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K-Infinite Trends with Burnup, Enrichment, and Cooling Time for BWR Fuel Assemblies
K-Infinite Trends with Burnup, Enrichment, and Cooling Time for BWR Fuel Assemblies
This report documents the work performed by ORNL for the Yucca Mountain Project (YMP)
M&O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k values for inf
infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various
burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting
criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon
Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation
Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation
The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality
safety limits for applications outside reactors be validated by comparison with appropriate critical
experiments. This report provides a detailed description of 34 fresh fuel critical experiments and
their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library.
The 34 critical experiments were selected based on geometry, material, and neutron interaction
SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5
SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5
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 is to be taken for
the reduced reactivity of burned or spent fuel relative to its original "fresh" composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to
Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages
Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages
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
Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel
Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel
This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance.
Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses
Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses
This report presents studies performed to support the development of a technically justifiable approach for
addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality
safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for
addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is
examined in detail to identify profiles that maximize the neutron multiplication factor, keff, assess its
Preliminary Site Requirements and Considerations for a Monitored Retrievable Storage Facility
Preliminary Site Requirements and Considerations for a Monitored Retrievable Storage Facility
In the November 1989 Report to Congress on Reassessment of the Civilian
Radioactive Waste Management Program (DOE/RW-0247), the Secretary of Energy
announced an initiative for developing a monitored retrievable storage (MRS) facility
that is to start spent-fuel acceptance in 1998. This facility, which will be licensed by
the U.S. Nuclear Regulatory Commission (NRC), will receive spent fuel from
commercial nuclear power plants and provide a limited amount of storage for this
Preliminary Feasibility Assessment for Several Specific MRS Design Alternatives with the Potential for Early Deployment Revision1
Preliminary Feasibility Assessment for Several Specific MRS Design Alternatives with the Potential for Early Deployment Revision1
This vintage 1990 document presents the results of WESTON'S preliminary assessment of the feasibility of several alternative fuel-transfer and storage concepts that have the potential for early spent-fuel acceptance at an MRS facility. The feasibility study was part of a series of studies conducted by the U.S. Department of Energy (DOE) during the late 1980's and early 1990's in an effort to establish an MRS design configuration.
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2
Unirradiated reactor fuel has a well-specified nuclide composition that provides a
straightforward and bounding approach to the criticality safety analysis of transport and storage
casks. As the fuel is irradiated in the reactor, the nuclide composition changes and, ignoring
the presence of burnable poisons, this composition change will cause the reactivity of the fuel to
decrease. Allowance in the criticality safety analysis for the decrease in fuel reactivity resulting
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks
Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport
and Storage Casks
Review of Subcritical Source-Driven Noise Analysis Measurements
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
SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3
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.
Assessment of Benefits for Extended Burnup Credit in Transporting PWR Spent Nuclear Fuel in the USA
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
Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel
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
A Coordinated U.S. Program to Address Full Burnup Credit in Transport and Storage Casks
A Coordinated U.S. Program to Address Full Burnup Credit in Transport and Storage Casks
The benefits of burnup credit and the technical issues associated with utilizing burnup credit in spent
nuclear fuel (SNF) casks have been studied in the United States for almost two decades. The issuance of the
U.S. Nuclear Regulatory Commission (NRC) staff guidance for actinide-only burnup credit in 2002 was a
significant step toward providing a regulatory framework for using burnup credit in transport casks. However,
adherence to the current regulatory guidance (e.g., limit credit to actinides) enables only about 30% of the existing
Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses
Validation of the SCALE System for PWR Spent Fuel Isotopic Composition Analyses
The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic
composition by the SCALE system depletion analysis was assessed using data presented in the report.
Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were
compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2,
and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of
predicted and measured concentrations for 14 actinides and 37 fission and activation products.
MRS Siting Briefing
MRS Siting Briefing
This briefing paper is a component of the comprehensive briefing package developed for the Negotiator, and describes previous DOE experience in its attempt to site an MRS facility. The Background section highlights, in chronological order, significant events in DOE's MRS siting history from enactment of the Nuclear Waste Policy Act of 1982 to the issuance of the "Report to Congress on Reassessment of the Civilian Radioactive Waste Management Program" in November 1989.
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
This report attempts to summarize and consolidate the existing knowledge on axial
burnup distribution issues that are important to burnup credit criticality safety calculations.
Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup
credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity
difference between the neutron multiplication factor (keff) calculated with explicit representation
Determination of the Accuracy of Utility Spent Fuel Burnup Records (Interim Report)
Determination of the Accuracy of Utility Spent Fuel Burnup Records (Interim Report)
This report summarizes the results of an initial investigation into the uncertainties associated with the burnup records maintained by nuclear power plants. The results indicate that there is an overall uncertainty of about 2 percent in the burnup records, which must be accounted for in spent fuel applications.
Strategies for Application of Isotopic Uncertainties in Burnup Credit
Strategies for Application of Isotopic Uncertainties in Burnup Credit
Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest
sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to
propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted
neutron multiplication factor (keff) of the system can have a significant effect on the uncertainty in the
safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport
OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results
OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results
Burnup credit is an ongoing technical concern for many countries that operate commercial
nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a
Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency
of the Organization for Economic Cooperation and Development. This working group has
established a set of well-defined calculational benchmarks designed to study significant aspects of
burnup credit computational methods. These benchmarks are intended to provide a means for the
Commercial Spent Nuclear Fuel Waste Package Misload Analysis
Commercial Spent Nuclear Fuel Waste Package Misload Analysis
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
Wyoming Decline of MRS
Wyoming Decline of MRS
This letter is written by Former Wyoming Governor Mike Sullivan to inform the Fremont County Commissioners of his conclusion to decline the MRC facility.