This calculation demonstrates how uncertainties in half-life and branching fraction affect postclosure isotopic concentrations and the quantification of those effects in terms of reactivity.

This report was prepared in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 221, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate.

The attached documents are an extensive list of references relevant to burnup credit criticality analysis. Some of the references may be available within the CURIE document collection.

The purpose of this calculation is to document the Three Mile Island Unit 1 pressurized water reactor {PWR) reactivity calculations performed as part o f the commercial reactor critical (CRC) evaluation program. CRC evaluation reactivity calculations are performed at a number of statepoints, representing reactor start-up critical conditions at either beginning of life (BOL), beginning of cycle (BOC), or mid- cycle when the reactor resumed operation after a shutdown.

The purpose of this analysis report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The analysis addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository.

This report presents the analysis and conclusions with respect to disposal criticality for canisters containing aluminum-based fuels from research reactors. The analysis has been divided into three phases. Phase I, dealt with breached and flooded waste packages containing relatively intact canisters and intact internal (basket) structures; Phase II, the subject of this report, covers the degradation of the spent nuclear fuel (SNF) and structures internal to the codisposal waste package including high level waste (HLW), canisters, and criticality control material.

The purpose of this Analysis/Model Report (AMR) is to describe probabilistically the main features of the geometry of the fracture system in the vicinity of the repository. They will be used to determine the quantity of fissile material that could accumulate in the fractured rock underneath a waste package as it degrades. This AMR is to feed the geochemical calculations for external criticality reports. This AMR is done in accordance with the technical work plan (BSC (Bechtel SAIC Company) 2001 b). The scope of this AMR is restricted to the relevant parameters of the fracture system.

The purpose of this calculation is to demonstrate that the preclosure performance objectives specified in 10 CFR 63.111(a) and 10 CFR 63.111(b) (Reference 2.2.1) have been met for the proposed design and operations in the geologic repository operations area (GROA) during normal operations and Category 1 event sequences, and following Category 2 event sequences. Category 1 event sequences are those natural and human-induced event sequences that are expected to occur one or more times before permanent closure of the repository.

The purpose of this calculation is to perform degraded mode criticality evaluations of Plutonium disposed in a ceramic waste form and emplaced in a Monitored Geologic Repository (MGR). A 5 Defense High-Level Waste (DHLW) Canister Waste Package (WP) design, incorporating the can-in-canister concept for Plutonium immobilization is considered for this calculation. Each HLW glass pour canister contains 7 tubes. Each tube contains 4 cans, with 20 ceramic disks (immobilzed Plutonium) in each.