resaturation en eau du site de stockage, il est nécessaire de disposer de données liées à leur distribution et à leur spéciation dans le graphite avant stockage. Ce travail a été centré sur l'étude du chlore. Il a eu pour objectif de retracer le comportement du 36Cl dans le graphite nucléaire durant “sa vie” en réacteur et, en particulier d'étudier les effets de la température et de la corrosion radiolytique de manière découplée. Nos résultats permettent de déduire qu'il se produit un relâchement rapide du 36Cl d'environ 20% dès les premières heures de fonctionnement du réacteur. Celui-ci est suivi par un relâchement beaucoup plus lent tout au long de la vie du réacteur. Nous avons identifié la présence de deux fractions distinctes de chlore correspondant à des formes chimiques différentes (n'ayant pas la même stabilité thermique) ou à deux localisations du chlore d'accessibilités différentes. Notre etude montre également que la corrosion radiolytique semble promouvoir le relâchement du chlore et cela quelle que soit la dose d'irradiation. La forme chimique du chlore est majoritairement organique.

This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the Frenche law of June 206. These wastes contain two long-lived radionuclides (^ 14C and ^36Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of ^36Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion indepently. Our results show a rapid release of around 20% ^36Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic.

Advisors/Committee Members: Moncoffre, Nathalie (thesis director), Toulhoat, Nelly (thesis director).

Subjects/Keywords: Chlore 36; Graphite nucléaire; Déchets nucléaires; Corrosion radiolytique; Speciation; Réacteurs UNGG; Chlorinee 36; Nuclear graphite; Nuclear wastes; Radiolytic corrosion; Speciation; UNGG reactors

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

APA (6^th Edition):

Vaudey, C. (2010). Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors. (Doctoral Dissertation). Université Claude Bernard – Lyon I. Retrieved from http://www.theses.fr/2010LYO10177

Chicago Manual of Style (16^th Edition):

Vaudey, Claire-Émilie. “Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors.” 2010. Doctoral Dissertation, Université Claude Bernard – Lyon I. Accessed January 18, 2021. http://www.theses.fr/2010LYO10177.

MLA Handbook (7^th Edition):

Vaudey, Claire-Émilie. “Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors.” 2010. Web. 18 Jan 2021.

Vancouver:

Vaudey C. Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors. [Internet] [Doctoral dissertation]. Université Claude Bernard – Lyon I; 2010. [cited 2021 Jan 18]. Available from: http://www.theses.fr/2010LYO10177.

Council of Science Editors:

Vaudey C. Effets de la température et de la corrosion radiolytique sur le comportement du chlore dans le graphite nucléaire : conséquences pour le stockage des graphites irradiés des réacteurs UNGG : Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite : consequences for the disposabl of irradiated graphite from UNGG reactors. [Doctoral Dissertation]. Université Claude Bernard – Lyon I; 2010. Available from: http://www.theses.fr/2010LYO10177

23. Silbermann, Gwennaelle. Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming.

Degree: Docteur es, Physico-chimie des matériaux, 2013, Université Claude Bernard – Lyon I

URL: http://www.theses.fr/2013LYO10168

►

Le démantèlement des réacteurs Uranium Naturel Graphite Gaz génèrera en France environ 23 000 tonnes de déchets radioactifs graphités. La gestion appropriée de ces déchets… (more)

▼

Le démantèlement des réacteurs Uranium Naturel Graphite Gaz génèrera en France environ 23 000 tonnes de déchets radioactifs graphités. La gestion appropriée de ces déchets nécessite de déterminer leur inventaire radiologique et de disposer de données fiables sur la localisation et la spéciation des radionucléides (RN). Le 14C a été identifié comme RN d'intérêt pour le stockage en raison de son inventaire initial important et du risque de présence d'une fraction organique mobile dans l'environnement, lors de la phase de stockage. A ce titre, l'objectif de cette thèse CIFRE, réalisée en partenariat avec EDF, est de mettre en œuvre des études expérimentales permettant de simuler et d'évaluer l'impact de la température, de l'irradiation et de la corrosion radiolytique du graphite sur le comportement migratoire en réacteur du 14C et de son précurseur azote. Les données ainsi acquises sont intégrées dans la deuxième partie de ce travail consacrée à l'étude d'un procédé de décontamination thermique du graphite en présence de vapeur d'eau. La démarche expérimentale consiste à simuler respectivement la présence de 14C et de 14N par implantation ionique de 13C et d'azote (14N ou 15N) dans un graphite de rondin SLA2 vierge. Cette étude montre que dans la gamme de températures du graphite en réacteur (100 - 500°C) et en absence de corrosion radiolytique, le 13C est stable thermiquement quel que soit l'état de structure du graphite. En revanche, les expériences d'irradiation du graphite chauffé à 500°C au contact d'un gaz représentatif du caloporteur radiolysé montrent le rôle synergique joué par les espèces oxydantes et l'endommagement du graphite favorisant la mobilité du 13C par gazéification des surfaces et/ou oxydation sélective du 13C plus faiblement lié. En ce qui concerne l'azote constitutif, il a tout d'abord été démontré que sa concentration en surface atteint plusieurs centaines de ppm (< 500 ppm at.) et décroît en profondeur jusqu'à environ 160 ppm at.. Contrairement au 13C implanté, l'azote implanté migre à 500°C lorsque le graphite est fortement déstructuré (environ 8 dpa) alors qu'il reste stable pour un taux de déstructuration moindre (0,14 dpa). Les expériences montrent également le rôle synergique des excitations électroniques et de la température qui accélèrent le transport de l'azote vers la surface du graphite. Cette migration de l'azote semble se faire sous forme moléculaire d'espèces C-N, C=N voire C N. Après huit heures d'irradiation ces espèces ne sont toutefois pas ou peu relâchées et restent bloquées à la surface. L'étude du procédé de décontamination thermique en présence de vapeur d'eau a nécessité la mise en place d'un dispositif de thermogravimétrie couplé à un générateur de vapeur d'eau ainsi que l'optimisation des paramètres de l'étude. Les influences de la température (700°C et 900°C) et de l'humidité relative (50 % HR et 90 % HR) ont été testées à un débit de gaz humide fixe de 50 mL/min. Dans ces conditions, l'oxydation sélective du carbone implanté a été confirmée

The dismantling of UNGG…

Advisors/Committee Members: Moncoffre, Nathalie (thesis director), Toulhoat, Nelly (thesis director).

Subjects/Keywords: Carbone 14; Graphite nucléaire; Réacteur UNGG; Démantèlement; Déchets nucléaires; Corrosion radiolytique; Décontamination; Carbon 14; Nuclear graphite; UNGG reactors; Dismantling; Nuclear wastes; Radiolytic; Corrosion; 530

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

APA (6^th Edition):

Silbermann, G. (2013). Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming. (Doctoral Dissertation). Université Claude Bernard – Lyon I. Retrieved from http://www.theses.fr/2013LYO10168

Chicago Manual of Style (16^th Edition):

Silbermann, Gwennaelle. “Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming.” 2013. Doctoral Dissertation, Université Claude Bernard – Lyon I. Accessed January 18, 2021. http://www.theses.fr/2013LYO10168.

MLA Handbook (7^th Edition):

Silbermann, Gwennaelle. “Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming.” 2013. Web. 18 Jan 2021.

Vancouver:

Silbermann G. Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming. [Internet] [Doctoral dissertation]. Université Claude Bernard – Lyon I; 2013. [cited 2021 Jan 18]. Available from: http://www.theses.fr/2013LYO10168.

Council of Science Editors:

Silbermann G. Effets de la température et de l'irradiation sur le comportement du 14C et de son précurseur 14N dans le graphite nucléaire. Étude de la décontamination thermique du graphite en présence de vapeur d'eau : Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming. [Doctoral Dissertation]. Université Claude Bernard – Lyon I; 2013. Available from: http://www.theses.fr/2013LYO10168

Texas A&M University

24. Rauch, Eric B. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.

Degree: MS, Nuclear Engineering, 2010, Texas A&M University

URL: http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739

► Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with… (more)

▼ Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable. Advisors/Committee Members: Charlton, William S. (advisor), Tsvetkov, Pavel V. (committee member), Sprecher, Christopher M. (committee member).

Subjects/Keywords: graphite moderated reactor; nuclear safeguards; DPRK

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

APA (6^th Edition):

Rauch, E. B. (2010). Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739

Chicago Manual of Style (16^th Edition):

Rauch, Eric B. “Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.” 2010. Masters Thesis, Texas A&M University. Accessed January 18, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739.

MLA Handbook (7^th Edition):

Rauch, Eric B. “Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.” 2010. Web. 18 Jan 2021.

Vancouver:

Rauch EB. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. [Internet] [Masters thesis]. Texas A&M University; 2010. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739.

Council of Science Editors:

Rauch EB. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. [Masters Thesis]. Texas A&M University; 2010. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739

University of Manchester

25. Crump, Timothy. Modelling dynamic cracking of graphite.

Degree: 2017, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:312603

► Advances in dynamic fracture modelling have become more frequent due to increases in computer speed, meaning that its application to industrial problems has become viable.… (more)

▼ Advances in dynamic fracture modelling have become more frequent due to increases in computer speed, meaning that its application to industrial problems has become viable. From this, the author has reviewed current literature in terms of graphite material properties, structural dynamics, fracture mechanics and modelling methodologies to be able to address operational issues related to the ageing of Advanced Gas-cooled Reactor (AGR) cores. In particular, the experimentally observed Prompt Secondary Cracking (PSC) of graphite moderator bricks which has yet to be observed within operational reactors, with the objective of supporting their plant life extension. A method known as eXtended Finite Element Method with Cohesive Zones (XCZM) was developed within Code_Aster open-source FEM software. This enabled the incorporation of velocity toughening, irradiation-induced material degradation effects and multiple 3D dynamic crack initiations, propagations and arrests into a single model, which covers the major known attributes of the PSC mechanism. Whilst developing XCZM, several publications were produced. This started with first demonstrating XCZM's ability to model the PSC mechanism in 2D and consequently that methane holes have a noticeable effect on crack propagation speeds. Following on from this, XCZM was benchmarked in 2D against literature experiments and available model data which consequently highlighted that velocity toughening was an integral feature in producing energetically correct fracture speeds. Leading on from this, XCZM was taken into 3D and demonstrated that it produced experimentally observed bifurcation angle from a literature example. This meant that when a 3D graphite brick was modelled that the crack profile was equivalent to an accepted quasi-static profile. As a consequence of this validation, the XCZM approach was able to model PSC and give insight into features that could not be investigated previously including: finer-scale heterogeneous effects on a dynamic crack profile, comparison between Primary and Secondary crack profiles and also, 3D crack interaction with a methane hole, including insight into possible crack arrest. XCZM was shown to improve upon previous 2D models of experiments that showed the plausibility of PSC; this was achieved by eliminating the need for user intervention and also incorporation of irradiation damage effects through User-defined Material properties (UMAT). Finally, while applying XCZM to a full-scale 3D graphite brick including reactor effects, it was shown that PSC is likely to occur under LEFM assumptions and that the Secondary crack initiates before the Primary crack arrests axially meaning that modal analysis would not be able to fully model PSC. Advisors/Committee Members: JIVKOV, ANDREY AP, Mummery, Paul, Jivkov, Andrey.

Subjects/Keywords: graphite; dynamic fracture; XFEM; Cohesive zone; Nuclear; XCZM; Structural integrity; meso; Prompt Secondary Cracking; AGR

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

APA (6^th Edition):

Crump, T. (2017). Modelling dynamic cracking of graphite. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:312603

Chicago Manual of Style (16^th Edition):

Crump, Timothy. “Modelling dynamic cracking of graphite.” 2017. Doctoral Dissertation, University of Manchester. Accessed January 18, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:312603.

MLA Handbook (7^th Edition):

Crump, Timothy. “Modelling dynamic cracking of graphite.” 2017. Web. 18 Jan 2021.

Vancouver:

Crump T. Modelling dynamic cracking of graphite. [Internet] [Doctoral dissertation]. University of Manchester; 2017. [cited 2021 Jan 18]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:312603.

Council of Science Editors:

Crump T. Modelling dynamic cracking of graphite. [Doctoral Dissertation]. University of Manchester; 2017. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:312603

University of Missouri – Columbia

26. Lee, Yoonjo J., 1983-. Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios.

Degree: 2016, University of Missouri – Columbia

URL: https://doi.org/10.32469/10355/59845

► One of the most severe accidents anticipated for the Very High Temperature Reactor (VHTR) is an air ingress accident caused by a pipe break, where… (more)

Subjects/Keywords: Nuclear reactors – Safety measures – Research; Graphite – Thermal properties; Oxidation; Materials at high temperatures – Testing.

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

APA (6^th Edition):

Lee, Yoonjo J., 1. (2016). Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios. (Thesis). University of Missouri – Columbia. Retrieved from https://doi.org/10.32469/10355/59845

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):

Lee, Yoonjo J., 1983-. “Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios.” 2016. Thesis, University of Missouri – Columbia. Accessed January 18, 2021. https://doi.org/10.32469/10355/59845.

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

MLA Handbook (7^th Edition):

Lee, Yoonjo J., 1983-. “Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios.” 2016. Web. 18 Jan 2021.

Vancouver:

Lee, Yoonjo J. 1. Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios. [Internet] [Thesis]. University of Missouri – Columbia; 2016. [cited 2021 Jan 18]. Available from: https://doi.org/10.32469/10355/59845.

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

Council of Science Editors:

Lee, Yoonjo J. 1. Oxidation of nuclear- and matrix-grade graphite for VHTR air ingress accident scenarios: Oxidation of nuclear- and matrix-grade graphite for Very High Temperature Reactor air ingress accident scenarios. [Thesis]. University of Missouri – Columbia; 2016. Available from: https://doi.org/10.32469/10355/59845

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

University of Southern California

27. White, Kevin Spencer. SQUID-based CW NMR system for measuring the magnetization of ³He films.

Degree: MS, Physics, 2012, University of Southern California

URL: http://digitallibrary.usc.edu/cdm/compoundobject/collection/p15799coll3/id/120082/rec/6026

► This thesis describes the design and construction of a SQUID-based CW NMR system together with its application in a study of the two dimensional magnetism… (more)

Subjects/Keywords: helium; low-temperature; magnetization; ferromagnetism; quantum; nuclear spin; graphite; two-dimensional; NMR; spectroscopy

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

APA (6^th Edition):

White, K. S. (2012). SQUID-based CW NMR system for measuring the magnetization of ³He films. (Masters Thesis). University of Southern California. Retrieved from http://digitallibrary.usc.edu/cdm/compoundobject/collection/p15799coll3/id/120082/rec/6026

Chicago Manual of Style (16^th Edition):

White, Kevin Spencer. “SQUID-based CW NMR system for measuring the magnetization of ³He films.” 2012. Masters Thesis, University of Southern California. Accessed January 18, 2021. http://digitallibrary.usc.edu/cdm/compoundobject/collection/p15799coll3/id/120082/rec/6026.

MLA Handbook (7^th Edition):

White, Kevin Spencer. “SQUID-based CW NMR system for measuring the magnetization of ³He films.” 2012. Web. 18 Jan 2021.

Vancouver:

White KS. SQUID-based CW NMR system for measuring the magnetization of ³He films. [Internet] [Masters thesis]. University of Southern California; 2012. [cited 2021 Jan 18]. Available from: http://digitallibrary.usc.edu/cdm/compoundobject/collection/p15799coll3/id/120082/rec/6026.

Council of Science Editors:

White KS. SQUID-based CW NMR system for measuring the magnetization of ³He films. [Masters Thesis]. University of Southern California; 2012. Available from: http://digitallibrary.usc.edu/cdm/compoundobject/collection/p15799coll3/id/120082/rec/6026

University of Pretoria

28. Magampa, Philemon Podile. Properties of graphitic composites.

Degree: Chemistry, 2013, University of Pretoria

URL: http://hdl.handle.net/2263/40244

► The Pebble Bed Modular Reactor (PBMR) is a high temperature graphite-moderated nuclear reactor that uses helium as a coolant. The triple coated (TRISO) particles contain… (more)

▼ The Pebble Bed Modular Reactor (PBMR) is a high temperature graphite-moderated nuclear reactor that uses helium as a coolant. The triple coated (TRISO) particles contain enriched uranium oxide fuel which is coated with layers of various forms of pyrolytic carbon and silicon carbide. The TRISO particles are further embedded in the matrix of spherical graphite pebbles. The graphite matrix is a composite moulded from a compound containing natural flake graphite (64 wt.%), synthetic graphite (16 wt.%) and a phenolic resin binder (20 wt.%) heated to 1800 °C in inert atmosphere. The graphitic composite provides structural integrity, encasement and act as a moderator material. In this work, low density model graphite composites similar to those used in nuclear applications as encasement material in fuel pebbles were made by uniaxial cold compression moulding. The graphitic composites contained various ratios of natural flake graphite and synthetic graphite at fixed phenolic novolac resin binder content of 20 wt.% (green state). The fabrication process employed entails mixing the graphite powders, followed by addition of methanol phenolic resin solution to the graphite powder mix, drying, grinding, milling and sieving; and finally compression moulding in a stainless steel die at 13 MPa using a hydraulic press. The green moulded disc specimens were then carbonized at 900 °C in nitrogen atmosphere to remove volatiles followed by annealing at 1800 °C in helium atmosphere. The annealing step diminishes structural defects and result in densification of the composites. The microstructure of fabricated graphitic composites was characterized using various techniques. Particle Size Distributions determined using Laser diffraction showed that the inclusion of the binder leads to agglomeration. The composite powders had larger mean particle sizes than the raw graphite powders showing the binding effect of the novolac phenolic resin. X-ray diffraction studies showed that the graphitic composites had a hexagonal crystal structure after annealing. Raman spectroscopy revealed the presence of the structurally disordered phase derived from the resin carbon (indicated by the pronounced D-band in the Raman spectra). XRD and Raman observations were consistent with literature and gave results supporting existing knowledge base. Optical microscopy revealed a flake-like microstructure for composites containing natural graphite and needle-coke like particles for composites containing mainly synthetic graphite. Optical microscopy confirmed that the effect of the manufacturing route employed here was to align the particles in the direction perpendicular to the compression moulding direction. As a result, the graphitic composites exhibited anisotropic property behavior. The bulk density of the composites increased with the increase in the natural graphite content due to compactability of natural flakes in the manufacturing route. Thermogravimetric analysis studies on the composites showed that they were stable in air to 650 °C. Composites containing… Advisors/Committee Members: Manyala, Ncholu I. (advisor), Focke, Walter Wilhelm (coadvisor).

Subjects/Keywords: Properties; Pebble Bed Modular Reactor (PBMR); Graphite-moderated nuclear reactor; TRISO; UCTD

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

APA (6^th Edition):

Magampa, P. P. (2013). Properties of graphitic composites. (Doctoral Dissertation). University of Pretoria. Retrieved from http://hdl.handle.net/2263/40244

Chicago Manual of Style (16^th Edition):

Magampa, Philemon Podile. “Properties of graphitic composites.” 2013. Doctoral Dissertation, University of Pretoria. Accessed January 18, 2021. http://hdl.handle.net/2263/40244.

MLA Handbook (7^th Edition):

Magampa, Philemon Podile. “Properties of graphitic composites.” 2013. Web. 18 Jan 2021.

Vancouver:

Magampa PP. Properties of graphitic composites. [Internet] [Doctoral dissertation]. University of Pretoria; 2013. [cited 2021 Jan 18]. Available from: http://hdl.handle.net/2263/40244.

Council of Science Editors:

Magampa PP. Properties of graphitic composites. [Doctoral Dissertation]. University of Pretoria; 2013. Available from: http://hdl.handle.net/2263/40244

University of Florida

29. Woods, Hugh Wilroy, 1941-. Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range.

Degree: 1969, University of Florida

URL: https://ufdc.ufl.edu/AA00068270

Subjects/Keywords: Graphite – Research; Graphite – Thermodynamics; Nuclear reactors – Materials; Nuclear Engineering Sciences thesis Ph. D; Graphite – Research ( fast ); Nuclear reactors – Materials ( fast )

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

APA (6^th Edition):

Woods, Hugh Wilroy, 1. (1969). Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range. (Thesis). University of Florida. Retrieved from https://ufdc.ufl.edu/AA00068270

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):

Woods, Hugh Wilroy, 1941-. “Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range.” 1969. Thesis, University of Florida. Accessed January 18, 2021. https://ufdc.ufl.edu/AA00068270.

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

MLA Handbook (7^th Edition):

Woods, Hugh Wilroy, 1941-. “Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range.” 1969. Web. 18 Jan 2021.

Vancouver:

Woods, Hugh Wilroy 1. Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range. [Internet] [Thesis]. University of Florida; 1969. [cited 2021 Jan 18]. Available from: https://ufdc.ufl.edu/AA00068270.

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

Council of Science Editors:

Woods, Hugh Wilroy 1. Theoretical and experimental investigation of the graphite phase diagram and graphite properties in the liquid core nuclear rocket pressure-temperature range. [Thesis]. University of Florida; 1969. Available from: https://ufdc.ufl.edu/AA00068270

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

30. Le Guillou, Maël. Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste.

Degree: Docteur es, Physique des matériaux, 2014, Université Claude Bernard – Lyon I

URL: http://www.theses.fr/2014LYO10227

►

En France, 23 000 tonnes de graphites irradiés générés par le démantèlement des réacteurs nucléaires de première génération Uranium Naturel-Graphite-Gaz (UNGG) sont en attente d'une… (more)

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En France, 23 000 tonnes de graphites irradiés générés par le démantèlement des réacteurs nucléaires de première génération Uranium Naturel-Graphite-Gaz (UNGG) sont en attente d'une solution de gestion à long terme. Cette thèse porte sur le comportement du tritium, l'un des principaux contributeurs à l'inventaire radiologique des graphites à l'arrêt des réacteurs. Afin d'anticiper des rejets de tritium lors du démantèlement ou de la gestion des déchets, il est indispensable d'obtenir des données sur sa migration, sa localisation et son inventaire. Notre étude repose sur la simulation du tritium par implantation de l'ordre de 3 % at. de deutérium jusqu'à environ 3 μm dans un graphite nucléaire vierge. Celui-ci a ensuite subi des recuits jusqu'à 300 h et 1300 ° C sous atmosphère inerte, gaz caloporteur UNGG et gaz humide, dans le but de reproduire des conditions proches de celles rencontrées en réacteur et lors des opérations de gestion des déchets. Les profils et la répartition spatiale du deutérium ont été analysés via la réaction nucléaire 2H(3He,p)4He. Les principaux résultats montrent un relâchement thermique du deutérium se produisant selon trois régimes contrôlés par le dépiégeage de sites superficiels ou interstitiels. L'extrapolation des données au cas du tritium tend à montrer que son relâchement thermique en réacteur pourrait avoir été inférieur à 30 % et localisé à proximité des surfaces libres du graphite. L'essentiel de l'inventaire en tritium à l'arrêt des réacteurs serait retenu en profondeur dans les graphites irradiés, dont la décontamination nécessiterait alors des températures supérieures à 1300 °C, et serait plus efficace sous gaz inerte que sous gaz humide

In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 °C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be…

Advisors/Committee Members: Moncoffre, Nathalie (thesis director), Toulhoat, Nelly (thesis director).

Subjects/Keywords: UNGG; Démantèlement; Déchets radioactifs; Graphite; Tritium; Deutérium; Traitements thermiques; Inventaire radiologique; GCR; Dismantling; Nuclear waste; Graphite; Tritium; Deuterium; Heat treatments; Radiological inventory; 539.7

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APA (6^th Edition):

Le Guillou, M. (2014). Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste. (Doctoral Dissertation). Université Claude Bernard – Lyon I. Retrieved from http://www.theses.fr/2014LYO10227

Chicago Manual of Style (16^th Edition):

Le Guillou, Maël. “Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste.” 2014. Doctoral Dissertation, Université Claude Bernard – Lyon I. Accessed January 18, 2021. http://www.theses.fr/2014LYO10227.

MLA Handbook (7^th Edition):

Le Guillou, Maël. “Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste.” 2014. Web. 18 Jan 2021.

Vancouver:

Le Guillou M. Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste. [Internet] [Doctoral dissertation]. Université Claude Bernard – Lyon I; 2014. [cited 2021 Jan 18]. Available from: http://www.theses.fr/2014LYO10227.

Council of Science Editors:

Le Guillou M. Migration du deutérium dans le graphite nucléaire : conséquences sur le comportement du tritium en réacteur UNGG et sur la décontamination des graphites irradiés : Deuterium migration in nuclear graphite : consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste. [Doctoral Dissertation]. Université Claude Bernard – Lyon I; 2014. Available from: http://www.theses.fr/2014LYO10227

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