Last edited by Galkis
Friday, November 20, 2020 | History

2 edition of Analysis of mixed oxide fuel irradiated in EBR-II found in the catalog.

Analysis of mixed oxide fuel irradiated in EBR-II

L. D Scott

Analysis of mixed oxide fuel irradiated in EBR-II

measured vs. predicted burnup

by L. D Scott

  • 191 Want to read
  • 24 Currently reading

Published by Dept. of Energy, for sale by the National Technical Information Service in [Washington], Springfield, Va .
Written in English

    Subjects:
  • Fuel burnup (Nuclear engineering),
  • Irradiation,
  • Oxides

  • Edition Notes

    StatementL. D. Scott, D. S. Dutt, D. R. Wilson, Hanford Engineering Develoopment Laboratory ; prepared for the U.S. Department of Energy
    SeriesHEDL-TME ; 77-60
    ContributionsDutt, D. S., joint author, Wilson, D. R. 1936- joint author, United States. Dept. of Energy, Hanford Engineering Development Laboratory
    The Physical Object
    Paginationi, 13 leaves :
    Number of Pages13
    ID Numbers
    Open LibraryOL14879951M

    This practice describes the procedure for preparing nuclear-grade uranium dioxide (UO 2) or mixed uranium-plutonium dioxide (MOX or (U,Pu)O 2)), sintered and non-irradiated pellets for subsequent microstructural analysis (hereafter referred to as ceramographic examination). "TEM Analysis of Irradiated Mixed-oxide Fuel" Assel Aitkaliyeva, Riley Parrish, Jason Harp, American Nuclear Society Student Conference April , () NSUF Articles: DOE-NE Awards 19 RTE Projects - New projects total approximately $K Thursday, February 6, - Announcement, Calls and Awards, Newsletter, News Release. model LWR fuel behavior under DBAs and severe accidents • Development and licensing of reduced-enrichment fuels for research reactors • Testing and evaluation of a variety of fuel types, including metallic, oxide, nitride, carbide, and dispersion fuels • Expertise in fuel fabrication, irradiation performance, safety analysis and.


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Analysis of mixed oxide fuel irradiated in EBR-II by L. D Scott Download PDF EPUB FB2

Get this from a library. Analysis of mixed oxide fuel irradiated in EBR-II: measured vs. predicted burnup. [L D Scott; D S Dutt; D R Wilson; United States. Department of Energy.; Hanford Engineering Development Laboratory.]. The irradiation programs for mixed oxide fuels in the s and early s in EBR-II were extensive.

In addition to the understanding of the fuel pellet behavior during irradiation, all the previously described irradiation effects on the cladding impacted fuel pin by:   Analyses of fuel pin irradiation behavior were carried out on the MOX fuel pins used in irradiation tests in the Joyo, EBR-II and Phenix reactors.

Typical characteristics of the irradiation tests are summarized in Table 2. Fuel pellets for these irradiation test pins were solid by: 7. The MOX (mixed oxide) fuels, wt% (analytical value is wt%) plutonium in or wt% enriched uranium were irradiated up to GWd/t in the MARK-II core of JOYO.

The irradiated MOX fuels were dissolved in 8 M nitric acid by:   The melting (solidus) temperatures of irradiated mixed oxide fuels were measured and the compositions of the fuels on the temperature measurement date were Analysis of mixed oxide fuel irradiated in EBR-II book.

The fuels contained about 29wt% Pu initially and were irradiated up to GWd/t in the experimental fast reactor JOYO. The formation and consequences of fuel-sodium reaction product (FSRP) in mixed-oxide fuel pins that were irradiated in EBR-II are described.

These results indicated that the amount of FSRP that. @article{osti_, title = {Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor}, author = {Chang, G S and Ryskamp, J M}, abstractNote = {An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory.

In this system, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. It was reported that MA-containing influences up to MA content less than ∼3 wt.% of the heavy metal amount on thermal fuel properties, i.e.

melting temperature and thermal conductivity, would be slight. The accumulated amount of spent fuel, whether normal uranium based fuel or irradiated mixed oxide, is stored either in the cooling pond or at the interim storage.

It is convenient to define auxiliary storage variables X U S (T) and X M S (T) to combine the total amounts of uranium based irradiated fuel and of irradiated mixed oxide, respectively. The burnup of fuel pins in the subassemblies irradiated at the range from to %FIMA in the JOYO MK-II core were measured by the isotope dilution analysis.

For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively.

Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN program language.

The mechanical analysis module implements the LIFE algorithm. 1. Introduction. The concept of an axially heterogeneous uranium plutonium mixed-oxide (MOX) core is a promising idea for liquid metal-cooled fast reactor (FR) cores because it flattens axial power distribution with the use of the internal blankets of the depleted uranium dioxide (UO 2), thereby mitigating void is also effective in reducing the neutron-irradiation-damage of the.

Fuel restructuring is also modeled, and includes the effects of porosity migration, irradiation-induced fuel densification and grain growth. (cont.) The FEAST-OXIDE predictions has been compared to the available FFTF, EBR-II and JOYO databases, and the agreement between the code and data was found to be satisfactory.

An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed.

The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm.

These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and. The character and extent of fuel/cladding chemical interaction (FCCI) have been established for mixed uranium-plutonium oxide, (U,Pu)O/sub 2/, fuels irradiated in Experimental Breeder Reactor-II.

Examples of fuel rod performance predictions are shown. Introduction 1. Analysis of fuel pin performance during transient overpower and/or undercooling events requires the use of some form of cladding damage correlation which includes the effects of neutron irradiation and fuel on cladding mechanical properties.

The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors.

As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of MWd/kg heavy metal. The US experience with mixed oxide, metal, and mixed carbide fuels is substantial, comprised of irradiation of over 50 MOX rods, over metal rods, and mixed carbide rods, in EBR-II.

Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of K and K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that. Much consideration is being given to the analysis of uranium and plutonium oxide samples and uranium-plutonium mixed fuels irradiated in the BOR using mass-spectrometric (isotope dilution method) and radiometric techniques.

The results of uranium and plutonium determination by. Fuel restructuring is also modeled, and includes the effects of porosity migration, irradiation-induced fuel densification and grain growth.

en_US: ct (cont.) The FEAST-OXIDE predictions has been compared to the available FFTF, EBR-II and JOYO databases, and the agreement between the code and data was found to be satisfactory. Conference: An expert system for fuel failure diagnosis in EBR The CDE clearly demonstrates that mixed-oxide fuel can achieve burnups in excess of MWd/kgM and fast fluences in excess of 30 {times} 10{sup 22} n/cm{sup 2} using the very low swelling.

Abstract The FFTF fuel pin design analysis is shown to be conservative through comparison with pin irradiation experience in EBR-II. This comparison shows that the actual lifetimes of EBR-II fuel pins are either greater t MWd/MTM or greater than the calculated allowable lifetimes based on.

The irradiation testing mission was directed primarily at development of oxide fuel for FFTF and CRBR, but also important was improvement of the performance of EBR-II metal fuel. In addition, nitride and carbide fuels were tested. EBR-II metal driver fuel was significantly improved over the course of the 30 year operating life of the reactor (6.

The U.S. Department of Energy's Office of Scientific and Technical Information. Burn-up measurement on an irradiated mixed oxide (MOX) test fuel pellet was carried out through measurements on the dissolver solution by HPLC-Thermal Ionization Mass Spectrometric (TIMS) technique.

The studies carried out using HPLC as well as TIMS for quantification of burn-up value are described. While in one case, both the separation and determination of elements of interest (U, Pu.

Extended overpower transient testing of LMFBR oxide pins in EBR-II Authors: H. Tsai, L. Neimark, S. Tani, I. Shibahara Source: Nuclear fuel performance, 1 Jan (1: –). Mixed Oxide Fuel (Mox) Exploitation and Destruction in Power Reactors. Editors (view affiliations) The 25 presentations in this book give an impressive overview of MOX technology.

The following issues are covered: an up to date report on the disposition of ex-weapons Pu in Russia; an analysis of safety features of MOX fuel configurations of.

A test containing 19 mixed-oxide fuel pins was operated in the Experimental Breeder Reactor II (EBRII) at peak cladding temperatures near °C. Two test pins that had been designed to fail at ∼5 at.

% burnup and two low-burnup environmental pins failed and then were operated in the run beyond cladding breach mode for 22 days. irradiation conducted by the EBR-II Project of Argonne National Laboratory. The three pin sodium-cooled fuel assemblies were desig-nated PNL-3, PNL-4 and PNL-5 and were similar in that they con-tained near-equal fuel smear densities of mixed oxide (85–88%), identical plenum volumes ( cc) and were irradiated at similar.

Testing of LMFBR Oxide Pins in EBR-II ibid 4 W. Lehto The EBR-II Breached Fuel Test Facility Proceedings of International Meeting on Fast Reactor Safety V 5 K.

Gross, and R. Strain Delayed-Neutron Signal Characterization in a Fast. An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST).

FEAST has several modules working in coupled form with an explicit numerical algorithm. Supplement Analysis 2 PURPOSE To implement the MOX fuel disposition alternatives considered in the Storage and Disposition PEIS and the SPD EIS, fuel lead assemblies need to be fabricated, irradiated, and inspected to support U.S.

Nuclear Regulatory Commission (NRC) licensing activities and fuel. Mixed oxide, or MOX fuel, is a blend of plutonium and natural or depleted uranium which behaves similarly (though not identically) to the enriched uranium feed for which most nuclear reactors were designed.

MOX fuel is an alternative to low enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation. Some concern has been expressed that used MOX. During post irradiation examination of a mixed-oxide fuel pin, the molybdenum contents of the fuel matrix and of the metallic inclusions are determined by electron microprobe analysis.

At a point in the fuel pin where the temperature is estimated to have been °K during irradiation, the fuel matrix contains mole % MoO 2 and the. The MCNP Monte Carlo code was applied to accurately calculate the fission rate of mixed oxide (MOX) fuel used in a fast reactor.

The test fuel pins to be analyzed were irradiated in the experimental fast reactor JOYO. The heterogeneous structure of the irradiation test subassembly was modeled in the MCNP calculation. The neutron flux distribution in the subassembly was calculated with the.

Radiation Effects on Metal Fuel 75 Radiation Effects on Mixed-Oxide Fuel iv Figure The effect of irradiation on 8 — Pu metal fuel 77 Figure The effect of irradiation on 20 — Pu metal fuel 78 diagnosing failed fuel elements in EBR-II.

Before the scope of this project is outlined. Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel Paul Edelmann Follow this and additional works at: This Dissertation is brought to you for free and open access by the Engineering ETDs at UNM.

The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!.

Oxide dispersion strengthened (ODS) steel cladding on mixed-oxide fuel pins offers high-temperature strength and creep rupture resistance properties for a long-life fast reactor.

The fabrication and irradiation of the large diameter ODS steel cladding with high-density mixed-oxide pellets were performed at the Argonne National Laboratory.irradiated LEU fuel and that from irradiated mixed uranium-plutonium oxide (MOX) fuel. Nothing contained in this document may be construed as having the force and effect of NRC regulations (except where the regulations are cited), or as indicating that applications supported by safety analyses and prepared in accordance with RG EBR-II was the backbone of the U.S.

breeder reactor effort from towhen research was terminated. The EBR-II accommodated as many as 65 experimental subassemblies at one time for irradiation and operational reliability tests. EBR-II also performed o irradiation tests.