Download or read book Neutronic Safety and Transient Analyses for Potential LEU Conversion of the IR 8 Research Reactor written by and published by . This book was released on 1999 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.
Download or read book Neutronic Safety Parameters and Transient Analyses for Potential LEU Conversion of the Budapest Research Reactor written by and published by . This book was released on 1999 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO2-Al dispersion fuel, and with LEU (19.75%)UO2-Al dispersion fuel that has a uranium density of about 2.5 g/cm3. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.
Download or read book A Neutronic Feasibility Study for LEU Conversion of the IR 8 Research Reactor written by and published by . This book was released on 1998 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU(90%), HEU(36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average 235U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm3 in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU(90%) IRT-3M FA and an LEU density of 3.7 g/cm3 is needed to match the cycle length of the HEU(36%) IRT-3M FA.
Download or read book Neutronics Methods for Transient and Safety Analysis of Fast Reactors written by Marchetti, Marco and published by KIT Scientific Publishing. This book was released on 2017-03-02 with total page 164 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Neutronic Analyses for HEU to LEU Fuel Conversion of the Massachusetts Institute of Technology written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.
Download or read book Neutronics Steady state and Transient Analyses for the Poland MARIA Reactor for Irradiation Testing of LEU Lead Test Fuel Assemblies from CERCA written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.
Download or read book TREAT Transient Analysis Benchmarking for the HEU Core written by and published by . This book was released on 2014 with total page 45 pages. Available in PDF, EPUB and Kindle. Book excerpt: This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used to determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term "reported" values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core's performance.
Download or read book Analysis of the TREAT LEU Conceptual Design written by and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy's Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration's Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.
Download or read book A Neutronic Feasibility Study for LEU Conversion of the Brookhaven Medical Research Reactor BMRR written by and published by . This book was released on 1998 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt: A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm3 or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.
Download or read book A Neutronic Feasibility Study for LEU Conversion of the Budapest Research Reactor written by and published by . This book was released on 1998 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: A neutronic feasibility study for conversion of the Budapest Research Reactor (BRR) from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with the KFKI Atomic Energy Research Institute in Hungary. Comparisons were made of the reactor performance with the current HEU (36%) fuel and with a proposed LEU (19.75%) fuel. Cycle lengths, thermal neutron fluxes, and rod worths were calculated in equilibrium-type cores for each type of fuel. Relative to the HEU fuel, the LEU fuel has up to a 50% longer fuel cycle length, but a 7-10% smaller thermal neutron flux in the experiment locations. The rod worths are smaller with the LEU fuel, but are still large enough to easily satisfy the BRR shutdown margin criteria. Irradiation testing of four VVR-M2 LEU fuel assemblies that are nearly the same as the proposed BRR LEU fuel assemblies is currently in progress at the Petersburg Nuclear Physics Institute.
Download or read book Neutronic Feasibility Studies for LEU Conversion of the HFR Petten Reactor written by and published by . This book was released on 2000 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Design and safety analyses to determine an optimum LEU fuel assembly design using U3Si2-Al fuel with up to 4.8 g/cm3 for conversion of the HFR Petten reactor were performed by the RERTR program in cooperation with the Joint Research Centre and NRG. Credibility of the calculational methods and models were established by comparing calculations with recent measurements by NRG for a core configuration setup for this purpose. This model and methodology were then used to study various LEU fissile loading and burnable poison options that would satisfy specific design criteria.
Download or read book Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study written by G. S. Chang and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2-s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequentoperating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile aresimilar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.
Download or read book Applications of Research Reactors written by International Atomic Energy Agency and published by International Atomic Energy Agency. This book was released on 2014 with total page 108 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication is a comprehensive study that reviews the current situation in a great number of applications of research reactors. It revises the contents of IAEA TECDOC-1234, The Applications of Research Reactors, giving detailed updates on each field of research reactor uses worldwide. Reactors of all sizes and capabilities can benefit from the sharing of current practices and research enabled via this updated version, which describes the requirements for practicing methods as diverse as neutron activation analysis, education and training, neutron scattering and neutron imaging, silicon doping and radioisotope production, material/fuel irradiation and testing, and some others. Many underutilised research reactors can learn how to diversify their technical capabilities, staff and potential commercial partners and users seeking research reactor services and products. The content of the publication has also been strengthened in terms of current issues facing the vast majority of research reactors by including sections describing user and customer relations as well as strategic planning considerations.
Download or read book Neutronic Analysis for the UVAR Reactor HEU to LEU Conversion Project written by David W. Freeman and published by . This book was released on 1990 with total page 192 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Neutronics Studies on the NIST Reactor Using the GA LEU Fuel written by Kyle Anthony Britton and published by . This book was released on 2018 with total page 60 pages. Available in PDF, EPUB and Kindle. Book excerpt: The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5x1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR's heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR's current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel.
Download or read book Planning Document for an NBSR Conversion Safety Analysis Report written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted. This report follows the recommended format and content from the NRC codified in NUREG-1537,?Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,? Chapter 18,?Highly Enriched to Low-Enriched Uranium Conversions.? The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.
