skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report

Abstract

Characterization data is reported for sodium bonded Experimental Breeder Reactor II (EBR-II) fuel which had been stored underwater in containers since the late 1970's. Sixteen stainless steel storage containers were retrieved from the ICPP-603 storage pool at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. Ten of the containers had leaked water due to improper sealing. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide particulate formed and filled the bottom of the container. Headspace gas analysis determined that greater than 99% hydrogen was present. Cesium-137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a caustic solution of NaOH.

Authors:
Publication Date:
Research Org.:
Argonne National Lab., IL (US)
Sponsoring Org.:
US Department of Energy (US)
OSTI Identifier:
755842
Report Number(s):
ANL-00/9
TRN: US0003626
DOE Contract Number:
W-31109-ENG-38
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 17 Apr 2000
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; EBR-2 REACTOR; SPENT FUELS; SPENT FUEL STORAGE; SPENT FUEL CASKS; LEAKS; LEACHING; CORROSION; URANIUM ALLOYS; URANIUM OXIDES; GAS ANALYSIS; HYDROGEN; CESIUM 137; SODIUM; SODIUM HYDROXIDES

Citation Formats

Pahl, R. G. Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report. United States: N. p., 2000. Web. doi:10.2172/755842.
Copy to clipboard
Pahl, R. G. Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report. United States. doi:10.2172/755842.
Copy to clipboard
Pahl, R. G. Mon . "Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report". United States. doi:10.2172/755842. https://www.osti.gov/servlets/purl/755842.
Copy to clipboard
@article{osti_755842,
title = {Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report},
author = {Pahl, R. G.},
abstractNote = {Characterization data is reported for sodium bonded Experimental Breeder Reactor II (EBR-II) fuel which had been stored underwater in containers since the late 1970's. Sixteen stainless steel storage containers were retrieved from the ICPP-603 storage pool at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. Ten of the containers had leaked water due to improper sealing. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide particulate formed and filled the bottom of the container. Headspace gas analysis determined that greater than 99% hydrogen was present. Cesium-137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a caustic solution of NaOH.},
doi = {10.2172/755842},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Apr 17 00:00:00 EDT 2000},
month = {Mon Apr 17 00:00:00 EDT 2000}
}
Copy to clipboard
Technical Report:
View Technical Report (1.31 MB)
DOI:  10.2172/755842
Save / Share:
Export Metadata
Save to My Library
You must Sign In or Create an Account in order to save documents to your library.

  • The oxidation behavior of hydride-bearing corrosion products from uranium metal ZPPR fuel plates was studied in Ar-O{sub 2}, Ar-H{sub 2}O, Ar-O{sub 2}-H{sub 2}O, dry air, and moist air environments. Both isothermal and burning curve tests in the different environments were performed using a thermo-gravimetric analyzer. The effect of pre-oxidation in each environment on subsequent ignition temperature was investigated by performing burning curve tests on samples after isothermal oxidation. Low-temperature oxidation rates in Ar-O{sub 2} and dry air environments were identical. Oxidation rates in moist environments were slightly higher, but the difference was not statistically significant at 95% confidence. Oxygen contaminationmore » was suspected to have lowered rates measured in the Ar-H{sub 2}O environment. ignition temperatures measured in air were 10-15 C higher than those measured in Ar-20% O{sub 2}; the ignition temperatures of samples pre-oxidized in moist gas environments appeared to be slightly lower than those of unreacted samples at equivalent hydride fractions. Burning rates in all environments were linearly dependent on hydride surface area for surface areas less than 200 cm{sup 2}. Burning rates were constant at higher surface areas.« less

  • A study was made to define the type of fuel that is the primary source of radionuclides ''leaking'' into the fuel storage basin at the Idaho Chemical Processing Plant. The isotopic compositions of plutonium and uranium in sludge on the basin floor, in the fuel storage basin water, and in the various types of fuel, as indicated by analysis of process solutions, were used to determine that the stainless steel-type (EBR-II) fuel was the main ''leaking'' fuel.

  • ;
    OLDAGE is a computer program used to calculate the amount of spent fuel and/or HLW transferred to and from the various fuel cycle facilities (i.e., reactors, reprocessing, MRS, or repository) within the nuclear waste management system. OLDAGE analyzes the logistics of transporting and storing spent fuel and high-level waste among fuel cycle facilities using the priority rule of oldest fuel/waste first. The transfer of spent fuel/HLW from one location to the other is limited by the handling capacity of each location. Although the program has the potential for a wide range of applications, the purpose for its development was tomore » provide the material flow balance input for another program, CASMO. CASMO is used to calculate the total miles for one or more MRS sites based on the amount of fuel transferred into and out of the MRS, and the corresponding transportation costs for specific MRS sites. The results produced by executing OLDAGE provide a decision tool to aid in planning and scheduling spent fuel/HLW shipments, siting MRS facilities, determining at-reactor storage capacity requirements, handling rates, and needed facility start-up dates. 2 references, 2 figures.« less

  • A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closedmore » conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford« less

  • The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng.