Fully Ceramic Microencapsulated Fuel in High-Temperature Gas-Cooled Reactors PDF Download
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Author: Cihang Lu
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Languages : en
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Highly innovative nuclear reactor technologies have the potential to meet the global energy demand while reducing carbon emissions. Generation IV reactors, with advances in safety, reliability, sustainability and economic benefits, are currently being investigated. The high-temperature gas-cooled reactor (HTGR), moderated by graphite and cooled by helium, has the highest technology readiness level compared to the other Generation IV reactor designs.Conventional HTGR fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a graphite matrix. Several historic HTGRs fueled with conventional fuel have been constructed and operated, including the Peach Bottom Unit No.1 reactor operated in Pennsylvania and the Fort St. Vrain reactor that was operated in Colorado.The FCM fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a silicon carbide (SiC) matrix. Compared to the conventional HTGR fuel, the FCM fuel could potentially enhance the safety of the reactor due to the numerous advantages provided by the SiC matrix. The FCM fuel features enhanced ability to retain fission products. The FCM fuel exhibits a greater stability under irradiation and less swelling after irradiation. Moreover, the FCM fuel has better mechanical characteristics and would be less sensitive to physical disturbances. The FCM fuel has higher oxidation resistance and would suffer less damage in air-ingress accidents. The SiC matrix may also increase the proliferation resistance of the FCM fuel.However, due to the replacement of the graphite matrix in the conventional HTGR fuel, the FCM fuel hardens the neutron spectrum in the reactor core. This may further cause economic penalties of an FCM-fueled HTGR as well as a higher fuel temperature which jeopardizes the core safety.This dissertation proved the viability of FCM-fueled HTGRs by answering the following six questions based on analysis of experimental data as well as neutronics and thermal-hydraulics numerical calculations:(1) What are the key changes in fuel cycle performance and fuel cost of HTGRs with the FCM fuel?(2) What is the potential impact of the FCM fuel on reactor performance and safety characteristics of HTGRs?(3) How does the FCM fuel impact anticipated transients and design-basis accidents?(4) What are the most important parameters for each of the design-basis accidents and their sensitivities to the maximum fuel temperature?(5) What is the kinetics of the annealing process of neutron-irradiated SiC?(6) Does the irradiation defect annealing process of SiC significantly impact fuel temperature during design-basis accidents?The reference HTGR core configuration considered was the General Atomics designed 350-MWt prismatic mHTGR which has a prismatic block configuration similar to the Fort St. Vrain reactor.In this dissertation, I identified three FCM fuel options which are able to maintain the fuel cycle length of the reference core. However, because of the higher natural resource requirement, the FCM fuel cycle cost could be up to 74% more expensive than the conventional HTGR fuel. The impact of the FCM fuel on the other parameters which are important to the core safety, including decay power, reactivity temperature coefficients and control rod worth, is minor. I investigated three typical design-basis accidents of the HTGRs, including the pressurized loss of forced cooling accident, the depressurized loss of forced cooling accident and the control rod withdrawal accident. The maximum fuel temperature of an FCM-fuel core could be up to 65 K higher than that of the reference core during normal operating conditions, and the peak maximum fuel temperature of an FCM-fuel core could be up to 55 K higher than that of the reference core during those design-basis accidents. I also studied the sensitivity of the maximum fuel temperature to various parameters of interest during different operating conditions and identified the steady-state power distribution to have the largest impact on the peak maximum fuel temperature during the design-basis accidents. By analyzing acquired experimental data, I elucidated information on the kinetics of the SiC annealing process. I further estimated the maximum possible impact of the SiC annealing process on the maximum fuel temperature of FCM-fueled HTGRs during design-basis accidents based on conservative assumptions. The SiC annealing process would at most increase the peak maximum fuel temperature of an FCM-fueled HTGR by 40 K.According to the calculations conducted, the increase of the maximum fuel temperature caused by the use of the FCM fuel in HTGRs would not exceed 100 K which is minor compared to the maximum fuel temperature of around 1200 K in the reference core during normal operating conditions. Therefore, the use of the FCM fuel in HTGRs is viable.Additionally, by comparing calculation results with experimental data, I demonstrated the validity of the system analysis code RELAP5-3D to conduct thermal-hydraulics calculations of transients in HTGRs.
Author: Cihang Lu
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
Highly innovative nuclear reactor technologies have the potential to meet the global energy demand while reducing carbon emissions. Generation IV reactors, with advances in safety, reliability, sustainability and economic benefits, are currently being investigated. The high-temperature gas-cooled reactor (HTGR), moderated by graphite and cooled by helium, has the highest technology readiness level compared to the other Generation IV reactor designs.Conventional HTGR fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a graphite matrix. Several historic HTGRs fueled with conventional fuel have been constructed and operated, including the Peach Bottom Unit No.1 reactor operated in Pennsylvania and the Fort St. Vrain reactor that was operated in Colorado.The FCM fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a silicon carbide (SiC) matrix. Compared to the conventional HTGR fuel, the FCM fuel could potentially enhance the safety of the reactor due to the numerous advantages provided by the SiC matrix. The FCM fuel features enhanced ability to retain fission products. The FCM fuel exhibits a greater stability under irradiation and less swelling after irradiation. Moreover, the FCM fuel has better mechanical characteristics and would be less sensitive to physical disturbances. The FCM fuel has higher oxidation resistance and would suffer less damage in air-ingress accidents. The SiC matrix may also increase the proliferation resistance of the FCM fuel.However, due to the replacement of the graphite matrix in the conventional HTGR fuel, the FCM fuel hardens the neutron spectrum in the reactor core. This may further cause economic penalties of an FCM-fueled HTGR as well as a higher fuel temperature which jeopardizes the core safety.This dissertation proved the viability of FCM-fueled HTGRs by answering the following six questions based on analysis of experimental data as well as neutronics and thermal-hydraulics numerical calculations:(1) What are the key changes in fuel cycle performance and fuel cost of HTGRs with the FCM fuel?(2) What is the potential impact of the FCM fuel on reactor performance and safety characteristics of HTGRs?(3) How does the FCM fuel impact anticipated transients and design-basis accidents?(4) What are the most important parameters for each of the design-basis accidents and their sensitivities to the maximum fuel temperature?(5) What is the kinetics of the annealing process of neutron-irradiated SiC?(6) Does the irradiation defect annealing process of SiC significantly impact fuel temperature during design-basis accidents?The reference HTGR core configuration considered was the General Atomics designed 350-MWt prismatic mHTGR which has a prismatic block configuration similar to the Fort St. Vrain reactor.In this dissertation, I identified three FCM fuel options which are able to maintain the fuel cycle length of the reference core. However, because of the higher natural resource requirement, the FCM fuel cycle cost could be up to 74% more expensive than the conventional HTGR fuel. The impact of the FCM fuel on the other parameters which are important to the core safety, including decay power, reactivity temperature coefficients and control rod worth, is minor. I investigated three typical design-basis accidents of the HTGRs, including the pressurized loss of forced cooling accident, the depressurized loss of forced cooling accident and the control rod withdrawal accident. The maximum fuel temperature of an FCM-fuel core could be up to 65 K higher than that of the reference core during normal operating conditions, and the peak maximum fuel temperature of an FCM-fuel core could be up to 55 K higher than that of the reference core during those design-basis accidents. I also studied the sensitivity of the maximum fuel temperature to various parameters of interest during different operating conditions and identified the steady-state power distribution to have the largest impact on the peak maximum fuel temperature during the design-basis accidents. By analyzing acquired experimental data, I elucidated information on the kinetics of the SiC annealing process. I further estimated the maximum possible impact of the SiC annealing process on the maximum fuel temperature of FCM-fueled HTGRs during design-basis accidents based on conservative assumptions. The SiC annealing process would at most increase the peak maximum fuel temperature of an FCM-fueled HTGR by 40 K.According to the calculations conducted, the increase of the maximum fuel temperature caused by the use of the FCM fuel in HTGRs would not exceed 100 K which is minor compared to the maximum fuel temperature of around 1200 K in the reference core during normal operating conditions. Therefore, the use of the FCM fuel in HTGRs is viable.Additionally, by comparing calculation results with experimental data, I demonstrated the validity of the system analysis code RELAP5-3D to conduct thermal-hydraulics calculations of transients in HTGRs.
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Languages : en
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"Member States of the IAEA have pursued activities focused on developing high temperature gas cooled reactors (HTGRs) with an aim to build innovative nuclear fuel cycles and reactor systems for high temperature applications such as process heat and hydrogen production, apart from electricity generation. This publication documents the knowledge and experience in the development of HTGRs gained over fifty years and will serve as a basis for further development of fuels and reactor systems ... Subject Classification : 0802 - Fuel fabrication and performance. Responsible Officer : Mr. Uddharan Basak, NEFW."--Résumé de l'éditeur
Author: Jinbiao Xiong
Publisher: Frontiers Media SA
ISBN: 2889745236
Category : Technology & Engineering
Languages : en
Pages : 129
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Languages : en
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The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO2 kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 {micro}mm kernel diameter, 100 {micro}mm buffer, 35 {micro}mm IPyC, 35 {micro}mm SiC, 40 {micro}mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10−2 failure probability. For a 'best-estimate' FGR fraction of 50% and a more modest burnup target level of 500 MWd/kg, the failure probability drops below 2.0 x 10−5, the typical performance of TRISO fuel made under the German HTR research program. An optimization study on particle design shows improved performance if the buffer size is increased from 100 to 120 {micro}mm while reducing the OPyC layer. The presence of the latter layer does not provide much benefit at high burnup levels (and fast fluence levels). Normally the shrinkage of the OPyC would result in a beneficial compressive force on the SiC coating. However, at high fluence levels the shrinkage is expected to turn into swelling, resulting in the opposite effect. However, this situation is different when the SiC-matrix, in which the particles are embedded, is also considered: the OPyC swelling can result in a beneficial compressive force on the SiC coating since outward displacement of the OPyC outer surface is inhibited by the presence of the also-swelling SiC matrix. Taking some credit for this effect by adopting a 5 {micro}mm SiC-matrix layer, the optimized particle (100 {micro}mm buffer and 10 {micro}mm OPyC), gives a failure probability of 1.9 x 10−4 for conservative conditions. During a LOCA transient, assuming core re-flood in 30 seconds, the temperature of the coated particle can be expected to be about 200K higher than nominal temperature (900K). For this event the particle failure fraction for a conservative case is 1.0 x 10−2, for the optimized particle design. For a FGR-fraction of 50% this value reduces to 6.4 x 10−4.
Author: Enrico De Micheli
Publisher: MDPI
ISBN: 3039434950
Category : Mathematics
Languages : en
Pages : 182
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The charm of Mathematical Physics resides in the conceptual difficulty of understanding why the language of Mathematics is so appropriate to formulate the laws of Physics and to make precise predictions. Citing Eugene Wigner, this “unreasonable appropriateness of Mathematics in the Natural Sciences” emerged soon at the beginning of the scientific thought and was splendidly depicted by the words of Galileo: “The grand book, the Universe, is written in the language of Mathematics.” In this marriage, what Bertrand Russell called the supreme beauty, cold and austere, of Mathematics complements the supreme beauty, warm and engaging, of Physics. This book, which consists of nine articles, gives a flavor of these beauties and covers an ample range of mathematical subjects that play a relevant role in the study of physics and engineering. This range includes the study of free probability measures associated with p-adic number fields, non-commutative measures of quantum discord, non-linear Schrödinger equation analysis, spectral operators related to holomorphic extensions of series expansions, Gibbs phenomenon, deformed wave equation analysis, and optimization methods in the numerical study of material properties.
Author: Michael A. Stawicki
Publisher:
ISBN:
Category :
Languages : en
Pages : 133
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(cont.) It is concluded that accurate modeling of TRISO particles depends on having very high accuracy data describing material properties and a very good understanding of the uncertainties in those measurements.
Author: Kevin M. Fox
Publisher: John Wiley & Sons
ISBN: 1118530381
Category : Technology & Engineering
Languages : en
Pages : 195
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Dedicated to the innovative design and use of ceramic materials for energy applications, this issue brings readers up to date with some of the most important research discoveries and new and emerging applications in the field. Contributions come from the proceedings of three symposia, as well as the European Union–USA Engineering Ceramics Summit. The three symposia are: Ceramics for Electric Energy Generation, Storage, and Distribution; Advanced Ceramics and Composites for Nuclear and Fusion Applications; and Advanced Materials and Technologies for Rechargeable Batteries. An abundance of charts, tables, and illustrations are included throughout.
Author:
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Languages : en
Pages : 17
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This is the final report on the high-temperature, gas-cooled reactor fuels project that was pursued at the Los Alamos Scientific Laboratory from early 1973 to mid-1977. This work followed two broad paths: (1) ZrC was utilized in various ways to improve the temperature capability of fission-product-retaining coated nuclear fuel particles and (2) techniques were developed for fabricating fuel rods by extruding the coated particles in high-density graphite matrices. The key to the ZrC coating project proved to be the ZrCl4 powder feeder development. The ZrCl4 reacted with H2 and CH4 or C3H6 to produce chemically vapor-deposited (CVD) ZrC. This CVD ZrC was used on nuclear fuel particles for replacing the SiC in the TRISO-II design, for making ZrC-C graded coats, and for making ZrC-alloyed isotropic carbon coats. Fuel particles with all of the described coats have been very successful in high-temperature, full-fluence (8 x 1021 n cm−2) irradiation tests. High-temperature diffusion experiments with cesium showed that ZrC is comparable to SiC as a diffusion barrier. Extruded fuel rods with matrix densities of 1.6 to 1.7 g cm−3 and with up to 45-vol percent coated particles were made without difficulty. All fuel rods tested in high-temperature, full-fluence irradiations had excellent results.
Author: Chengmin Liu
Publisher: Springer Nature
ISBN: 9811988994
Category : Science
Languages : en
Pages : 1260
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This is the third in a series of three volumes of proceedings of the 23rd Pacific Basin Nuclear Conference (PBNC 2022) which was held by Chinese Nuclear Society. As one in the most important and influential conference series of nuclear science and technology, the 23rd PBNC was held in Beijing and Chengdu, China in 2022 with the theme “Nuclear Innovation for Zero-carbon Future”. For taking solid steps toward the goals of achieving peak carbon emissions and carbon neutrality, future-oriented nuclear energy should be developed in an innovative way for meeting global energy demands and coordinating the deployment mechanism. It brought together outstanding nuclear scientists and technical experts, senior industry executives, senior government officials and international energy organization leaders from all across the world. The proceedings highlight the latest scientific, technological and industrial advances in Nuclear Safety and Security, Operations and Maintenance, New Builds, Waste Management, Spent Fuel, Decommissioning, Supply Capability and Quality Management, Fuel Cycles, Digital Reactor and New Technology, Innovative Reactors and New Applications, Irradiation Effects, Public Acceptance and Education, Economics, Medical and Biological Applications, and also the student program that intends to raise students’ awareness in fully engaging in this career and keep them updated on the current situation and future trends. These proceedings are not only a good summary of the new developments nuclear science and technology, but also a useful guideline for the researchers, engineers and graduate students.
