Results for case A for task order 041-001

A design change in the Multicanister Overpack (MCO) container scrap basket has required a reevaluation for criticality safety. The bounding case, the drop accident case for MarkIV-loaded MCOs, was reanalyzed with less mass in the baseplate of the upper scrap basket. The reduction in mass was from 50 lb to 27 lb. No statistically significant difference between the two cases was found. The original analysis is therefore considered valid and no additional analysis is required

Use of WIMS-E lattice code for prediction of the transuranic source term for spent fuel dose estimation

A recent source term analysis has shown a discrepancy between ORIGEN2 transuranic isotopic production estimates and those produced with the WIMS-E lattice physics code. Excellent agreement between relevant experimental measurements and WIMS-E was shown, thus exposing an error in the cross section library used by ORIGEN2

Criticality safety evaluation report for spent nuclear fuelprocessing and storage facilities

This criticality evaluation is for Spent N Reactor fuel unloaded from the existing canisters in both KE and KW Basins, and loaded into multiple canister overpack (MCO) containers with specially- built baskets containing either 54 Mark IV or 48 Mark IA fuel assemblies. The criticality evaluations include loading baskets into the MCO/Cask, operations at the Cold Vacuum Drying Facility (CVDF), and storage in the Canister Storage Building (CSB). Many conservatisms have been built into this analysis, the primary one being the selection of the k{sub eff} @ 0.95 criticality safety limit

Calculation of the Fast Flux Test Facility fuel pin tests with the WIMS-E and MCNP codes

The Fuel Assembly Area (FAA) at the Fast Flux Test Facility site on the Hanford Site at Richland, Washington currently is being prepared to fabricate mixed oxide fuel (U, Pu) for the FFTF. Calculational tools are required to perform criticality safety analyses for various process locations and to establish safe limits for fissile material handling at the FAA. These codes require validation against experimental data appropriate for the compositions that will be handled. Critical array experiments performed by Bierman provide such data for mixed oxide fuel in the range Pu/(U+Pu) = 22 wt %, and with Pu-240 contents equal to 12 wt %. Both the Monte Carlo Neutron Photon (MCNP) and the Winfrith Improved Multigroup Scheme (WIMS-E) computer codes were used to calculate the neutron multiplication factor for explicit models of the various critical arrays. The W-CACTUS modules within the WIMS-E code system was used to calculate k{infinity} for the explicit array configuration, as well as few-group cross sections that were then used in a three-dimensional diffusion theory code for the calculation of k{sub eff} for the finite array. 10 refs., 15 figs., 7 tabs

Results for additional calculations for Task Order 98-009B-01,addendum 3 to: HNF-SD-SNF-CSER-005, Revision 3

Several sets of new calculations were performed to support the Spent Nuclear Fuel project nuclear criticality safety evaluation. These new calculations include partial loading of Mark IA inner elements after the outers have been loaded, a new, more robust design for the central pipe insert for the Mark IA fuel baskets, Single Pass Reactor fuel loading, the lowering of a Mark IV-loaded MCO through the concrete operating deck as-it is inserted into the Canister Storage Building storage array, and one additional scrap basket loading error. None of these calculations exceeded the criticality safety limit

ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

Verification and validation plan for reactor analysis computer codes

This report presents a verification and validation (V V) plan for reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. This plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River (DOE-SR) as identified in a letter to R.E. Tiller (Reference 1). The plan stresses verification and validation by demonstrating successful application of the codes to predict reactor data, special measurements, and benchmarks. This is in compliance with the intent of the WSRC quality assurance requirements. Restructuring of software especially to achieve verification compliance is not recommended