Baselining a Spent Nuclear Fuel Cask Shielding Model

Cumberland, Riley M.; Banerjee, Kaushik

Title: Baselining a Spent Nuclear Fuel Cask Shielding Model

Abstract

Calculating more accurate dose rates to provide additional flexibility to both nuclear facility designers and operational planners.

Authors:

^[1];

^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Publication Date:: Sun Jan 28 00:00:00 EST 2018

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE

OSTI Identifier:: 1436949

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Transactions of the American Nuclear Society

Additional Journal Information:: Journal Volume: 117; Journal Issue: 1; Journal ID: ISSN 0003-018X

Publisher:: American Nuclear Society

Country of Publication:: United States

Language:: English

Subject:: 12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES

Citation Formats


                    Cumberland, Riley M., and Banerjee, Kaushik. Baselining a Spent Nuclear Fuel Cask Shielding Model.  United States: N. p., 2018. 
Web.

Copy to clipboard


                    Cumberland, Riley M., & Banerjee, Kaushik. Baselining a Spent Nuclear Fuel Cask Shielding Model.  United States.

Copy to clipboard


                    Cumberland, Riley M., and Banerjee, Kaushik. Sun .  
"Baselining a Spent Nuclear Fuel Cask Shielding Model".  United States.  https://www.osti.gov/servlets/purl/1436949.

Copy to clipboard


                    
@article{osti_1436949,

  title        = {Baselining a Spent Nuclear Fuel Cask Shielding Model},

  author       = {Cumberland, Riley M. and Banerjee, Kaushik},

  abstractNote = {Calculating more accurate dose rates to provide additional flexibility to both nuclear facility designers and operational planners.},

  doi          = {},

  journal      = {Transactions of the American Nuclear Society},

  number       = 1,

  volume       = 117,

  place        = {United States},

  year         = {Sun Jan 28 00:00:00 EST 2018},

  month        = {Sun Jan 28 00:00:00 EST 2018}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (DOE)

Publisher's Version of Record

The DOI is not currently available

Other availability

Search WorldCat to find libraries that may hold this journal

Figures / Tables:

Fig. 1: NAC-LWT radial slice.

All figures and tables (5 total)

Save / Share:

Export Metadata

Save to My Library

Figures / Tables found in this record:

Figures/Tables have been extracted from DOE-funded journal article accepted manuscripts.

Similar Records in DOE PAGES and OSTI.GOV collections:

Shielding analysis considerations for spent-fuel transport and storage casks

Conference Lesser, T H ; Wells, A H - Transactions of the American Nuclear Society; (United States)

Spent-fuel transport and storage casks are expected to be major components of the solution to the problem of spent-fuel disposal in the United States. In keeping with as low as reasonably achievable dose levels, these casks are designed to incorporate shielding for neutrons and gamma rays to minimize the radiation exposure associated with spent-fuel cask storage, transport, and maintenance. Nuclear Assurance Corporation (NAC) has obtained good results from the QAD-CG computer code for essentially monoenergetic {sup 60}Co gamma-ray calculations. The QAD-CG code is also used in the evaluation of the dose rate contribution from the active fuel region. The energymore »« less
Evaluation of shielding analysis methods in spent-fuel cask environments

Journal Article Broadhead, B L ; Tang, J S ; Childs, R L ; ... - Nuclear Technology

The three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, is applied to the analysis of a series of simple geometry benchmark experiments and prototypic spent-fuel storage cask measurements. The simple geometry experiments were performed in Japan and at the General Electric-Morris Operation facility; the cask measurements were performed at the Idaho National Engineering Laboratory. A total of five storage cask problems and two simple geometry problems were analyzed to determine the expected accuracies of computational analyses using well-established source-generation and Monte Carlo codes. The general trends seen in this work are in agreement withinmore »« less
The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Technical Report Cumberland, Riley ; Radulescu, Georgeta ; Banerjee, Kaushik

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNFmore »« less
https://doi.org/10.2172/1649326

Full Text Available
Calculation of radiation dose rates from a spent nuclear fuel shipping cask

Conference Chen, S Y ; Yuan, Y C

Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask andmore »« less
Full Text Available
Calculation of radiation dose rates from a spent nuclear fuel shipping cask

Conference Chen, S Y ; Yuan, Y C - Transactions of the American Nuclear Society; (USA)

Radiation doses from a spent nuclear fuel cask are usually incurred from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the caskmore »« less

Similar Records