Review of Subcritical Source-Driven Noise Analysis Measurements

Subcritical source-driven noise measurements are simultaneous Rossi-a and randomly pulsed neutron measurements that provide measured quantities that can be related to the subcritical neutron multiplication factor. In fact, subcritical source-driven noise measurements should be performed iii lieu of Rossi-a rneasurements because of the additional information that is obtained from noise measurements such as the spectral ratio and the coherencc functions. The basic understanding of source-driven noisc analysis

slides - ISFSI Pad Design Issues

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

Aging and Phase Stability of Waste Package Outer Barrier

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.

SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors.

Geochemistry Model Validation Report: External Accumulation Model

The purpose of this report is to document and validate the external accumulation model that predicts accumulation of fissile materials in the invert, fractures and lithophysae in the rock beneath a degrading waste package containing spent nuclear fuel (SNF) in the monitored geologic repository at Yucca Mountain. (Lithophysae are hollow, bubblelike structures in the rock composed of concentric shells of finely crystalline alkali feldspar, quartz, and other materials (Bates and Jackson 1984 [DIRS 128109], p.

SAS2H Analysis of Radiochemical Assay Samples from Trino Vercelles PWR Reactor

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.

Letter - Request Approval to Establish and Populate the Ad Hoc Subcommittee on Co‐mingling of Defense and Commercial Wastes

Dear Tim: As we work to complete our final recommendations to the Secretary by January 29, 2012, we have determined that our efforts would be aided by the formation of an ad hoc subcommittee to investigate the issue of co‐mingling of defense and commercial wastes. Specifically, the ad hoc subcommittee would review and make a recommendation to the Commission on the issue of whether the 1985 Presidential decision to co‐mingle defense and commercial wastes for disposal should be revisited in light of changes that have

EBS Radionuclide Transport Abstraction

The purpose of this report is to develop and analyze the Engineered Barrier System (EBS) Radionuclide Transport Abstraction Model, consistent with Level I and Level II model validation, as identified in Technical Work Plan for: Near-Field Environment: Engineered Barrier System: Radionuclide Transport Abstraction Model Report (BSC 2006 [DIRS 177739]). The EBS Radionuclide Transport Abstraction (or RTA) is the conceptual model used in the Total System Performance Assessment (TSPA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ).

Code to Code Comparison of One- and Two-Dimensional Methods

This calculation file provides comparisons of one- and two-dimensional methods for calculating the isotopic content of spent nuclear fuel. The one-dimensional methods use the SAS2H sequence of SCALE 4.4a (Reference 7.1) and the SAS2 sequence of SCALE 5.0 (Reference 7.2). The two-dimensional method uses the TRITON control module along with the T-DEPL sequence of SCALE 5.0 (Reference 7.3). The SAS2H results for SCALE 4.4a are taken from Reference 7.4. Data from previous two-dimensional calculations (Reference 7.5) using CASM03 will also be used for comparisons with TRITON.

From Three Mile Island to the Future Improving Worker Safety and Health In the U.S. Nuclear Power Industry

The Blue Ribbon Commission on America’s Nuclear Future (BRC) asked us to study whether occupational safety and health conditions in today's U.S. nuclear industry are reasonably safe, and if those conditions have improved since the Three Mile Island event in 1979. The BRC also asked us to look to the future, to try to anticipate worker safety and health risks that should be addressed by the industry, its government regulators and private watchdogs. Over the eight weeks allotted, we performed a limited review of the literature and spoke with

Letter to Mr. Tim Frazier, Designated Federal Official - DOE Inputs Needed for High Level Waste Program Cost and Financing Overview

Dear Mr. Frazier: At our request, the Commission staff is in the process of assembling information on the costs and financing of the US program to manage used fuel and high-level nuclear wastes. To assist in the completion of this effort, it would be most helpful if the Department could provide the information listed int the attachment. ...

Assessment of Benefits for Extended Burnup Credit in Transporting PWR Spent Nuclear Fuel in the USA

This paper presents an assessment of the benefits for extended burnup credit in transporting pressurized-water-reactor (PWR) spent nuclear fuel (SNF) in the United States. A prototypic 32- assembly cask and the current regulatory guidance were used as bases for this assessment. By comparing recently released PWR discharge data with actinide-only-based loading curves, this evaluation shows that additional negative reactivity (through either increased credit for fuel burnup or

Fast Flux Test Facility (FFTF) Reactor Fuel Criticality Calculations

The purpose of these calculations is to characterize the criticality safety concerns for the storage of Fast Flux Test Facility (FFTF) nuclear fuel in a Department of Energy spent nuclear fuel (DOE SNF) canister in a co-disposal waste package. These results will be used to support the analysis that will be done to demonstrate concept viability related to use in the Monitored Geologic Repository (MGR) environment.

Vacuum Drying Test Plan - Public Version

NRC initiated a research activity with the Center for Nuclear Waste Regulatory Analyses (CNWRA®) to develop a conceptual test plan for measuring the quantity of residual water remaining in a canister following vacuum drying to the criterion referenced in NUREG–1536 (NRC, 2010). The test plan will be used to help NRC assess options for independently performing an experimental program or to support engagements with industry or the Department of Energy should they undertake a similar effort.

Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted

Analysis of the Total System Life Cycle Cost of the Civilian Radioactive Waste Management Program

The Analysis of the Total System Life Cycle Cost (TSLCC) of the Civilian Radioactive Waste Management Program represents the Office of Civilian Radioactive Waste Management's most recent estimate of the costs to dispose of the Nations's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). This TSLCC analysis projects all Program costs through 2119 for a surrogate, single potential repository. The design and emplacement concepts in this TSLCC analysis are the same as those presented in the Monitored Geologic Repository Project Description Document.

slides - From a Chairman's Perspective

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

Consolidated Interim Storage of Commercial Spent Nuclear Fuel

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.

CRC Reactivity Calculations for Three Mile Island Unit 1

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.

Isotopic Model for Commercial SNF Burnup Credit

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.

Total System Model Version 6.0 Transportation Validation Report

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

Rod Consolidation Waste Package Criticality Calculations

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.

ORIGEN-S Decay Data Library and Half-Life Uncertainties

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

Sensitivity Study of Reactivity Consequences to Waste Package Egress Area

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 of Waste Package Outer Barrier and Drip Shield Materials

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.