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Author: Publisher: ISBN: Category : Languages : en Pages : 13
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 13
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 13
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 11
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 12
Book Description
The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm3 and 3.8 gU/cm3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.
Author: Publisher: ISBN: Category : Languages : en Pages : 17
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 14
Book Description
The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about (approximately)50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm3 is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm3 is needed to match the cycle length of the HEU (36%) core.
Author: Publisher: ISBN: Category : Languages : en Pages : 9
Book Description
A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept.
Author: Russian Academy of Sciences Publisher: National Academies Press ISBN: 0309253209 Category : Science Languages : en Pages : 126
Book Description
Highly enriched uranium (HEU) is used for two major civilian purposes: as fuel for research reactors and as targets for medical isotope production. This material can be dangerous in the wrong hands. Stolen or diverted HEU can be used-in conjunction with some knowledge of physics-to build nuclear explosive devices. Thus, the continued civilian use of HEU is of concern particularly because this material may not be uniformly well-protected. To address these concerns, the National Research Council (NRC) of the U.S. National Academies and the Russian Academy of Sciences (RAS) held a joint symposium on June 8-10, 2011. Progress, Challenges, and Opportunities for Converting U.S. and Russian Research Reactors summarizes the proceedings of this joint symposium. This report addresses: (1) recent progress on conversion of research reactors, with a focus on U.S.- and R.F.-origin reactors; (2) lessons learned for overcoming conversion challenges, increasing the effectiveness of research reactor use, and enabling new reactor missions; (3) future research reactor conversion plans, challenges, and opportunities; and (4) actions that could be taken by U.S. and Russian organizations to promote conversion. The agenda for the symposium is provided in Appendix A, biographical sketches of the committee members are provided in Appendix B, and the report concludes with the statement of task in Appendix C.
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Author: Russian Academy of Sciences