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Relevant bibliographies by topics / Htr fuel

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Academic literature on the topic 'Htr fuel'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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Journal articles on the topic "Htr fuel"

1

Marmier, Alain, Michael A. Fütterer, Kamil Tuček, Jim C. Kuijper, Jaap Oppe, Biser Petrov, Jérôme Jonnet, Jan Leen Kloosterman, and Brian Boer. "Fuel Cycle Investigation for Wallpaper-Type HTR Fuel." Nuclear Technology 181, no. 2 (February 2013): 317–30. http://dx.doi.org/10.13182/nt13-a15786.

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Nabielek, H., W. Kühnlein, W. Schenk, W. Heit, A. Christ, and H. Ragoss. "Development of advanced HTR fuel elements." Nuclear Engineering and Design 121, no. 2 (July 1990): 199–210. http://dx.doi.org/10.1016/0029-5493(90)90105-7.

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Zhang, Hai Quan, Xin Wang, Hong Ke Li, Jun Feng Nie, and Ji Guo Liu. "Design and Engineering Verification of HTR-PM Fuel Handling." Advanced Materials Research 621 (December 2012): 317–25. http://dx.doi.org/10.4028/www.scientific.net/amr.621.317.

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Abstract. HTR-PM is a twin-reactor structure pebble bed modular reactor. With the gather-scatter fuel handling system (FHS), handling and circulating function of twin reactors’ fuel elements are performed under a non-shutdown continuous condition. Relying on separate pipeline system of sphere fuel main circulation, FHS achieved automatic operation in the structural model by sphere fuel’s gravity flowing and pneumatic conveying in the twin reactors. The FHS adopted the international experience at design and operation of similar systems, especially based on that of HTR-10. However, some key components and technologies were improved so that fuel handling of HTR-PM becomes more reliable. All of the improved components and technologies will be tested in a full-scale hot testing facility, and some of them were verified and validated with the help of separated cold testing facilities. The functions and design of HTR-PM FHS is introduced in this paper. Design and engineering test of the FHS in HTR-PM demonstration power plant are reviewed.

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Wijaya, Rokhmadi, Bebeh Wahid Nuryadin, Khotib Maulani, and Topan Setiadipura. "CALCULATION OF PROBABILITY OF TRISO PARTICLE FAILURE USING TIMCOAT AND PEBBED CODE." SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir 24, no. 1 (April 30, 2020): 17. http://dx.doi.org/10.17146/sigma.2020.24.1.5786.

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CALCULATION OF PROBABILITY OF TRISO PARTICLE FAILURE USING TIMCOAT AND PEBBED CODE. The calculation of the failure probability for fuel particles (TRISO) in the HTGR type reactor has been successfully carried out. This study aimed to estimate the failure probability of the fuel particles in the HTR-10 and HTR-PM, as well as to analyze the fuels of those reactors by varying the SiC thickness. The initial layer thickness of SiC in the HTR-10 and HTR-PM is 35 µm. The PEBBED code was used to simulate calculations resulting in the power distribution data, which is then compared with the results from the TIMCOAT simulation process. The TIMCOAT simulation calculation results, which are based on the SiC thickness variation, showed that the thickness failure is smaller if applied to the HTR-10 and HTR-PM. Based on the comparison between the two reactors, the failure probability of HTR-PM fuel particle has the value smaller than that of the HTR-10 with the difference of 10-5 .Keywords: failure particle, HTR-10, HTR-PM, TRISO, TIMCOAT.

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Marmier, A., M. A. Fütterer, K. Tuček, Han de Haas, Jim C. Kuijper, and Jan Leen Kloosterman. "Revisiting the concept of HTR wallpaper fuel." Nuclear Engineering and Design 240, no. 10 (October 2010): 2485–92. http://dx.doi.org/10.1016/j.nucengdes.2010.02.043.

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de Groot, Sander, Pierre Guillermier, Kazuhiro Sawa, Jean-Michel Escleine, Shohei Ueta, Virginie Basini, Klaas Bakker, Young-Woo Lee, Marc Perez, and Bong-Goo Kim. "HTR fuel coating separate effect test PYCASSO." Nuclear Engineering and Design 240, no. 10 (October 2010): 2392–400. http://dx.doi.org/10.1016/j.nucengdes.2010.05.052.

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Helary, D., O. Dugne, X. Bourrat, P. H. Jouneau, and F. Cellier. "EBSD investigation of SiC for HTR fuel particles." Journal of Nuclear Materials 350, no. 3 (May 2006): 332–35. http://dx.doi.org/10.1016/j.jnucmat.2006.01.010.

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Brähler, Georg, Markus Hartung, Johannes Fachinger, Karl-Heinz Grosse, and Richard Seemann. "Improvements in the fabrication of HTR fuel elements." Nuclear Engineering and Design 251 (October 2012): 239–43. http://dx.doi.org/10.1016/j.nucengdes.2011.10.036.

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9

FU, Xiaoming, Tongxiang LIANG, Yaping TANG, Zhichang XU, and Chunhe TANG. "Preparation of UO2Kernel for HTR-10 Fuel Element." Journal of Nuclear Science and Technology 41, no. 9 (September 2004): 943–48. http://dx.doi.org/10.1080/18811248.2004.9715568.

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10

Sembiring, Tagor Malem, and Pungky Ayu Artiani. "SUBCRITICALITY ANALYSIS OF HTR-10 SPENT FUEL CASK MODEL FOR THE 10 MW HTR INDONESIAN EXPERIMENTAL POWER REACTOR." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 20, no. 3 (October 31, 2018): 151. http://dx.doi.org/10.17146/tdm.2018.20.3.4630.

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The 10 MW HTR Indonesian Experimental Power Reactor (RDE reactor) is designed identical with the HTR-10 in China, conceptually. However, the review results showed that the spent fuel cask model which is used between two reactors is fully different, such as size and capacity. The proposed cask model in RDE reactor can hold 15 times more fuel pebbles than HTR-10 has. This research activities deal with the subcriticality analysis for the spent fuel cask of RDE reactor if using the HTR-10 cask model. The subcriticality condition is designed to meet the limit of safety value. The objective of this research is to determine the subcriticality value in the normal and accident events for the spent fuel cask when it is in the reactor building and the spent fuel cask room. All calculations were carried out by MCNP6.1 code. The selected external events are the water ingress (reactor room), water flood and the combination event of water flood and earthquake. The calculation results showed that the maximum value of keff (3σ) are 0.47510 and 0.19214 for the cask in the reactor building and in the spent fuel cask room, respectively. This value is far from the limit value of 0.95. The calculation results showed that the spent fuel cask are in the safe condition eventhough in the worst combination events, the cask is flooded and earthquake. The HTR-10 spent fuel cask can be proposed as an alternative for the RDE reactor to get an efficient reactor building.Keywords: spent pebble fuel element, HTGR, subcriticality, MCNP6.1, RDE reactor ANALISIS SUBKRITIKALITAS PENYIMPAN BAHAN BAKAR BEKAS MODEL CASK REAKTOR HTR-10 UNTUK REAKTOR DAYA EKSPERIMENTAL 10 MW TERMAL. Reaktor Daya Eksperimental (RDE) secara konseptual didesain identik dengan reaktor HTR-10 di Tiongkok. Meskipun demikian, terdapat perbedaan yang signifikan untuk desain konseptual cask penyimpan bahan bakar bekas di kedua reaktor seperti dimensi dan kapasitas. Kegiatan penelitian ini berkaitan dengan analisis subkritikalitas cask penyimpan elemen bahan bakar bekas tipe pebble di RDE jika menggunakan model cask yang dipakai di HTR-10. Kondisi sub-kritikalitas didesain memenuhi nilai batas keselamatan. Tujuan penelitian adalah menentukan nilai subkritikalitas dalam keadaan normal atau kondisi kecelakaan di gedung reaktor dan di gudang penyimpan bahan bakar bekas. Perhitungan dilakukan dengan paket program MCNP6.1. Kejadian kecelakaan yang dipilih adalah masuknya air ke dalam cask, cask terendam air dan kombinasi cask terendam air dan kejadian gempa. Hasil perhitungan menunjukkan bahwa nilai maksimum keff (3σ) untuk cask di gedung reaktor dan di gudang penyimpan bahan bakar bekas masing-masing adalah 0,47510 dan 0,19214. Nilai ini masih jauh dari batas 0,95. Hasil perhitungan menunjukkan bahwa cask penyimpan bahan bakar bekas tetap dalam keadaan selamat meski terjadi kombinasi 2 kejadian eksternal.Kata kunci: elemen bahan bakar bekas tipe pebble, HTGR, subkritikalitas, MCNP6.1, RDE

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Dissertations / Theses on the topic "Htr fuel"

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Van, der Merwe Jacobus Johannes. "Modelling silver transport in spherical HTR fuel." Thesis, Pretoria : [s. n.], 2009. http://upetd.up.ac.za/thesis/available/etd-10172009-102536/.

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Murovhi, Phathutshedzo. "Low temperature thermal properties of HTR nuclear fuel composite graphite." Diss., University of Pretoria, 2013. http://hdl.handle.net/2263/33156.

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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.
Dissertation (MSc)--University of Pretoria, 2013.
gm2014
Physics
Unrestricted

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Rohbeck, Nadia. "The high temperature mechanical properties of silicon carbide in TRISO particle fuel." Thesis, University of Manchester, 2014. https://www.research.manchester.ac.uk/portal/en/theses/the-high-temperature-mechanical-properties-of-silicon-carbide-in-triso-particle-fuel(275b2e07-8a5e-4b22-b575-3ded9c6b9008).html.

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The high temperature reactor (HTR) requires a completely new fuel design as it operates at around 1000°C in normal conditions and can reach up to 1600°C in case of an accident. The fuel and its cladding consist fully of ceramic materials, which precludes the possibility of a core meltdown and thus ensures inherent safety. The integral part of all HTR core designs is the tristructural-isotropic (TRISO) particle, which encapsulates the fissionable materials in succeeding coatings of pyrolytic carbon and silicon carbide (SiC). An exceptional mechanical integrity of the silicon carbide layer in all conditions is required to ensure full fission product retention. Within this work simulated TRISO fuel has been fabricated by fluidized bed chemical vapour deposition and was annealed in protective atmosphere up to 2200°C for short durations. Subsequent mechanical tests showed only minor reductions in the fracture strength of the SiC up to 2000°C. Substantial weight loss and crystal growth were observed after annealing at 2100°C and above. Raman spectroscopy identified the formation of a multi-layered graphene film covering the SiC grains after annealing and scanning electron microscopy revealed significant porosity formation within the coating from 1800°C onwards. These observations were attributed towards an evaporation-precipitation mechanism of SiC at very elevated temperatures that only slightly diminishes the hardness, elastic modulus or fracture strength, but might still be problematic in respect to fission product retention of the SiC layer. The new technique of high temperature nanoindentation was applied to measure the elastic modulus and hardness of SiC in-situ up to 500°C in argon atmosphere. The elastic modulus was found to be only slightly reduced over the measurement range, while the hardness showed a significant drop. Investigations of the deformation zone beneath the indenter tip executed by transmission electron microscopy showed slip and deformation twinning. On specimens that had been subject to neutron irradiation an irradiation hardening effect was noted. The elastic modulus showed only a minor increase compared with the non-irradiated samples. Oxidation experiments were carried out in air up to 1500°C. Analysis of the oxidation layer showed the formation of amorphous silica and cristobalite for the highest temperatures.

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Zhang, Zhan. "Neutron energy spectrum reconstruction method based for htr reactor calculations." Thesis, Georgia Institute of Technology, 2011. http://hdl.handle.net/1853/41195.

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In the deep burn research of Very High Temperature Reactor (VHTR), it is desired to make an accurate estimation of absorption cross sections and absorption rates in burnable poison (BP) pins. However, in traditional methods, multi-group cross sections are generated from single bundle calculations with specular reflection boundary condition, in which the energy spectral effect in the core environment is not taken into account. This approximation introduces errors to the absorption cross sections especially for BPs neighboring reflectors and control rods. In order to correct the BP absorption cross sections in whole core diffusion calculations, energy spectrum reconstruction (ESR) methods have been developed to reconstruct the fine group spectrum (and in-core continuous energy spectrum). Then, using the reconstructed spectrum as boundary condition, a BP pin cell local transport calculation serves an imbedded module within the whole core diffusion code to iteratively correct the BP absorption cross sections for improved results. The ESR methods were tested in a 2D prismatic High Temperature Reactor (HTR) problem. The reconstructed fine-group spectra have shown good agreement with the reference spectra. Comparing with the cross sections calculated by single block calculation with specular reflection boundary conditions, the BP absorption cross sections are effectively improved by ESR methods. A preliminary study was also performed to extend the ESR methods to a 2D Pebble Bed Reactor (PBR) problem. The results demonstrate that the ESR can reproduce the energy spectra on the fuel-outer reflector interface accurately.

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Chiuta, Steven. "The potential utilization of nuclear hydrogen for synthetic fuels production at a coal–to–liquid facility / Steven Chiuta." Thesis, North-West University, 2010. http://hdl.handle.net/10394/4840.

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The production of synthetic fuels (synfuels) in coal–to–liquids (CTL) facilities has contributed to global warming due to the huge CO2 emissions of the process. This corresponds to inefficient carbon conversion, a problem growing in importance particularly given the limited lifespan of coal reserves. These simultaneous challenges of environmental sustainability and energy security associated with CTL facilities have been defined in earlier studies. To reduce the environmental impact and improve the carbon conversion of existing CTL facilities, this paper proposes the concept of a nuclear–assisted CTL plant where a hybrid sulphur (HyS) plant powered by 10 modules of the high temperature nuclear reactor (HTR) splits water to produce hydrogen (nuclear hydrogen) and oxygen, which are in turn utilised in the CTL plant. A synthesis gas (syngas) plant mass–analysis model described in this paper demonstrates that the water–gas shift (WGS) and combustion reactions occurring in hypothetical gasifiers contribute 67% and 33% to the CO2 emissions, respectively. The nuclear–assisted CTL plant concept that we have developed is entirely based on the elimination of the WGS reaction, and the consequent benefits are investigated. In this kind of plant, the nuclear hydrogen is mixed with the outlet stream of the Rectisol unit and the oxygen forms part of the feed to the gasifier. The significant potential benefits include a 75% reduction in CO2 emissions, a 40% reduction in the coal requirement for the gasification plant and a 50% reduction in installed syngas plant costs, all to achieve the same syngas output. In addition, we have developed a financial model for use as a strategic decision analysis (SDA) tool that compares the relative syngas manufacturing costs for conventional and nuclear–assisted syngas plants. Our model predicts that syngas manufactured in the nuclear–assisted CTL plant would cost 21% more than that produced in the conventional CTL plant when the average cost of producing nuclear hydrogen is US$3/kg H2. The model also evaluates the cost of CO2 avoided as $58/t CO2. Sensitivity analyses performed on the costing model reveal, however, that the cost of CO2 avoided is zero at a hydrogen production cost of US$2/kg H2 or at a delivered coal cost of US$128/t coal. The economic advantages of the nuclear–assisted plant are lost above the threshold cost of $100/t CO2. However, the cost of CO2 avoided in our model works out to below this threshold for the range of critical assumptions considered in the sensitivity analyses. Consequently, this paper demonstrates the practicality, feasibility and economic attractiveness of the nuclear–assisted CTL plant.
Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

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Izenson, Michael G. (Michael Gary). "Effects of fuel particle and reactor core design on modular HTGR source terms." Thesis, Massachusetts Institute of Technology, 1986. http://hdl.handle.net/1721.1/14787.

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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1987.
MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.
Bibliography: v.3, leaves 516-522.
by Michael G. Izenson.
Ph.D.

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7

Gopalan, Babu. "INVESTIGATION OF HYDROGEN STORAGE IN IDEAL HPR INNER MATRIX MICROSTRUCTURE USING FINITE ELEMENT ANALYSIS." Ohio University / OhioLINK, 2006. http://rave.ohiolink.edu/etdc/view?acc_num=ohiou1159476259.

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Tshamala, Mubenga Carl. "Simulation and control implications of a high-temperature modular reactor (HTMR) cogeneration plant." Thesis, Stellenbosch : Stellenbosch University, 2014. http://hdl.handle.net/10019.1/86264.

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Thesis (MScEng)--Stellenbosch University, 2014.
ENGLISH ABSTRACT: Traditionally nuclear reactor power plants have been optimised for electrical power generation only. In the light of the ever-rising cost of dwindling fossil fuel resources as well the global polluting effects and consequences of their usage, the use of nuclear energy for process heating is becoming increasingly attractive. In this study the use of a so-called cogeneration plant in which a nuclear reactor energy source is optimised for the simultaneous production of superheated steam for electrical power generation and process heat is considered and analysed. The process heat superheated steam is generated in a once-through steam generator of heat pipe heat exchanger with intermediate fluid while steam for power generation is generated separately in a once-through helical coil steam generator. A 750 °C, 7 MPa helium cooled HTMR has been conceptually designed to simultaneously provide steam at 540 °C, 13.5 MPa for the power unit and steam at 430 °C, 4 MPa for a coal-to-liquid fuel process. The simulation and dynamic control of such a typical cogeneration plant is considered. In particular, a theoretical model of a typical plant will be simulated with the aim of predicting the transient and dynamic behaviour of the HTMR in order to provide guideline for the control of the plant under various operating conditions. It was found that the simulation model captured the behaviour of the plant reasonably well and it is recommended that it could be used in the detailed design of plant control strategies. It was also found that using a 1500 MW-thermal HTMR the South African contribution to global pollution can be reduced by 1.58%.
AFRIKAANSE OPSOMMING: Tradisioneel is kernkragaanlegte vir slegs elektriese kragopwekking geoptimeer. In die lig van die immer stygende koste van uitputbare fossielbrandstohulpbronne asook die besoedelingsimpak daarvan wêreldwyd, word die gebruik van kernkrag vir prosesverhitting al hoe meer aanlokliker. In hierdie studie word die gebruik van ‘n sogenaamde mede-opwekkingsaanleg waarin ‘n kernkragreaktor-energiebron vir die gelyktydige produksie van oorverhitte stoom vir elektriese kragopwekking en proseshitte oorweeg ontleed word. Die oorvehitte stoom word in ‘n enkeldeurvloei-stoomopwekking van die hittepyp-hitteruiler met tussenvloeistof opgewek en stoom vir kragopwekking word apart in ‘n enkeldeurvloei-spiraalspoel-stoomopwekker opgewek. ‘n 750 °C, 7 MPa heliumverkoelde HTMR is konseptueel ontwerp vir die gelytydige veskaffing van stoom by 540 °C, 13.5 MPa, vir die kragopwekkings eenheid, en stoom by 430 °C, 4 MPa, vir ‘n steenkool-tot-vloeibare (CTL) brandstoff proses. Die simulasie en dinamiese beheer van ‘n tipiese HTMR mede-opwekkingsaanleg word beskou. ‘n die besonder word ‘n teoretiese model van die transiënte en dinamiese gedrag van die aanleg gesimuleer om sodoene riglyne te identifiseer vir die ontwikkeling van dinamiese beheer strategië vir verskillende werkstoestande van die aanleg. Daar was ook gevind dat die simulasie model van die aanleg se gedrag goed nageboots word en dat dit dus gebruik kan word vir beheer strategie doeleindes. Indien so ‘n 1500 MW-termies HTMR gebruik word sal dit die Suid Afrikaanse besoedling met 1.58% sal kan verminder.

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Magnusson, Ann-Sofie. "Sveriges universitets- och högskoleförbunds-modellen : Har införandet av SUHF-modellen ökat förtroendet för lärosätenas redovisning av indirekta kostnader hos forskningsfinansiärerna?" Thesis, Högskolan i Gävle, Avdelningen för ekonomi, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:hig:diva-14899.

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Abstrakt Titel: SUHF-modellen Nivå: C-uppsats i ämnet företagsekonomi Författare: Ann-Sofie Magnusson Handledare: Jan Svanberg Datum: 2013 – juni Syfte: Syftet med studien är att utforska hur SUHF-modellen är uppbyggd och om övergången till SUHF-modellen har lett till att forskningsfinansiärernas förtroende för lärosätenas redovisning av indirekta kostnader har förbättrats. Metod: I studien har en hermeneutisk kvalitativ metod med förstående epistemologi använts. Metoden innehöll en induktiv empirisk prövning av problemformuleringen. Den induktiva empiriska undersökningen bestod av semi-strukturerade intervjuer med fyra lärosäten och fyra forskningsfinansiärer. Resultat och slutsats: Studien visar att SUHF-modellen inte är tillräckligt tydlig när det gäller gränsdragningen mellan vilka kostnader som är direkta respektive indirekta. Kommunikationen och samarbetet mellan parterna har ökat. Lärosätena upplever att förtroendet har ökat hos finansiärerna samtidigt som forskningsfinansiärernas upplevelse är olika, allt från minskat förtroende till ökat förtroende. Förslag till fortsatt forskning: Undersöka ytterligare forskningsfinansiärer och lärosäten för att kunna säkerställa att resultaten från denna studie är korrekta samt undersöka orsaken till varför förtroendet är oförändrat och dessutom har minskat hos forskningsfinansiärer efter införandet av SUHF-modellen. Nyckelord: SUHF-modell, full kostnadstäckning, indirekta kostnader, förtroende, engagemang, kommunikation, samarbete
Abstract Title: SUHF model Level: Final assignment for Bachelor Degree in Business Administration Author: Ann-Sofie Magnusson Supervisor: Jan Svanberg Date: 2013 - June Aim: The purpose of this study is to explore how the SUHF model is built and if the transition to the SUHF model has led to improvement of research funder’s confidence in the institutions´ accounting of indirect costs. Method: The study has been carried out with a qualitative hermeneutic approach with understanding epistemology. The method included an inductive empirical test of the problem formulation. The inductive empirical study consisted of semi-structured interviews with four universities and four research funders. Result and Conclusions: The study shows that the SUHF model is not sufficiently clear about the distinction between the costs that are direct and indirect. Communication and cooperation between the parties has increased. The universities feel that their confidence has improved among financiers while the research funders' experience is different, ranging from loss of confidence to increased confidence. Suggestions for future research: Investigate additional research funders and universities in order to ensure that the results from this study are reliable and investigate the reasons why the confidence is unchanged and also has reduced with research funders after the introduction of the SUHF model. Key words: SUHF model, full costs recovery, indirect costs, trust, commitment, communication, cooperation

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Tiwari, Housila. "INVESTIGATION OF THE FEASIBILTY OF METALS, POLYMERIC FOAMS, AND COMPOSITE FOAM FOR ON-BOARD VEHICULAR HYDROGEN STORAGE VIA HYDROSTATIC PRESSURE RETAINMENT (HPR) USING IDEAL BCC MICROSTRUCTURE." Ohio University / OhioLINK, 2007. http://rave.ohiolink.edu/etdc/view?acc_num=ohiou1186967436.

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Books on the topic "Htr fuel"

1

United States. Congress. House. Committee on Energy and Commerce. Subcommittee on Energy and Power. Alternative automotive fuels: Hearings before the Subcommittee on Energy and Power of the Committee on Energy and Commerce, House of Representatives, One Hundredth Congress, first session, on H.R. 168, H.R. 1595, H.R. 2031, and H.R. 2052 ... June 17, 24, and July 9, 1987. Washington: U.S. G.P.O., 1988.

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To establish a grant program whereby moneys collected from violations of the Corporate Average Fuel Economy Program are used to expand infrastructure necessary to increase the availability of alternative fuels: Report (to accompany H.R. 5534) (including cost estimate of the Congressional Budget Office). [Washington, D.C: U.S. G.P.O., 2006.

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United States. Congress. House. Committee on Energy and Commerce. To establish a grant program whereby moneys collected from violations of the Corporate Average Fuel Economy Program are used to expand infrastructure necessary to increase the availability of alternative fuels: Report (to accompany H.R. 5534) (including cost estimate of the Congressional Budget Office). [Washington, D.C: U.S. G.P.O., 2006.

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Williams, Aleta L. Why her?: A full novel. [North Charleston, SC]: [CreateSpace Independent Publishing Platform], 2013.

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Strømsted, Finn. En fugl har tent meg. Oslo: Aschehoug, 1995.

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United States. Congress. Senate. Committee on Armed Services. Full committee consideration of H.R. 3283 ... H.R. 3140 ... H.R. 2873 ... Washington: U.S. G.P.O., 1988.

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Economy, United States Congress House Committee on Energy and Commerce Subcommittee on Environment and the. H.R. 4345, the Domestic Fuels Protection Act of 2012: Hearing before the Subcommittee on Environment and Economy of the Committee on Energy and Commerce, House of Representatives, One Hundred Twelfth Congress, second session, April 19, 2012. Washington: U.S. Government Printing Office, 2013.

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Markgraf, J. F. W. HFR irradiation testing of light water reactor (LWR) fuel. Luxembourg: Commission of the European Communities, 1985.

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United States. Congress. Senate. Committee on Armed Services. Full committee consideration of H.R. 2948 ... and H.R. 2974 ... Washington: U.S. G.P.O., 1987.

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United States. Congress. Senate. Committee on Armed Services. Full committee consideration of H.R. 2948 ... and H.R. 2974 ... Washington: U.S. G.P.O., 1987.

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Book chapters on the topic "Htr fuel"

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Kania, Michael J., Heinz Nabielek, and Karl Verfondern. "SiC-Coated HTR Fuel Particle Performance." In Ceramic Engineering and Science Proceedings, 33–70. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118217535.ch4.

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Li, Ning, Hongjun Zhang, and Xiaogang Xu. "Structural Design and Verification of the CNFC-HTR New Fuel Transport Container." In Proceedings of The 20th Pacific Basin Nuclear Conference, 873–81. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2317-0_82.

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Karriem, Z., C. Stoker, and F. Reitsma. "MCNP Modelling of HTGR Pebble-Type Fuel." In Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, 841–46. Berlin, Heidelberg: Springer Berlin Heidelberg, 2001. http://dx.doi.org/10.1007/978-3-642-18211-2_134.

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Choi, Hangbok, Robert Schleicher, and Myunghee Choi. "Physics Analysis of Alternative Fuel Options for HTGR." In Proceedings of The 20th Pacific Basin Nuclear Conference, 801–11. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2317-0_76.

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Bornhöft, Gudrun, Klaus v. Ammon, Marco Righetti, André Thuneysen, and Peter F. Matthiessen. "Full discussion of the HTA results." In Homeopathy in Healthcare – Effectiveness, Appropriateness, Safety, Costs, 193–204. Berlin, Heidelberg: Springer Berlin Heidelberg, 2011. http://dx.doi.org/10.1007/978-3-642-20638-2_13.

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Karplus, Rivka. "People Facing the Question of Euthanasia: Patients, Family and Friends, Healthcare Workers." In Euthanasia: Searching for the Full Story, 49–59. Cham: Springer International Publishing, 2021. http://dx.doi.org/10.1007/978-3-030-56795-8_5.

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AbstractSince the Oregon Death with Dignity Act was adopted in 1997, an increasing number of people have requested euthanasia, showing that life has become unbearable for them. However, a person who expresses the wish to die by euthanasia is not saying that he/she prefers death to life, but rather that death seems preferable to life under the actual circumstances. In order to respond to a person’s suffering, we need to understand the nature of that suffering, as they experience it. Suffering may be physical, psychological, relational, spiritual, or existential; frequently these different aspects overlap or intermingle, particularly in a serious illness. Euthanasia does not improve life—it ends it by giving death. But when the response involves listening and accepting the person in his/her present situation, it becomes possible to work together with the person to see what can be done to help reduce suffering. We can look for means of relief for the person’s individual, unique suffering, in partnership with the patient and his/her family and friends, using the resources of both medical knowledge and our shared humanity. The willingness to walk on this shared path with the sick person is in itself an affirmation of his/her human dignity.

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Kim, Jun Hwan, Jong Hyuk Baek, Sung Ho Kim, and Chan Bock Lee. "Effect of Heat Treatment on the Mechanical Properties of HT9 Fuel Cladding Tube for Sodium-Cooled Fast Reactor (SFR)." In Proceedings of the 8th Pacific Rim International Congress on Advanced Materials and Processing, 2431–34. Cham: Springer International Publishing, 2013. http://dx.doi.org/10.1007/978-3-319-48764-9_300.

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Kim, Jun Hwan, Jong Hyuk Baek, Sung Ho Kim, and Chan Bock Lee. "Effect of Heat Treatment on the Mechanical Properties of HT9 Fuel Cladding Tube For Sodium-Cooled Fast Reactor (SFR)." In PRICM, 2431–34. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118792148.ch300.

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"Full Status." In Her Oxford, 274–84. Vanderbilt University Press, 2008. http://dx.doi.org/10.2307/j.ctv16h2nb9.26.

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"THE MOVE TOWARD FULL PARTICIPATION." In Her Story, 273–303. 2nd ed. Fortress Press, 2006. http://dx.doi.org/10.2307/j.ctv1hqdj65.11.

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Conference papers on the topic "Htr fuel"

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Marmier, Alain, Michael A. Fu¨tterer, Mathias Laurie, and Chunhe Tang. "Preliminary Results of the HFR-EU1 Fuel Irradiation of INET and AVR Pebbles in the HFR Petten." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58049.

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The irradiation experiment HFR-EU1 in the HFR Petten is currently being conducted by the European Commission’s Joint Research Centre – Institute for Energy (JRC-IE). The irradiation targets are 5 spherical High Temperature Reactor (HTR) fuel pebbles, 2 of INET production and 3 of former German production. Both types are made of TRISO coated particles and are tested for their potential for very high temperature performance and high burn-up. The irradiation started on 29 September 2006 and, by 24 February 2008, had accumulated 12 reactor cycles totaling 332.8 efpd and a calculated maximum burn-up of 8.9% FIMA (INET) and 11.2% FIMA (AVR). The objective of the HFR-EU1 test is to irradiate 5 HTR fuel pebbles at conditions beyond the characteristics of current HTR reactor designs with pebble bed cores, e.g. HTR-Modul, HTR-10 and PMBR. This should demonstrate that pebble bed HTRs are capable of enhanced performance in terms of sustainability (further increased power conversion efficiency, improved fuel use) and thus reduced waste production. The surface temperature of all pebbles was held constant during the irradiation, with the exception of HFR downtime and power transients. HFR-EU1 should demonstrate the feasibility of low coated particle failure fractions under normal operating conditions and more specifically: • high fuel surface temperature of 900°C (INET) and 950°C (AVR); • very high burn-up of 17% FIMA (INET) and 20% FIMA (AVR) which is significantly higher than the license limit of the HTR-Modul (approx. 8% FIMA); it will be explained in this paper why this objective had to be somewhat reduced due to excessive irradiation time requirements and technological difficulties. This paper provides the irradiation history of the experiment performed so far including data on fission gas release.

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van Heek, Aliki, Florence Charpin, Steven van der Marck, Jorrit Wolters, Christos Trakas, Luis Aguiar, Eleonora Bomboni, et al. "HTR Pebble Fuel Burnup Experimental Benchmark." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58134.

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The HTR pebble fuel experiment HFR EU1bis was irradiated in the High Flux Reactor, Petten, The Netherlands, in 2004 and 2005. It consisted of five fuel pebbles from the German HTR program (GLE4 type, UO2 fuel, 16.75% enrichment) and six minisamples (UO2 fuel, 9.75% enrichment). Its instrumentation included three flux monitor sets. The experiment was loaded in a REFA-170 rig, surrounded by a strongly moderating filler element. The central fuel temperature was held at 1250°C during the irradiation. In the framework of the European RAPHAEL project, Post Irradiation Examination (PIE) has been done at NRG in Petten, The Netherlands and at JRC ITU in Karlsruhe, Germany. In Petten, flux monitor analysis has been done, whereas in Karlsruhe, a quantitative evaluation of γ-emitters was used to make a burn-up determination. A benchmark description based on this experiment has been written by NRG. Until now, five RAPHAEL project participants have modeled the experiment, each with their own neutronics code system. Participating codes are three versions of MONTEBURNS (MCNP with ORIGEN), MURE/MCNP and OCTOPUS (MCNP with FISPACT). The pebble burnup and isotopic inventories (Bq/gram initial HM) of selected fission products and actinides in the fuel pebble samples are both calculated and determined by gamma spectrometry, mass spectrometry and ion chromatography by JRC-ITU. Additionally, two participants calculated the flux monitor activities that were measured by NRG. A burnup measurement of 11.0 % FIMA by gamma spectrometry could be confirmed by calculation. Differences between the various modeling approaches and the experimental burn-up determination will be discussed.

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Guittonneau, Fabrice, Abdesselam Abdelouas, Bernd Grambow, Manoe¨l Dialinas, and Franc¸ois Cellier. "New Methods for HTR Fuel Waste Management." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58112.

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Considering the need to reduce waste production and greenhouse emissions by still keeping high energy efficiency, various 4th generation nuclear energy systems have been proposed. As far as graphite moderated reactors are concerned, one of the key issues is the large volumes of irradiated graphite encountered (1770 m3 for fuel elements and 840 m3 for reflector elements during the lifetime (60 years) of a single reactor module [1]). With the objective to reduce volume of waste in the HTR concept, it is very important to be able to separate the fuel from low level activity graphite. This requires to separate TRISO particles from the graphite matrix with the sine qua non condition to not break TRISO particles in case of future embedding of particles in a matrix for disposal. According to National Regulatory Systems, in case of limited graphite waste production or of short duration HTR projects (e.g. in Germany), direct disposal without separation is acceptable. Nevertheless, in case of large scale deployment of HTR technology, such approach is not economical and sustainable. Previous attempts in graphite management (furnace, fluidised bed and laser incinerations and encapsulation matrices) dealt with graphite matrix only. These are the reasons why we studied the management of irradiated compact-type fuel element. We simulated the presence of fuel in the particles by using ZrO2 kernels. Compacts with ZrO2 TRISO particles were manufactured by AREVA NP. Two original methods have been studied. First, we tested high pressure jet to erode graphite and clean TRISO particles. Best erosion rate reached about 0.18 kg/h for a single nose ending. Examination of treated graphite showed a mixture of undamaged TRISO particles, particles that have lost the outer pyrolytic carbon layer and ZrO2 kernels. Secondly, we studied the thermal shock method by immerging successively graphite into liquid nitrogen and hot water to cause fracturing of the compact. This produced particles and graphite fragments with diameter ranging from several centimetres to less than 500 μm. This relatively simple and economic method may potentially be considered as a pretreatment step and be coupled with other method(s) before reprocessing and recycling for example.

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Marmier, Alain, Michael A. Fu¨tterer, Kamil Tucˇek, Han de Haas, Jim C. Kuijper, and Jan Leen Kloosterman. "Revisiting the Concept of HTR Wallpaper Fuel." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58114.

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Good safety characteristics are an outstanding feature of High Temperature Reactors (HTR): • The high graphite inventory in the core provides significant thermal inertia. Graphite also has a high thermal conductivity, which facilitates the transfer of heat to the reflector, and it can withstand high temperatures; • The strongly negative Doppler coefficient gives a negative feedback, such that the reactor shuts down by itself in overpower accidental conditions; • The high quality of fuel elements — tri-isotropic (TRISO) coated particles — minimizes operational and accidental fission gas release. The materials selected have resistance to high temperatures; • The low power density enables stabilization of core temperature significantly below the maximum allowable, even in case of severe accidents (such as loss-of-coolant accident). Together, these aspects significantly reduce the risk of massive fission product release, which is one of the attractive features of HTRs. The fuel that is currently used in pebble bed reactors such as AVR, HTR-10 and soon PBMR is based on a homogeneous distribution of coated particles within a fuel pebble. This homogenizes power density in the pebble, but creates a radial temperature gradient across the fuel sphere. Fuel particles placed at its centre has the highest temperature. Reducing the average temperature of particles would help preserve their integrity and maintain the resistance of the first barrier against fission product release. As early as the 1970s, attempts were made to reduce the peak fuel temperature by means of so-called “wallpaper fuel”, in which the fuel is arranged in a spherical shell within a pebble. At that time, the production process was not sufficiently mature and had caused unacceptable damage to the (less performing) BISO particles, which is why this fundamentally promising concept was abandoned. In this paper, proposals will be put forward to improve the production process. This paper further exploits the wallpaper concept, not only from the point of view of temperature reduction, but also for enhanced neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. Parameters modified were the density of the central fuel-free graphite zone and the packing fraction of the fuel zone. It is demonstrated that this fuel type impacts positively on the fuel cycle, reduces production of minor actinides (MA) and improves the safety-relevant parameters of the reactor. A comparison of these characteristics with PBMR-type fuel is presented. The calculations were performed using Monte Carlo neutron transport and depletion codes MCNP/MCB and the deterministic code WIMS. By comparison with PBMR fuel, the “wallpaper design” of the fuel pebble results in an effective neutron multiplication coefficient increase (by about 2%), which is combined with a decrease of between 3 and 15% in MA production. An improved neutron economy of the heterogeneous design enables enrichment of the “wallpaper type” of fuel to be reduced by more than 6%.

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Freis, D., P. D. Bottomley, J. P. Hiernaut, J. Y. Colle, J. Ejton, and W. de Weerd. "Post Irradiation Examination of HTR Fuel at ITU Karlruhe." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58329.

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In the last years considerable efforts have been made at the Institute for Transuranium Elements (ITU) in order to reestablish European knowledge and ability in safety testing of irradiated high temperature reactor (HTR) Fuel Elements. In the framework of the 6th European framework programme a cold finger apparatus (Ku¨FA) furnace, formerly installed at FZ-Ju¨lich (FzJ), has been installed in a hot cell at ITU [Freis 2008] in order to test fission product release under high temperature and non-oxidising conditions. Several analytical methods (e.g. Gamma-spectrometry, mass-spectrometry) have been applied in order to analyse different isotopes released during Ku¨FA tests. After the heating tests, examinations of the fuel elements were performed including scanning electron microscopy (SEM) and micro-hardness testing of coated particles. Individual coated particles were object of heating tests in a Knudsen cell with a coupled mass spectrometer measuring all released species. In order to cover more accident scenarios, a second furnace for oxidising-conditions (air- or water-ingress) was constructed and installed in a cold lab. Furthermore a disintegration apparatus, based on anodic oxidation, was constructed and fuel elements were dissolved obtaining thousands of individual coated particles for further examination. A fully automated irradiated microsphere gamma analyzer (IMGA) is under construction and will be used, in particular, to identify and sort out failed particles.

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Hittner, Dominique. "The Renewal of HTR Development in Europe." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22423.

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The European HTR-Technology Network (HTR-TN), created in 2000, presently groups 20 organisations from European nuclear research and industry for developing the technologies of direct-cycle modular HTRs, which presently raise a large world-wide interest, because of their high potential for economic competitiveness, natural resource sparing, safety and minimisation of the waste impacts, in line with the goals of sustainable development of Generation IV. All aspects of HTR technologies are addressed by HTR-TN, from the reactor physics to the development of materials, fuel and components. Most of this activity is supported by the European Commission in the frame of its 5th Euratom Framework Programme. The first results of HTR-TN programme are given: the analysis of the reactor physics international benchmark on the commissioning tests of HTTR (Japan), the long term behaviour of spent HTR fuel in geologic disposal conditions, the preparation of a very high burnup fuel irradiation and the development of fabrication processes for producing high performance coated particles, etc.

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Grambow, B., A. Abdelouas, F. Guittonneau, J. Vandenborre, J. Fachinger, W. von Lensa, P. Bros, et al. "The Backend of the Fuel Cycle of HTR/VHTR Reactors." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58177.

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For various countries, the direct disposal of high level nuclear fuel wastes is a key option for the backend of the fuel cycle. For HTR/VHTR reactors this is assumed for the introductory phase of this reactor system. However, closed fuel cycles or a separation of spent coated-particles from the graphite moderator and specific treatment, conditioning and disposal of these waste streams are also possible. In the European Community project “RAPHAEL”, fuel waste performance is going to be studied in depth, including post-irradiation fuel characterization, analysis of the stability and failure mechanism of coatings and of fuel kernels and overall performance of waste packages with compact fuel and/or only with fuel particles in geological disposal environments. Different confinement matrices for separated fuel particles (vitrification, SiC, ZrO2) have been adapted to limit release of radionuclides into groundwater at low temperatures over geological time spans. The investigations are limited to Low-Enriched Uranium (LEU) fuel with uranium oxide and uranium oxycarbide kernels that will allow higher burn-up, but may be more susceptible to leaching.

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Wang, Meng-Jen, Jinn-Jer Peir, Chen-Wei Chi, and Jenq-Horng Liang. "A Parametric Study of Fuel Lattice Design for HTR-10." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29253.

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In this study, the multiplication factor and neutron spectrum behaviors were investigated against the moderator-to-fuel ratio, the fuel loading height, and the detector location in HTR-10. The MCNP5 computer code (version 1.51) was employed to perform all the simulation computations. The results revealed that the multiplication factor varies significantly depending on the moderator-to-fuel ratio and the fuel loading height due to the competition among the neutron moderation and absorption abilities of the moderator as well as the neutron production ability of the fuel. Due to its inherent stability, HTR-10 is deliberately designed such that the multiplication factor decreases and the neutron spectrum softens as the moderator-to-fuel ratio increases. The average neutron energy level in the HTR-10 fuel balls is approximately 200 keV and ranges from smallest to largest at the middle, bottom, and top of the reactor core, respectively.

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Wang, Jinhua, Bing Wang, Bin Wu, and Yue Li. "Design of the Spent Fuel Storage Well of HTR-PM." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60051.

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There are more than 400 reactors in operation to generate electricity in the world, most of them are pressurized water reactors and boiling water reactors, which generate great amount of spent fuel every year. The residual heat power of the spent fuel just discharged from the reactor core is high, it is required to store the spent fuel in the spent fuel storage pool at the first 5 years after discharged from the reactor, and then the spent fuel could be moved to the interim storage facility for long term storage, or be moved to the factory for final treatment. In the accident of the Fukushima in 2011, the spent fuel pool ruptured, which led to the loss of coolant accident, it was very danger to the spent fuel assemblies stored in the pool. On the other hand, the spent fuel stored in the dry storage facility was safe in the whole process of earthquake and tsunami, which proved inherent safety of the spent fuel dry storage facility. In china, the High Temperature gas cooled Reactor (HTR) is developing for a long time in support of the government. At the first stage, HTR-10 with 10MW thermal power was designed and constructed in the Institute of Nuclear Energy Technology (INET) of Tsinghua University, and then the High Temperature Reactor-Pebble bed Modules (HTR-PM) is designed to meet the commercial application, which is in constructing process in Shandong Province. HTR has some features of the generation four nuclear power plant, including inherent safety, avoiding nuclear proliferation, could generate high temperature industrial heat, and so on. Spherical fuel elements would be used as fuel in HTR-PM, there are many coating fuel particles separated in the fuel element. As the fuel is different for the HTR and the PWR, the fuel element would be discharged into the appropriate spent fuel canister, and the canister would be stored in the appropriate interim storage facility. As the residual power density is very low for the spent fuel of HTR, the spent fuel canister could be cooled with air ventilation without water cooling process. The advantage of air cooling mode is that it is no need to consider the residual heat removal depravation due to loss of coolant accident, so as to increase the inherent safety of the spent fuel storage system. This paper introduced the design, arrangement and safety characteristics of the spent fuel storage well of HTR-PM. The spent fuel storage wells have enough capacity to hold the total spent fuel canisters for the HTR-PM. The spent fuel storage facility includes several storage wells, cold intake cabin, hot air discharge cabin, heat shield cylinders, well lids and so on. The cold intake cabin links the inlets of all the wells, which would be used to import cold air to every well. The hot air discharge cabin links the outlets of all the wells, which would be used to gather heated air discharged from every well, the heated air would be discharged to the atmosphere through the ventilating pipe at the top of the hot air cabin. The design of the spent fuel storage well and the ventilating pipe could discharge the residual heat of the spent fuel canisters in the storage wells, which could ensure the operating safety of the spent fuel storage system.

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Renze, Wang, Zhang Jiangang, Li Guoqiang, Zhuang Dajie, Meng Dongyuan, Wang Xuexin, Sun Hongchao, and Sun Shutang. "PSA Research of Transport of New Fuel of HTR-PM." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66012.

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Probabilistic safety assessment (PSA) method was explored to be introduced in radiation risk assessment of radioactive material transport (RMT), and then radiation risk of road transport (RT) of new fuel element (FE) of commercial High Temperature Reactor-Pebblebed Modules (HTR-PM) was analyzed. By accident scenario analysis, mechanical analysis and criticality analysis, both accident conditions of package radiation level hoist and criticality were chosen for accident frequency analysis. Since the results show that accident frequency of package radiation level hoist is low and accident frequency of criticality is extremely low, criticality accident can be neglected in the sequential risk assessment. The consequence of typical accident scenarios was estimated, and the results show that the maximum external exposure dose arising from the accident for emergency workers is 0.55mSv, and 4.55×10−3mSv for public people around, which is acceptable. The global radiation risk is 1.24×10−10person·Sv/(vehicle·each transport), for which impact accident contributes the maximum percent.

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Reports on the topic "Htr fuel"

1

Gerhard Strydom. Reactor Physics Characterization of the HTR Module with UCO Fuel. Office of Scientific and Technical Information (OSTI), January 2011. http://dx.doi.org/10.2172/1009138.

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Francesco Venneri, Chang-Keun Jo, Jae-Man Noh, Yonghee Kim, Claudio Filippone, Jonghwa Chang, Chris Hamilton, et al. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor. Office of Scientific and Technical Information (OSTI), September 2010. http://dx.doi.org/10.2172/991901.

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Brian Boer and Abderrafi M. Ougouag. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors. Office of Scientific and Technical Information (OSTI), March 2011. http://dx.doi.org/10.2172/1013722.

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Bess, John D., Leland M. Montierth, James W. Sterbentz, J. Blair Briggs, Hans D. Gougar, Luka Snoj, Igor Lengar, and Oliver Koberl. HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio. Office of Scientific and Technical Information (OSTI), March 2014. http://dx.doi.org/10.2172/1117731.

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John D. Bess and Leland M. Montierth. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO. Office of Scientific and Technical Information (OSTI), March 2013. http://dx.doi.org/10.2172/1073776.

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Vincent Descotes. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation. Office of Scientific and Technical Information (OSTI), March 2011. http://dx.doi.org/10.2172/1013715.

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Martin, William R., John C. Lee, Alan baxter, and Chuck Wemple. Creation of a Full-Core HTR Benchmark with the Fort St. Vrain Initial Core and Assessment of Uncertainties in the FSV Fuel Composition and Geometry. Office of Scientific and Technical Information (OSTI), March 2012. http://dx.doi.org/10.2172/1047488.

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Michael A. Pope. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities. Office of Scientific and Technical Information (OSTI), October 2011. http://dx.doi.org/10.2172/1042392.

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John D. Bess, Barbara H. Dolphin, James W. Sterbentz, Luka Snoj, Igor Lengar, and Oliver Köberl. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio. Office of Scientific and Technical Information (OSTI), March 2013. http://dx.doi.org/10.2172/1064064.

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John D. Bess, Barbara H. Dolphin, James W. Sterbentz, Luka Snoj, Igor Lengar, and Oliver Köberl. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio. Office of Scientific and Technical Information (OSTI), March 2012. http://dx.doi.org/10.2172/1042385.

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