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Agr 1 Post Irradiation Examination Final Report
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Book Synopsis AGR-1 Post Irradiation Examination Final Report by :
Download or read book AGR-1 Post Irradiation Examination Final Report written by and published by . This book was released on 2015 with total page 132 pages. Available in PDF, EPUB and Kindle. Book excerpt: The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building on the as-fabricated fuel characterization and irradiation data. In addition to the extensive volume of results generated, the work also resulted in a number of novel analysis techniques and lessons learned that are being applied to the examination of fuel from subsequent TRISO fuel irradiations. This report provides a summary of the results obtained as part of the AGR-1 PIE campaign over its approximately 5-year duration.
Book Synopsis AGR-2 Irradiation Test Final As-Run Report by :
Download or read book AGR-2 Irradiation Test Final As-Run Report written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47́1025 n/m2 (E>0.18 MeV) for UCO fuel, and from 3.05 to 3.53́1025 n/m2 (E>0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2́10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.
Book Synopsis AGR-2 Irradiation Test Final As-Run Report, Rev 2 by :
Download or read book AGR-2 Irradiation Test Final As-Run Report, Rev 2 written by and published by . This book was released on 2014 with total page 103 pages. Available in PDF, EPUB and Kindle. Book excerpt: This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 1025 n/m2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 1025 n/m2 (E >0.18 MeV) for UO2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10-6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10-6. In the UO2 capsule (Capsule 3), the R/B values during the first three cycles were below 10-7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.
Book Synopsis AGR-1 Data Qualification Interim Report by :
Download or read book AGR-1 Data Qualification Interim Report written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.
Book Synopsis AGR-1 Data Qualification Report by :
Download or read book AGR-1 Data Qualification Report written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: 1. Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). 2. Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. 3. FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report currently in review. 4. ATR Operating Conditions Data. Data for all AGR-1 cycles have been stored and capture tested. These data, which come from outside the VHTR program, are assumed to be qualified by ATR quality control procedures. 5. Neutronics and Thermal Modeling Data. NDMAS processing is in progress. These data will not be qualified by NDMAS.
Book Synopsis PIE on Safety-Tested AGR-1 Compact 5-1-1 by :
Download or read book PIE on Safety-Tested AGR-1 Compact 5-1-1 written by and published by . This book was released on 2015 with total page 42 pages. Available in PDF, EPUB and Kindle. Book excerpt: Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.
Book Synopsis Comprehensive Nuclear Materials by :
Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field
Book Synopsis Data Compilation for AGR-1 Variant 1 Compact Lot LEU01-47T-Z by :
Download or read book Data Compilation for AGR-1 Variant 1 Compact Lot LEU01-47T-Z written by and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This document is a compilation of characterization data for the AGR-1 variant 1 compact lot LEU01-47T-Z. The compacts were produced by ORNL for the ADvanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-47T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an ≈ 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrcoarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified at LEU01-?? (where?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 1 coated particle composite LEU01-47T can be found in ORNL/TM-2006/020. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel Materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.
Book Synopsis Comprehensive Nuclear Materials by : Todd R Allen
Download or read book Comprehensive Nuclear Materials written by Todd R Allen and published by Elsevier. This book was released on 2011-05-12 with total page 3552 pages. Available in PDF, EPUB and Kindle. Book excerpt: Comprehensive Nuclear Materials, Five Volume Set discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerators, and for diverse functions in fuels, cladding, moderator and control materials, structural, functional, and waste materials. The work addresses the full panorama of contemporary international research in nuclear materials, from Actinides to Zirconium alloys, from the worlds' leading scientists and engineers. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environment Fully integrated with F-elements.net, a proprietary database containing useful cross-referenced property data on the lanthanides and actinides Details contemporary developments in numerical simulation, modelling, experimentation, and computational analysis, for effective implementation in labs and plants
Book Synopsis Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR-1 Irradiation Experiment by :
Download or read book Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR-1 Irradiation Experiment written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.
Book Synopsis Nuclear Energy for Hydrogen Production by : Karl Verfondern
Download or read book Nuclear Energy for Hydrogen Production written by Karl Verfondern and published by Forschungszentrum Jülich. This book was released on 2007 with total page 199 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2 by :
Download or read book PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2 written by and published by . This book was released on 2016 with total page 63 pages. Available in PDF, EPUB and Kindle. Book excerpt: Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program [Petti et al. 2010]. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.
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Book Synopsis Scientific and Technical Aerospace Reports by :
Download or read book Scientific and Technical Aerospace Reports written by and published by . This book was released on 1981 with total page 788 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Monthly Catalogue, United States Public Documents by :
Download or read book Monthly Catalogue, United States Public Documents written by and published by . This book was released on 1982 with total page 1094 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Government Reports Announcements by :
Download or read book Government Reports Announcements written by and published by . This book was released on 1974 with total page 864 pages. Available in PDF, EPUB and Kindle. Book excerpt:
