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Download or read book Advanced Test Reactor Core Modeling Update Project written by and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Advanced Test Reactor Core Modeling Update Project: Annual Report for Fiscal 2014.
Download or read book Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010 written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V & V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V & V, within the next 3-4 years via the ATR Core Modeling and Simulation and V & V Update (or "Core Modeling Update") Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).
Download or read book Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013 written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V & V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).
Download or read book Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011 written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V & V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V & V, within the next 3-4 years via the ATR Core Modeling and Simulation and V & V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.
Download or read book Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012 written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V & V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.
Author :
Publisher :
ISBN 13 :
Total Pages : 232 pages
Book Rating : 4.3/5 (91 download)
Download or read book The Advanced Test Reactor (ATR) written by and published by . This book was released on 1960 with total page 232 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : Chad L. Pope
Publisher : BoD – Books on Demand
ISBN 13 : 1839699396
Total Pages : 196 pages
Book Rating : 4.8/5 (396 download)
Download or read book Nuclear Reactors written by Chad L. Pope and published by BoD – Books on Demand. This book was released on 2022-09-14 with total page 196 pages. Available in PDF, EPUB and Kindle. Book excerpt: Worldwide interest in nuclear reactors continues to increase and significant focus has been placed on advanced nuclear reactors intended to produce electricity and process heat. However, there is limited literature on the importance of research reactors and certain specialized reactor analysis topics. Thus, this book addresses these topics over three sections: “Nuclear Reactors for Spacecraft Propulsion”, “Research Reactors”, and “Select Reactor Analysis Techniques”. It provides detailed information on the use of nuclear reactors for spacecraft propulsion, presents research conducted on reactors in Idaho, USA, and discusses reactor analysis topics such as cyber-informed engineering for nuclear reactor digital instrumentation and control, the effect of plenum gas on fuel temperature, and more.
Author : Samuel E. Bays
Publisher :
ISBN 13 :
Total Pages : 0 pages
Book Rating : 4.:/5 (139 download)
Download or read book Core Reload Analysis Techniques in the Advanced Test Reactor written by Samuel E. Bays and published by . This book was released on 2019 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Since becoming a national user facility in 2007, the type of irradiation campaigns the Advanced Test Reactor (ATR) supports has become much more diverse and complex. In prior years, test complexity was limited by the computational ability to analyze the tests,Äô influence on the fuel. Large volume tests are irradiated in flux traps which are designed to receive excess neutrons from the surrounding fuel elements. Typically, fuel elements drive the test conditions, not vice versa. The computational tool, PDQ, was used for core physics analysis for decades. The PDQ code was adequate so long as the diffusion approximation between test and fuel element remained valid. This paradigm changed with the introduction of the Ki-Jang Research Reactor,ÄîFuel Assembly Irradiation (KJRR-FAI) in 2015. The KJRR-FAI was a prototypic fuel element for the KJRR research reactor project in the Republic of Korea. The KJRR-FAI irradiation presented multiple modeling and simulation challenges for which PDQ was ill suited. To demonstrate that the KJRR-FAI could be irradiated and meet safety requirements, the modern neutron transport codes, HELIOS and MCNP, were extensively verified and validated to replace PDQ. The hybrid 3D/2D methodology devised with these codes made analysis of the ATR with KJRR-FAI possible. The KJRR-FAI was irradiated in 2015-2016.
Download or read book Scoping Analysis of the Advanced Test Reactor Using SN2ND. written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature requirements for the ATR, and also to demonstrate the feasibility of performing this analysis with a deterministic transport code capable of modeling heterogeneous geometries. The work performed indicates that a minimum of 260,000 linear finite elements combined with a L3T11 cubature (96 angles on the sphere) is required for both eigenvalue and flux convergence of the ATR. A critical finding was that the fuel meat and water channels must each be meshed with at least 3 'radial zones' for accurate flux convergence. A small number of 3D calculations were also performed to show axial mesh and eigenvalue convergence for a full core problem. Finally, a brief analysis was performed with different cross sections sets generated from DRAGON and SCALE, and the findings show that more effort will be required to improve the multigroup cross section generation process. The total number of degrees of freedom for a converged 27 group, 2D ATR problem is ≈340 million. This number increases to ≈25 billion for a 3D ATR problem. This scoping study shows that both 2D and 3D calculations are well within the capabilities of the current SN2ND solver, given the availability of a large-scale computing center such as BlueGene/P. However, dynamics calculations are not realistic without the implementation of improvements in the solver.
Download or read book A Spectral Verification of the HELIOS-2 Lattice Physics Code written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Core modeling of the Advanced Test Reactor (ATR) at INL is currently undergoing a significant update through the Core Modeling Update Project1. The intent of the project is to bring ATR core modeling in line with today's standard of computational efficiency and verification and validation practices. The HELIOS-2 lattice physics code2 is the lead code of several reactor physics codes to be dedicated to modernize ATR core analysis. This presentation is concerned with an independent verification of the HELIOS-2 spectral representation including the slowing down and thermalization algorithm and its data dependency. Here, we will describe and demonstrate a recently developed simple cross section generation algorithm based entirely on analytical multigroup parameters for both the slowing down and thermal spectrum. The new capability features fine group detail to assess the flux and multiplication factor dependencies on cross section data sets using the fundamental infinite medium as an example.
Author :
Publisher :
ISBN 13 :
Total Pages : 206 pages
Book Rating : 4.3/5 (91 download)
Download or read book Design Study of an Advanced Test Reactor written by and published by . This book was released on 1957 with total page 206 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : J. O. Howard
Publisher :
ISBN 13 :
Total Pages : 206 pages
Book Rating : 4.:/5 (319 download)
Download or read book Advanced Test Reactor written by J. O. Howard and published by . This book was released on 1963 with total page 206 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.
Author :
Publisher :
ISBN 13 :
Total Pages : 490 pages
Book Rating : 4.0/5 ( download)
Download or read book Energy Research Abstracts written by and published by . This book was released on 1993 with total page 490 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Advanced Test Reactor Design Basis Reconstitution Project Issue Resolution Process written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Advanced Test Reactor (ATR) Design Basis Reconstitution Program (DBRP) is a structured assessment and reconstitution of the design basis for the ATR. The DBRP is designed to establish and document the ties between the Document Safety Analysis (DSA), design basis, and actual system configurations. Where the DBRP assessment team cannot establish a link between these three major elements, a gap is identified. Resolutions to identified gaps represent configuration management and design basis recovery actions. The proposed paper discusses the process being applied to define, evaluate, report, and address gaps that are identified through the ATR DBRP. Design basis verification may be performed or required for a nuclear facility safety basis on various levels. The process is applicable to large-scale design basis reconstitution efforts, such as the ATR DBRP, or may be scaled for application on smaller projects. The concepts are applicable to long-term maintenance of a nuclear facility safety basis and recovery of degraded safety basis components. The ATR DBRP assessment team has observed numerous examples where a clear and accurate link between the DSA, design basis, and actual system configuration was not immediately identifiable in supporting documentation. As a result, a systematic approach to effectively document, prioritize, and evaluate each observation is required. The DBRP issue resolution process provides direction for consistent identification, documentation, categorization, and evaluation, and where applicable, entry into the determination process for a potential inadequacy in the safety analysis (PISA). The issue resolution process is a key element for execution of the DBRP. Application of the process facilitates collection, assessment, and reporting of issues identified by the DBRP team. Application of the process results in an organized database of safety basis gaps and prioritized corrective action planning and resolution. The DBRP team follows the ATR DBRP issue resolution process which provides a method for the team to promptly sort and prioritize questions and issues between those that can be addressed as a normal part of the reconstitution project and those that are to be handle as PISAs. Presentation of the DBRP issue resolution process provides an example for similar activities that may be required at other facilities within the Department of Energy complex.
Download or read book Deterministic Modeling of the High Temperature Test Reactor written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green's Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2-3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.
Author : Bruce G. Schnitzler
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (316 download)
Download or read book VALIDATION OF IMPROVED 3D ATR MODEL. written by Bruce G. Schnitzler and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A full-core Monte Carlo based 3D model of the Advanced Test Reactor (ATR) was previously developed. [1] An improved 3D model has been developed by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) to eliminate homogeneity of fuel plates of the old model, incorporate core changes into the new model, and to validate against a newer, more complicated core configuration. This new 3D model adds capability for fuel loading design and azimuthal power peaking studies of the ATR fuel elements.
