Evaluation of measured LWR spent fuel composition data for use in code validation

PDF Version Also Available for Download.

Description

Burnup credit (BUC) is a concept applied in the criticality safety analysis of spent nuclear fuel in which credit or partial credit is taken for the reduced reactivity worth of the fuel due to both fissile depletion and the buildup of actinides and fission products that act as net neutron absorbers. Typically, a two-step process is applied in BUC analysis: first, depletion calculations are performed to estimate the isotopic content of spent fuel based on its burnup history; second, three-dimensional (3-D) criticality calculations are performed based on specific spent fuel packaging configurations. In seeking licensing approval of any BUC approach ... continued below

Physical Description

30 p.

Creation Information

Hermann, O.W.; DeHart, M.D. & Murphy, B.D. February 1, 1998.

Context

This report is part of the collection entitled: Office of Scientific & Technical Information Technical Reports and was provided by UNT Libraries Government Documents Department to Digital Library, a digital repository hosted by the UNT Libraries. It has been viewed 13 times . More information about this report can be viewed below.

Who

People and organizations associated with either the creation of this report or its content.

Sponsor

Publisher

Provided By

UNT Libraries Government Documents Department

Serving as both a federal and a state depository library, the UNT Libraries Government Documents Department maintains millions of items in a variety of formats. The department is a member of the FDLP Content Partnerships Program and an Affiliated Archive of the National Archives.

Contact Us

What

Descriptive information to help identify this report. Follow the links below to find similar items on the Digital Library.

Description

Burnup credit (BUC) is a concept applied in the criticality safety analysis of spent nuclear fuel in which credit or partial credit is taken for the reduced reactivity worth of the fuel due to both fissile depletion and the buildup of actinides and fission products that act as net neutron absorbers. Typically, a two-step process is applied in BUC analysis: first, depletion calculations are performed to estimate the isotopic content of spent fuel based on its burnup history; second, three-dimensional (3-D) criticality calculations are performed based on specific spent fuel packaging configurations. In seeking licensing approval of any BUC approach (e.g., disposal, transportation, or storage) both of these two computational procedures must be validated. This report was prepared in support of the validation process for depletion methods applied in the analysis of spent fuel from commercial light-water-reactor (LWR) designs. Such validation requires the comparison of computed isotopic compositions with those measured via radiochemical assay to assess the ability of a computer code to predict the contents of spent fuel samples. The purpose of this report is to address the availability and appropriateness of measured data for use in the validation of isotopic depletion methods. Although validation efforts to date at ORNL have been based on calculations using the SAS2H depletion sequence of the SCALE code system, this report has been prepared as an overview of potential sources of validation data independent of the code system used. However, data that are identified as in use in this report refer to earlier validation work performed using SAS2H in support of BUC. This report is the result of a study of available assay data, using the experience gained in spent fuel isotopic validation and with a consideration of the validation issues described earlier. This report recommends the suitability of each set of data for validation work similar in scope to the earlier work.

Physical Description

30 p.

Notes

INIS; OSTI as DE98003778

Source

  • Other Information: PBD: Feb 1998

Language

Item Type

Identifier

Unique identifying numbers for this report in the Digital Library or other systems.

  • Other: DE98003778
  • Report No.: ORNL/M--6121
  • Grant Number: AC05-96OR22464
  • DOI: 10.2172/629472 | External Link
  • Office of Scientific & Technical Information Report Number: 629472
  • Archival Resource Key: ark:/67531/metadc691876

Collections

This report is part of the following collection of related materials.

Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

What responsibilities do I have when using this report?

When

Dates and time periods associated with this report.

Creation Date

  • February 1, 1998

Added to The UNT Digital Library

  • Aug. 14, 2015, 8:43 a.m.

Description Last Updated

  • Jan. 21, 2016, 8:21 p.m.

Usage Statistics

When was this report last used?

Yesterday: 0
Past 30 days: 2
Total Uses: 13

Interact With This Report

Here are some suggestions for what to do next.

Start Reading

PDF Version Also Available for Download.

International Image Interoperability Framework

IIF Logo

We support the IIIF Presentation API

Hermann, O.W.; DeHart, M.D. & Murphy, B.D. Evaluation of measured LWR spent fuel composition data for use in code validation, report, February 1, 1998; Tennessee. (digital.library.unt.edu/ark:/67531/metadc691876/: accessed October 17, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.