Bounding criticality safety analyses for shipments of unconfigured spent nuclear fuel
In November 1996, a request was made to the US Department of Energy for a waiver for three shipments of spent nuclear fuel (SNF) from Oak Ridge National Laboratory (ORNL) to the Savannah River Site (SRS) in the US NRC certified BMI-1 cask (CoC 5957). Although the post-irradiation fissile mass (based on chemical assays) in each shipment was less than 800 g, a criticality safety analysis was needed because the pre-irradiation mass exceeded 800 g, the fissile material limit in the CoC. The analyses were performed on SNF consisting of aluminum-clad U{sub 3}O{sub 8}, UAl{sub x}, and U{sub 3}Si{sub 2} plates, fragments and pieces that had been irradiated at ORNL during the Reduced Enrichment Research and Test Reactor Program of the 1980s. The highlights of the approach used to analyze this unique SNF and the benefits of the waiver are presented in this paper.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Assistant Secretary for Management and Administration, Washington, DC (United States)
- DOE Contract Number:
- AC05-96OR22464
- OSTI ID:
- 663352
- Report Number(s):
- ORNL/CP-96300; CONF-980906-; ON: DE98005754; TRN: 99:000068
- Resource Relation:
- Conference: 3. American Nuclear Society (ANS) topical meeting on DOE spent nuclear fuel and fissile materials management, Charleston, SC (United States), 8-11 Sep 1998; Other Information: PBD: [1998]
- Country of Publication:
- United States
- Language:
- English
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