Fuel Handling Accident Analysis for the University of Missouri Research Reactors High Enriched Uranium to Low Enriched Uranium Fuel Conversion Initiative PDF Download
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Author: Benjamin Rickman
Publisher:
ISBN:
Category :
Languages : en
Pages : 40
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Book Description
In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's per- formance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staat the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was estab- lished by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Eective Dose Equivalent (TEDE) values were calculated in addi- tion to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regula- tory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem,which is above the HEU MHA TEDE dose of 132 mrem. Since the new method- ology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecied dose.
Author: Benjamin Rickman
Publisher:
ISBN:
Category :
Languages : en
Pages : 40
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Book Description
In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's per- formance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staat the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was estab- lished by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Eective Dose Equivalent (TEDE) values were calculated in addi- tion to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regula- tory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem,which is above the HEU MHA TEDE dose of 132 mrem. Since the new method- ology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecied dose.
Author:
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Category :
Languages : en
Pages :
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This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program - College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U.S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program - College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U.S. government. In the framework of 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 LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).
Author: Wilson Cowherd
Publisher:
ISBN:
Category :
Languages : en
Pages : 63
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Book Description
As a part of the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) programs, the University of Missouri Research Reactor (MURR) intends to convert from highly-enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As a part of the conversion, the maximum hypothetical accident (MHA) analysis had to be done to determine dose consequences for both radiation workers and members of the public. For the dose to the radiation workers inside the containment building, the committed effective dose equivalent (CEDE) was 0.889 mrem and the total effective dose equivalent (TEDE) was 1403 mrem using the 10 CFR 20 default derived air concentration (DAC) and 403 mrem using the pre-2007 version of 10 CFR 835 DACs. For the dose to a person outside the exclusionary boundary area (EBA) of MURR, the CEDE was 0 mrem and the TEDE was 8.58 mrem using the 10 CFR 20 default DAC and 2.45 mrem using the pre-2007 version of 10 CFR 835 DACs. These doses to both the radiation workers and the general public are lower than the 10 CFR 20 guidance for dose consequences.
Author:
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Category :
Languages : en
Pages :
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Author:
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Category :
Languages : en
Pages :
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Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.
Author: International Atomic Energy Agency
Publisher: International Atomic Energy Agency
ISBN: 9789201450104
Category : Science
Languages : en
Pages : 108
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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.
Author:
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Category :
Languages : en
Pages :
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ISBN: 9780982800843
Category : Energy policy
Languages : en
Pages : 237
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Book Description
"In this analysis we have presented a method that provides insight into future fuel cycle alternatives by clarifying the complexity of choosing an appropriate fuel cycle in the context of the distribution of burdens and benefits between generations. The current nuclear power deployment practices, together with three future fuel cycles were assessed."--Page 227.
Author: Christopher M. Carroll
Publisher:
ISBN:
Category : Nuclear reactor accidents
Languages : en
Pages : 262
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Book Description
"The Safety Analysis Report, SAR, is very important to commercial and research nuclear reactors alike and should be considered a living document. The Nuclear Regulatory Commission has instituted changes over the past few years to rely more on the SAR for the processes of licensing, relicensing, and uprating in power. As result of these changes the NRC has streamlined the licensing processes such that the length of time expended for application approval is shorter and in addition, easier for facilities. The University of Missouri-Rolla Reactor, UMRR, is up for relicensing January 2005 and has begun to review and revise its SAR to update to current NRC standards. A preliminary investigation into the possibility of a power uprate has also begun. As a part of this investigation, PARET-ANL and CONVEC were used to analyze accident scenarios at the currently licensed power of 200 kW[subscript t] and at possible uprate powers of 400 kW[subscript t] and 500 kW[subscript t] . CONVEC is a natural convection modeling code and PARET is a code used to investigate reactor reactivity transients. Due to the NRC permitting credit to be taken for a reactor SCRAM function, the hypothetical severe accident scenarios have been evaluated with the use of the UMRR safety channel SCRAM set-point of 150% of licensed power. Several improbable, but significant accident scenarios are expected in the SAR. The first accident is an insertion of excess reactivity that assumes a fuel element being placed accidentally next to the UMRR core. For the second accident, a loss of coolant accident (LOCA) in which all coolant is assumed to drain was evaluated. As a final accident, a startup accident was analyzed in which the UMRR's control elements were assumed to uncontrollably withdraw from the core. Each of these accidents is important to analyze due to the risk of fuel cladding failure that could potentially result in the release of fission products. The analysis described in this thesis proves that the UMRR would maintain fuel integrity during the aforementioned accident scenarios and thus avert the release or fission products "--Abstract, leaf iii.
Author: International Atomic Energy Agency
Publisher:
ISBN:
Category : Business & Economics
Languages : en
Pages : 120
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Book Description
Provides a critical review of the thorium fuel cycle: potential benefits and challenges in the thorium fuel cycle, mainly based on the latest developments at the front end of the fuel cycle, applying thorium fuel cycle options, and at the back end of the thorium fuel cycle.
