slides - NRC Management Perspectives
slides - NRC Management Perspectives
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
The objective of these calculations is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) Three Mile Island- Unit 2 (TMI-2) spent nuclear fuel (SNF) in canisters. This analysis evaluates codisposal in a 5-Defense High-Level Waste (5-DHLW/DOE SNF) Long Waste Package (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M&O] 2000b, Attachment V), which is to be placed in a potential monitored geologic repository (MGR).
The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site.
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having some or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment.
The purpose of this calculation is to perform partially and fully degraded mode criticality evaluations of plutonium disposed of in a ceramic waste form and emplaced in a Monitored Geologic Repository. The partially degraded mode is represented by the immobilized plutonium ceramic discs piled in the bottom of the waste package (WP) while neutron absorbers begin to leach out of the discs.
The purpose of this analysis is to identify, extract, and reformat weather (meteorological) data that is appropriate for use as input to an infiltration model, within the Yucca Mountain region. The analysis uses relevant meteorological data (e.g., precipitation and temperature) from source stations, and reformats or converts the data into a form suitable for the generation of meteorological conditions for a 10,000-year future climate in the Yucca Mountain region.
The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall.
The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared ·according to Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]).
The Global Nuclear Energy Partnership (GNEP) Program, a United States (U.S.) Department of
Energy (DOE) program, is intended to support a safe, secure, and sustainable expansion of
nuclear energy, both domestically and internationally. Domestically, the GNEP Program would
promote technologies that support economic, sustained
production of nuclear-generated electricity, while
reducing the impacts associated with spent nuclear fuel
disposal and reducing proliferation risks. DOE envisions
changing the U.S. nuclear energy fuel cycle1 from an
This calculation uses regression (CLReg V1.0 computer code) and non-parametric statistical methods, as specified in References 1 and 12, to develop the critical limit for the 21 Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) waste package (WP) in the proposed geologic repository at Yucca Mountain, Nevada. The critical limit is a limiting value of the effective neutron multiplication factor at which a WP configuration is considered potentially critical.
The purpose of this calculation is to perform a parametric study to determine the effects of fission product leaching, assembly collapse, and iron oxide loss on the reactivity of a waste package (WP) containing mixed oxide (MOX) spent nuclear fuel (SNF). Previous calculations (CRWMS M&O 1998a) have shown that the criticality control features of the WP are adequate to prevent criticality of a flooded WP for all the enrichment/burnup pairs expected for the MOX SNF.
The purpose of this calculation is to apply the process described in the Preclosure Criticality Analysis Process Report (Ref. 2.2.12) to establish the bias for keff calculations performed for commercial nuclear fuels using the MCNP code system. This bias will be used in criticality safety analyses as part of the basis for establishing the upper subcritical limit (USL). This calculation also defines the range of applicability (ROA) for which the bias may be used directly without need to consider additional penalties on the USL.
The objective of this calculation is to document the Grand Gulf Unit 1 (GGl) reactivity calculations for sixteen critical statepoints in· cycles 4 through 8. The GG1 reactor is a boiling water reactor (BWR) owned and operated by Entergy Operations Inc. The Commercial Reactor Criticality (CRC) evaluations support the development and validation of the neutronic models used for criticality analyses involving commercial spent nuclear fuel to be placed in a geologic repository. This calculation is performed as part of the evaluation in the CRC program.
The waste forms under consideration for disposal in the repository at Yucca Mountain contain scores of radionuclides. It would be impractical and highly inefficient to model all of these radionuclides in a total system performance assessment (TSPA). Thus, the purpose of this radionuclide screening analysis is to remove from further consideration (screen out) radionuclides that are unlikely to significantly contribute to radiation dose to the public from a nuclear waste repository at Yucca Mountain.
The Department of Energy’s Office of Fuel Cycle Technologies (FCT) in the Office of Nuclear Energy (DOE-NE) has conducted a technical review and assessment of the total current inventory [~70,150 MTHM (metric ton of heavy metal) as of 2011] of domestic discharged used nuclear fuel (UNF) and estimated that up to ~1700 MTHM of existing commercial UNF should be considered for retention to support research, development, and demonstration (RD&D) needs and national security interests.
Spent fuel transportation and storage cask designs based on a burnup credit approach must
consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For
example, the spent fuel composition must be adequately characterized and the criticality analysis
model can be complicated by the need to consider axial burnup variations. Parametric analyses are
needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel
DOE decided to terminate the Yucca Mountain repository program because, according to DOE officials, it is not a workable option and there are better solutions that can achieve a broader national consensus. DOE did not cite technical or safety issues. DOE also did not identify alternatives, but it did create a Blue Ribbon Commission to evaluate and recommend alternatives.
This evaluation investigates the potential benefits of separating the transuranic elements from spent reactor fuel before it is disposed of in geologic repositories. It addresses the question: Would the benefits to radioactive waste disposal justify both processing the spent fuel and deploying liquid metal reactors (LMRs) to transmute the separated transuranics?
http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=NP-7…
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
The purpose of this calculation is to apply the process described in the TDR-DS0-NU-000001 Rev. 02, Preclosure Criticality Analysis Process Report (Ref. 2.2.25) to aid in establishing design and operational criteria important to criticality safety and to identify potential control parameters and their limits important to the criticality safety of commercial spent nuclear fuel (CSNF) handling operations in the Wet Handling Facility (WHF)
This calculation file uses the MCNP neutron transport code to determine the range of parameters for Pressurized Water Reactor Spent Nuclear Fuel contained with a 21 PWR waste package (WP). Four base geometry patterns were considered in this work and included the following: intact fuel assemblies with intact WP internal components, intact fuel assemblies with degraded WP internal components, degraded fuel assemblies with intact WP internal components, and degraded fuel assemblies with degraded WP internal components.
The American Nuclear Society (ANS) supports the safe, controlled, licensed, and regulated interim
storage of used nuclear fuel (UNF) (irradiated, spent fuel from a nuclear power reactor) until disposition
can be determined and completed. ANS supports the U.S. Nuclear Regulatory Commission’s (NRC’s)
determination that “spent fuel generated in any reactor can be stored safely and without significant
environmental impacts for at least 30 years beyond the licensed life for operation.
This analysis provides information necessary for total system performance assessment (TSPA) for the license application (LA) to include the excess U.S. Department of Energy (DOE) plutonium in the form of mixed oxide (MOX) spent nuclear fuel and lanthanide borosilicate (LaBS) glass. This information includes the additional radionuclide inventory due to MOX spent nuclear fuel and LaBS glass and the analysis that shows that the TSPA models for commercial spent nuclear fuel (CSNF) and high-level waste (HLW) degradation are appropriate for MOX spent nuclear fuel and LaBS glass, respectively.