All of the pieces for the Waste Control Specialists (WCS) Safety Analysis Report (SAR) and Environmental Report (ER).

By letter dated April 28, 2016, Waste Control Specialists, LLC (WCS) submitted a specific
license application under 10 CFR Part 72 requesting authorization to construct and operate a
Consolidated Interim Storage Facility for Spent Nuclear Fuel and Reactor-Related Greater Than
Class C Low-Level Waste in Andrews County, Texas. The Nuclear Regulatory Commission
(NRC) performed an acceptance review of the application to determine if the application
contains sufficient technical information to allow the NRC staff to complete the detailed technical

The U.S. Nuclear Regulatory Commission (NRC) staff has issued its final "Supplement to the
U.S. Department of Energy's Environmental Impact Statement for a Geologic Repository for the
Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain,
Nye County, Nevada," NUREG-2184. This document is available in the NRC's Agencywide
Documents Access and Management System (ADAMS) at Accession No. ML 16125A032. On
May 13, 2016, the NRC will announce in the Federal Register the availability of this document.

These slides were presented by Waste Control Specialists LLC (WCS) to the NRC at the June 16, 2015 pre-application public meeting at the NRC offices in Rockville, Maryland.

In the existing U.S. Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations governing the spent nuclear fuel and high-level radioactive waste site at Yucca Mountain, Nevada, the time period of compliance was set at 10,000 years. Recently, a Court ordered that EPA and NRC either revise the regulation on this topic to be "based upon and consistent with" recommendations made by a panel of the National Academy of Sciences, who recommended a time period of compliance out to as long as one million years, or seek congressional relief.

The purpose of this interim staff guidance (ISG) is to supplement standard review plan guidance
for evaluating the helium leakage testing and ASME Code1
required pressure
(hydrostatic/pneumatic) testing that is specified for the dry storage system (DSS) confinement
boundary. These acceptance tests are necessary to clearly demonstrate that the DSS
confinement boundary has been fabricated in accordance with the design criteria, and that its
operation complies with the intended safety bases of the confinement system and regulatory

Under the current guidance in ISG-1, Revision 1, “Damaged Fuel,” the definition of intact fuel
includes fuel rods containing no cladding defects greater than pinhole leaks or hairline cracks.
During the cask water removal process parts of, or all of, the fuel rods will be exposed to a
gaseous atmosphere. If the gaseous atmosphere is oxidizing, oxidation of fuel pellets or fuel
fragments can occur if a cladding breach exists (such as a pinhole). Oxidation may occur

Given the growing industry need to store spent reactor fuel of increasingly higher burnups and
heat loads in dry storage casks, and eventually to transport that same spent fuel in
transportation packages, analyzing the performance of casks and other radioactive material
packages has become a greater challenge. Finite Element, Finite Difference, and Finite
Volume analysis computer codes, defined here as Computational Modeling Software (CMS),
are tools used by many licensees to analyze cask and package performance in the structural

The standard review plans for storage of spent nuclear fuel and transportation of fissile
materials do not address, in detail, the technical considerations for crediting the neutron
absorber content of metal matrix composites used for preventing nuclear criticality. The Division
of Spent Fuel Storage and Transportation (SFST) considers the application of acceptance
criteria and methodology described in the recently developed American Standard for Testing
and Materials (ASTM) standard practice C1671-07, “Standard Practice for Qualification and

Authority for licensees to transport radioactive material comes from 10 CFR Part 71. Licensees
are authorized to transport Type B quantities and fissile materials in NRC-certified packages
under the general license in 71.17. Unlike 10 CFR Part 72, Part 71 does not include change
authority, that is, there is no specific Part 71 regulation that allows licensees to make changes in
the design or operation of an NRC-certified package without prior NRC approval. However,