Background To determine whether current and former construction workers are at
significant risk for occupational illnesses from work at the Department of Energy’s (DOE)
nuclear weapons facilities, screening programs were undertaken at the Hanford Nuclear
Reservation, Oak Ridge Reservation, and the Savannah River Site.

Spent Fuel Project Office Interim Staff Guidance - 8

Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 1

The Nuclear Waste Policy Act of 1982 (the Act), established a
step-by-step process for the siting of the nation's first repository for
high-level radioactive waste and spent fuel. The Act gave the Department of
Energy (DOE) the primary responsibility for conducting this siting process.
The first step in the process laid out in the Act was the development by
the DOE, with the concurrence of the Nuclear Regulatory Commission (NRC), of
general guidelines to be used by the Secretary of the DOE (the Secretary) in

Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport
and Storage Casks

Background Medical screening programs were begun in 1996 and 1997 at three Department
of Energy (DOE) nuclear weapons facilities (Hanford Nuclear Reservation, Oak
Ridge, and the Savannah River Site) to evaluate whether current and former construction
workers are at significant risk for occupational illnesses. The focus of this report is
pneumoconiosis associated with exposures to asbestos and silica among workers enrolled
in the screening programs through September 30, 2001.

Background Medical screening programs at three Departments of Energy (DOE)
nuclear weapons facilities (Hanford Nuclear Reservation, Oak Ridge, and the Savannah
River Site) have included audiometric testing since approximately 1996. This report
summarizes hearing evaluations through March 31, 2003.
Methods Occupational examinations included a medical history, limited physical
examination, and tests for medical effects from specific hazards, including audiometric
testing. Hearing thresholds by frequency for DOE workers were compared to agestandardized

Background The U.S. Department of Energy (DOE) established medical screening
programs at the Hanford Nuclear Reservation, Oak Ridge Reservation, the Savannah
River Site, and the Amchitka site starting in 1996.Workers participating in these programs
have been followed to determine their vital status and mortality experience through
December 31, 2004.
Methods A cohort of 8,976 former construction workers from Hanford, Savannah River,
Oak Ridge, and Amchitka was followed using the National Death Index through December

This Interim Staff Guidance (ISG) provides guidance to the staff for determining if
storage systems to be licensed under 10 CFR Part 72 allow ready retrieval of spent fuel.
This guidance is not a regulation or a requirement.

This Interim Staff Guidance (ISG) provides guidance to the staff on classifying spent nuclear
fuel as either (1) damaged, (2) undamaged, or (3) intact, before interim storage or
transportation. This is not a regulation or requirement and can be modified or superseded by
an applicant with supportable technical arguments.

Revision 2

The closure weld for the outer cover plate for austenitic stainless steel designs may be
inspected using either volumetric or multiple pass dye penetrant techniques subject to the
following conditions:
• Dye penetrant (PT) examination may only be used in lieu of volumetric
examination only on austenitic stainless steels. PT examination should be done
in accordance with ASME Section V, Article 6, “Liquid Penetrant Examination.”
• For either ultrasonic examination (UT) or PT examination, the minimum

Several changes have occurred since the issuance of NUREG-1536, “Standard Review Plan
(SRP) for Dry Cask Storage Systems,” that affect the staff’s approach to confinement
evaluation. The attachment to this ISG integrates the current staff approach into a revision of
ISG-5. The highlights of the changes include:
• Reflects October 1998 revisions to 10 CFR 72.104 and 10 CFR 72.106.
• Expands and clarifies acceptance criteria associated with confinement analysis and
acceptance of “leak tight” testing instead of detailed confinement analysis.

The Standard Review Plan, NUREG-1536, Chapter 5, Section V, 2 recommends that “the
applicant calculate the source term on the basis of the fuel that will actually provide the
bounding source term,” and states that the applicant should, “either specify the minimum initial
enrichment or establish the specific source terms as operating controls and limits for cask use.”
A specified source term is difficult for most cask users to determine and for inspectors to verify.

Staff raised two major issues concerning the adverse effects of fission gases to the gas-mixture
thermal conductivity in a spent fuel canister in a post accident environment. The two major
concerns were: (1) the reduction of the thermal conductivity of the canister gas by the mixing of
fission gases expelled from failed fuel pins and (2) the resultant temperature and pressure rise
within the canister. Since the fission gas is typically of a lower conductivity than the cover gas,

Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of
Radioactive Material, and 10 CFR Part 72, Licensing Requirements for the Independent
Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater
Than Class C Waste, require that spent nuclear fuel (SNF) remain subcritical in transportation
and storage, respectively. Unirradiated reactor fuel has a well-specified nuclide composition
that provides a straightforward and bounding approach to the criticality safety analysis of

The purpose of this ISG is to clarify the technical criteria for types of materials that will be |
considered associated with the storage of spent fuel assemblies. While control rods are |
mentioned in the Standard Review Plan as possible contents, specific information and guidance
is lacking.

Revision 1

There is no existing American Society of Mechanical Engineers (ASME) Code for the design
and fabrication of spent fuel dry storage casks. Therefore, ASME Code Section III, is
referenced by NUREG-1536, “Standard Review Plan for Dry Cask Storage Systems,” as an
acceptable standard for the design and fabrication of dry storage casks. However, since dry
storage casks are not pressure vessels, ASME Code Section III, cannot be implemented
without allowing some alternatives to its requirements.

Revision 1

The staff has broadened the technical basis for the storage of spent fuel including assemblies
with average burnups exceeding 45 GWd/MTU. This revision to Interim Staff Guidance No. 11
(ISG-11) addresses the technical review aspects of and specifies the acceptance criteria for
limiting spent fuel reconfiguration in storage casks. It modifies the previous revision of the ISG
in three ways: (1) by clarifying the meaning of some of the acceptance criteria contained in

Fuel rod buckling analyses under bottom end drop conditions have traditionally been performed
to demonstrate integrity of the fuel following a cask drop accident. The methodology described
by Lawrence Livermore National Laboratory (LLNL) to analyze the buckling of irradiated spent
fuel assembly under a bottom end drop in their report UCID-21246 is a simplified approach. It
assumed that buckling occurred when the fuel rod segment between the bottom two spacer
grids reached the Euler buckling limit. The weight of fuel pellets was neglected in the analysis;

The purpose of this guidance is to, (1) clarify the meaning of a real individual as used in 10
CFR 72.104, (2) specify how the applicant may perform dose evaluations beyond the controlled
area for site-specific and general Independent Spent Fuel Storage Installation (ISFSI) licenses,
and (3) clarify standard review plan (SRP) text regarding dose calculations outside the
controlled area.

Guidance regarding supplemental shielding that may be installed at an independent
spent fuel storage installation (ISFSI) to meet the requirements of 10 CFR 72.104(a).

Due, in part, to a number of material-related issues identified during dry cask storage system
(DCSS) and transportation package application reviews and field implementation, the staff has
recognized the need for specific guidance for the review of materials selected by the applicant
for its DCSS or transportation package.