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Title: Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

Abstract

The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.

Authors:
Publication Date:
Research Org.:
Savannah River Site (US)
Sponsoring Org.:
US Department of Energy (US)
OSTI Identifier:
14491
Report Number(s):
WSRC-TR-97-0153
TRN: US0111055
DOE Contract Number:
AC09-96SR18500
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 21 Oct 1999
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; EVALUATION; SPENT FUELS; FUEL STORAGE POOLS; FISSION PRODUCT RELEASE; FUEL CANS; ALUMINIUM ALLOYS

Citation Formats

Sindelar, R.L. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage. United States: N. p., 1999. Web. doi:10.2172/14491.
Sindelar, R.L. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage. United States. doi:10.2172/14491.
Sindelar, R.L. Thu . "Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage". United States. doi:10.2172/14491. https://www.osti.gov/servlets/purl/14491.
@article{osti_14491,
title = {Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage},
author = {Sindelar, R.L.},
abstractNote = {The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.},
doi = {10.2172/14491},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Oct 21 00:00:00 EDT 1999},
month = {Thu Oct 21 00:00:00 EDT 1999}
}

Technical Report:

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  • Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements havemore » been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.« less
  • A nuclear power reactor operator, confronted with rising spent fuel inventories that would soon exceed storage capacity, has several options to remedy this situation: transshipment, reracking of existing at-reactor storage basins (ARB), new ARB's, and away-from-reactor (AFR) basins. This report examines the latter two cases where a macro-economic comparison of each storage system is developed. The AFR storage system proved to be significantly more economical. This analysis indicates that 46 ARB's would be needed to meet storage requirements compared to only three AFR's. The total discounted costs associated with building new ARB's and AFR's are $4.7 billion and $1.2 billionmore » respectively.« less
  • The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a rangemore » of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.« less
  • This report delineates the results obtained from laboratory testing of IONISIV{reg_sign} IE-95 to determine the efficacy of the zeolite for the removal of radioactive cesium from the KE Basin water prior to transport to the Effluent Treatment Facility, as described in RPP-PLAN-36158, IONSIV{reg_sign} IE-95 Studies for the removal of Radioactive Cesium from KE Basin Spent Nuclear Fuel Pool during Decommissioning Activities. The spent nuclear fuel was removed from KE Basin and the remaining sludge was layered with a grout mixture consisting of 26% Lehigh Type I/II portland cement and 74% Boral Mohave type F fly ash with a water-to-cement ratiomore » of 0.43. The first grout pour was added to the basin floor to a depth of approximately 14 in. covering an area of 12,000 square feet. A grout layer was also added to the sludge containers located in the attached Weasel and Technical View pits.« less
  • This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, andmore » point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs.« less