Don Mueller
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View article: Upper Subcritical Limit Calculations with Correlated Integral Experiments
Upper Subcritical Limit Calculations with Correlated Integral Experiments Open
The American National Standards Institute (ANSI) and American Nuclear Society (ANS) standard for Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations defines the Upper Subcritical Limit (USL) as “a limit on t…
View article: DETAILED DESIGN OF AN EPITHERMAL/INTERMEDIATE CRITICAL EXPERIMENT USING THE SANDIA NATIONAL LABORATORIES CRITICAL FACILITY
DETAILED DESIGN OF AN EPITHERMAL/INTERMEDIATE CRITICAL EXPERIMENT USING THE SANDIA NATIONAL LABORATORIES CRITICAL FACILITY Open
Nuclear criticality safety evaluations commonly take credit for neutron-absorbing materials. The normal practice for validating the calculational process to develop safety limits and margins of safety is to compare the safety case models w…
View article: Development of Criticality Safety Validation Guidance for NRC-Regulated Activities
Development of Criticality Safety Validation Guidance for NRC-Regulated Activities Open
Some guidance documents support validation of criticality safety evaluations for many system types, and some accommodate different regulators or regulatory requirements. While these reports address specific applications, no single document…
View article: Critical Experiment Design Phase 1 Report for Integral Request 441
Critical Experiment Design Phase 1 Report for Integral Request 441 Open
This report documents the analytical methods used in the first phase of critical experiment design (CED-1) conducted as part of integral experiment request (IER) 441. The purpose of IER-441 is to develop a capability to test the epithermal…
View article: Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models
Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models Open
In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments wer…
View article: Criticality safety enhancements for SCALE 6.2 and beyond
Criticality safety enhancements for SCALE 6.2 and beyond Open
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity…
View article: Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP K<sub>eff</sub> Calculations for PWR Burnup Credit Casks
Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP K<sub>eff</sub> Calculations for PWR Burnup Credit Casks Open
The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and…
View article: Upper Subcritical Calculations Based on Correlated Data
Upper Subcritical Calculations Based on Correlated Data Open
The American National Standards Institute and American Nuclear Society standard for Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations defines the upper subcritical limit (USL) as “a limit on the calculated…
View article: Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations
Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations Open
The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BU…
View article: Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2
Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2 Open
Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/…