Most analysis and simulation codes rely on evaluated neutron cross-section data from libraries such as ENDF/B-VII.1, JEFF3.1, or JENDL-4.0. High-quality nuclear data with covariances are required for the design and analysis of nuclear systems. However, data deficiencies and missing covariances in some existing cross-section evaluations make new neutron induced cross-section measurements necessary.
Essential for a large range of nuclear reactor applications, evaluated nuclear data, such as cross sections and fission products, are generated on the basis of experimental measurements using nuclear reaction models depending on adjustable parameters
Evaluations of nuclear cross sections data and their uncertainties using physics-based formalism is needed to decrease cross section data uncertainties for advanced simulations of nuclear application.
Provide technical support to the NRC in support of their review of nuclear criticality safety (NCS) licensing submittals and other issues associated with storage and handling of nuclear fuel at power plants and in storage and transportation systems.
Vendors are beginning to request credit for burnup in boiling water reactor (BWR) spent nuclear fuel (SNF). The NRC staff requires a technical basis for reviewing and acting on these applications.
The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation, is offering Safety Analysis Report for Packaging (SARP) shielding and nuclear criticality safety (NCS) courses for SARP generalists and analysts.