Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

Approximately 54,000 tons of spent nuclear fuel are stored at operating nuclear power plants and several decommissioned power plants throughout the country. Spent fuel storage at these sites was never intended to be permanent. The current Federal plan is to place the fuel in a repository for permanent disposal in Nevada at Yucca Mountain.

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.

Disposal Criticality Analysis Methodology Topical Report describes a methodology for performing postclosure criticality analyses within the repository at Yucca Mountain, Nevada. An important component of the postclosure criticality analysis is the calculation of conservative isotopic concentrations for spent nuclear fuel. This report documents the isotopic calculation methodology. The isotopic calculation methodology is shown to be conservative based upon current data for pressurized water reactor and boiling water reactor spent nuclear fuel.

The objectives of this report are to document the development of data and assumptions used in
developing the transportation modules of the Total System Model (TSM) and to demonstrate that
transportation and transportation timing work correctly. This validation supports the
Transportation Design and Basis Report (BSC 2007a) and is issued with Version 6.0 of the TSM.
This report is largely based on the analysis approach and content in the transportation calculation
provided for TSM Version 4.0 (BSC 2006a). Much of the information in that report is repeated

The purpose of this calculation file is to document criticality calculations performed on two different rod consolidation waste package designs. The results presented in this calculation file may be used to support further evaluation of the rod consolidation waste package design.

The results of an extensive update of the decay data of the ORIGEN-S library are presented
in this report. The updated decay data were provided for both the ORIGEN-S and ORIGEN2
libraries in the same project. A complete edit of the decay data plus the available half-life
uncertainties are included in Appendix A. A detailed description of the types of data contained in the
library, the format of the library, and the data sources are also presented. Approximately 24% of the

The criticality consequence analysis for pressurized water reactor (PWR) waste packages (WP) (Civilian Radioactive Waste Management System [CRWMS] Management and Operating Contractor [M&O] 1997) focused on results obtained by maximizing postulated rates of reactivity insertion to assure no synergistic reactions could occur among waste packages from hypothetical criticality events. Other variables potentially influencing the criticality consequences were held constant during the above referenced analysis.

Stress corrosion cracking (SCC) is one of the most common corrosion-related causes for premature breach of metal structural components. SCC is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously (Jones 1992 [DIRS 169906], Section 8.1): metallurgical susceptibility, critical environment, and sustained tensile stresses.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

This study provides results supporting the conclusion that the repository can be operated over a varying range of thermal modes and therefore temperatures. In particular, this work focused on limiting the peak, postclosure waste package surface temperature to less than 85 degrees Celsius, a possible limit due to corrosion considerations. These operating modes were compared by varying the waste package in drift spacing (0.1-2.83 meters), drift pitch (drift spacing centerline to centerline of 40-120 meters), ventilation duration (75-300 years), and ventilation efficiency (50-80%).

At the request of the staff to the Blue Ribbon Commission on America’s Nuclear Future (“BRC”), we have reviewed whether certain recommendations in the BRC’s July 29, 2011 Draft Report respecting near-term actions by the Department of Energy (“DOE”) or other officers or agencies in the Executive Branch can be implemented under existing law. These recommendations relate to:
(1) Initial steps to site, license and construct consolidated interim storage facilities for spent nuclear fuel (“spent fuel”);

The objective of this calculation is to characterize the criticality aspect of a Department of Energy Spent Nuclear Fuel (DOE SNF) canister containing 5 Fast Flux Test Facility (FFTF) assemblies in a Five-Pack defense High-Level Waste (HLW) waste package. The purpose of this calculation is to investigate the criticality issues for the waste package (WP) containing HLW and DOE SNF canisters in various stages of degradation.

The management of spent nuclear fuel (SNF) and defense high level waste (HLW) is a complex sociotechnical
systems challenge. Coordinated, reliable, and safe performance will be required over very long
periods of time within evolving social and technical contexts. To accomplish these goals, a waste
management system will involve a host of facilities for interim storage and longterm disposal, a
transportation infrastructure, and research and development centers. The complexity of SNF and HLW

Spent Nuclear Fuel Litigation - Court of Federal Claims decision in Maine Yankee II, Conn Yankee II and Yankee Atomic II

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

The objective of this safety requirements publication is to set down the protection objectives and criteria for geological disposal and to establish the requirements that must be met to ensure the safety of this disposal option, consistent with the established principles of safety for radioactive waste management.

The purpose of this calculation is to estimate the probability of misloading a commercial spent
nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or
burnup) outside the waste package design. The waste package designs are based on the expected
commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1
and Table 1). For this calculation, a misloaded waste package is defined as a waste package that

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of the transportation, aging and disposal (TAD) canister-based systems.

About 20,000 metric tons of spent, or used, nuclear
fuel have accumulated since the beginning of commercial
nuclear power prbduction in the United States. At the end
of the currently licensed period of all existing nuclear power
plants and those under construction, the amount of spent
nuclear fuel is expected to total 87,000 metric tons.
Thus far, practically all of the spent nuclear fuel is
stored in water-filled pools at reactor sites. However, space
does not exist in the pools to store all the spent fuel expected

This report is developed from Technical Work Plan for: Thermodynamic Databases for Chemical Modeling (BSC 2006 [DIRS 177885]). The purpose of this analysis report is to update the thermochemical database data0.ymp.R4 (Output DTN: SN0410T0510404.002). Various data have been added, corrected, or corroborated, partly in response to four Condition Reports (CRs): CR 6489, CR 6731, CR 7542, and CR 7756. The most notable changes are a general revision of phosphate data to achieve consistency with the recommendations from the Committee on Data for Science and Technology (CODATA) (Cox. et al.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied·to future depletion calculations using SAS2H in which no measurements are available. ·

In February, 2011 the Blue Ribbon Commission (BRC) on America’s Nuclear Future requested the Department of Energy
(DOE) to provide a white paper summarizing the quantities and characteristics of potential waste generated by various
nuclear fuel cycles. The BRC request expressed interest in two classes of radioactive wastes:
 Existing waste that are or might be destined for a civilian deep geologic repository or equivalent.
 Potential future waste, generated by alternative nuclear fuel cycles (e.g. wastes from reprocessing, mixed-oxide

This report provides information on the inventory of commercial spent nuclear fuel, referred to herein as used nuclear fuel (UNF), as well as Government-owned UNF and high-level radioactive waste (HLW). Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining a different future commercial nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g. type, packaging, heat generation rate, decay curves).