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                                       UNITED STATES
                               NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS
                           OFFICE OF NUCLEAR REACTOR REGULATION
                               WASHINGTON, D.C.  20555-0001

                                       June 26, 1997


NRC INFORMATION NOTICES 97-39:  INADEQUATE 10 CFR 72.48 SAFETY EVALUATIONS OF  
                              INDEPENDENT SPENT FUEL STORAGE INSTALLATIONS


Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.  All holders of licenses for independent spent fuel storage
installations (ISFSIs).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to inadequate safety evaluations performed under
Section 72.48 of Title 10 of the Code of Federal Regulations (10 CFR 72.48). 
It is expected that the recipients will review this information notice for
applicability to their facilities and consider actions, as appropriate. 
However, suggestions contained in this information notice are not NRC
requirements; therefore, no specific action or written response is required.

Background

Section 72.48, "Changes, tests, and experiments," states that a holder of an
ISFSI license may make changes in the ISFSI described in the safety analysis
report (SAR), may make changes in the procedures described in the SAR, or may
conduct tests or experiments not described in the SAR, without prior NRC
approval, unless the proposed change, test, or experiment involves a change in
the license conditions incorporated in the license, an unreviewed safety
question, a significant increase in occupational exposure, or a significant
unreviewed environmental impact.  A proposed change is deemed to involve an
unreviewed safety question if the probability of occurrence or the
consequences of an accident or malfunction of equipment important to safety
previously evaluated in the SAR may be increased, if a possibility for an
accident or malfunction of a different type than any evaluated previously in
the SAR may be created, or if the margin of safety as defined in the basis for
any technical specification is reduced.  The licensee is required to maintain
records of changes made to the ISFSI, which include a written safety
evaluation that provides the bases for the determination that each change,
test, or experiment does not involve an unreviewed safety question. 

  

9706260174.                                                                     IN 97-39
                                                                     June 26, 1997        
                                                                   Page 2 of 4

Description of Circumstances

The NRC performs inspections of ISFSI licensees to verify adequate
implementation of 10 CFR 72.48 requirements.  Recent NRC inspections at
several ISFSI licensees have identified violations of these requirements. 
These violations involve failure by the licensees to perform 10 CFR 72.48
safety evaluations when applicable and failure to document sufficient
technical justification to support safety evaluation conclusions.  A summary
of inspection findings follows.

The NRC issued a violation to Arkansas Nuclear One (ANO), owned and operated
by Entergy Operations, Inc.  The violation was based on an inspection
conducted at ANO (Inspection Report Nos. 50-313/96-25; 72-13/96-02). The
licensee has a Part 72 general license to store spent fuel at an ISFSI and
uses the Sierra Nuclear Corporation ventilated storage cask, model VSC-24. 
The inspection identified 12 nonconformances that were not evaluated in
accordance with 10 CFR 72.48.  The licensee considered nonconformances as one-
time changes to its SAR that did not require a safety evaluation because they
were not permanent design changes.  The licensee accepted the nonconformances
as "use as-is" without addressing 10 CFR 72.48 requirements.  The inspectors
concluded that the licensee did not adequately evaluate the conditions to
ensure that an unreviewed safety question did not exist.  The licensee
reevaluated the identified nonconformances under 10 CFR 72.48 requirements and
did not identify any unreviewed safety questions.  The licensee also
reevaluated nine other nonconformances that had originally been resolved as
"use as-is" and did not identify any unreviewed safety questions.  In
addition, the licensee revised its procedures to require an engineering
evaluation for a "use as-is" determination, which requires a 10 CFR 72.48
review, when applicable.

The NRC issued a violation to Point Beach Nuclear Plant (PBNP), owned and
operated by Wisconsin Electric Power Company.  The violation was based on
inspections conducted at PBNP (Inspection Report Nos. 50-266/301-95008; 50-
266/301-95014).  The licensee has a Part 72 general license to store spent
fuel at an ISFSI and uses the VSC-24 cask.  The inspections identified several
one-time dimensional changes and other changes that either should have
received a 10 CFR 72.48 safety evaluation or that did not receive an adequate
safety evaluation.  In one such safety evaluation, the licensee did not
adequately discuss the geometric effects on criticality when reducing the
thickness of spent fuel guide sleeves in the VSC-24 fuel basket.  The licensee
subsequently revised its safety evaluation to address the geometric effects on
criticality.  In another case, the licensee did not perform a safety
evaluation to allow the VSC-24 to be reflooded with a flow rate between 34
L/min (9 gpm) and 42 L/min (11 gpm) during loading and unloading.  This
oversight was of concern because the pressure transient calculation for the
cask pressure was only valid for a flow rate of 38 L/min (10 gpm) or less. 
The licensee subsequently revised reflooding procedures to ensure that a flow
rate of 38 L/min (10 gpm) is not exceeded.  In response to the inspection
findings, the licensee screened all other design and procedure changes made to
its ISFSI and performed full 10 CFR 72.48 safety evaluations, as necessary. 
The licensee did not identify any unreviewed safety questions..                                                                     IN 97-39
                                                                     June 26, 1997
                                                                     Page 3 of 4

Another violation was issued to PBNP, in part, because the licensee failed to
perform adequate safety evaluations for two procedures.  The violation was
based on an augmented inspection team inspection at PBNP (Enforcement Action
[EA] 96-273 and Inspection Report No. 50-266/301-96005).  The licensee did not
perform a 10 CFR 72.48 safety evaluation for a lifting evolution that created
a potential for dropping the VSC-24 multi-assembly sealed basket (MSB)
transfer cask off the top of the ventilated concrete cask, an accident not
described in the SAR.  The licensee also did not provide sufficient technical
justification to support conclusions in a safety evaluation for an MSB
weighing procedure.  The safety evaluation did not include supporting
information that ensured that the shield lid would not be inadvertently
removed from the MSB and expose the workers to spent fuel.  Before lifting the
lid, the licensee developed a new weighing procedure that did not create the
potential to inadvertently remove the lid.

The NRC issued a violation to Prairie Island Nuclear Plant (PI), owned and
operated by Northern States Power Company.  The violation was based on an
inspection conducted at PI (Inspection Report Nos. 50-282/95002; 50-306/95002;
72-10/95002).  The licensee has a Part 72 site-specific license to store spent
fuel in the Transnuclear, Inc., metal storage cask, model TN-40.  Upon initial
inspection, the inspectors determined that the licensee did not have a
procedure in place for conducting 10 CFR 72.48 safety evaluations.  The
licensee subsequently revised its procedures to incorporate 10 CFR 72.48
safety evaluations into its existing 10 CFR 50.59 review process.  The
inspectors then reviewed a sample safety evaluation and found that it was not
adequate.  The sample did not address 10 CFR 72.48 requirements on a TN-40
lifting beam, which is described in its ISFSI SAR.  The licensee believed that
only a 10 CFR 50.59 safety evaluation was required because the cask-handling
equipment was used in the Auxiliary Building.  The inspectors discussed with
the licensee the need to conduct a 10 CFR 72.48 safety evaluation if changes
were made to any equipment that was described in the ISFSI SAR, independent of
whether the equipment was used to handle the cask in the Auxiliary Building. 
The licensee subsequently performed a 10 CFR 72.48 safety evaluation on the
TN-40 lifting beam.

Discussion

The NRC inspects licensees to assess the adequacy of their programs to perform
safety evaluations in accordance with 10 CFR 72.48.  Holders of an ISFSI
license are required by 10 CFR 72.48 to perform a safety evaluation when
changing a component or a procedure described in the ISFSI SAR and, in the
case of a general licensee, the cask SAR.  The licensee's determination that a
modification to an ISFSI does not involve an unreviewed safety question
provides confidence that the bases on which the NRC issued a license to
operate an ISFSI are preserved. 

An NRC-licensed ISFSI that uses dry-storage casks is a passive system in which
its structural, criticality-control, thermal, and shielding performances
depend on the detailed drawings and descriptions provided as the design bases
in the SAR.  It is important that the licensee perform an adequate 10 CFR
72.48 safety evaluation of any modification in the ISFSI SAR or cask SAR,
including any changes in dimensions, materials, and procedures.  In addition,
one-time changes to the ISFSI SAR or the cask SAR constitute a modification
that also requires an adequate 10 CFR 72.48 safety evaluation..                                                                     IN 97-39
                                                                     June 26, 1997
                                                                     Page 4 of 4


Related Generic Communications

IN 95-29,  "Oversight of Design and Fabrication Activities for Metal
Components Used in Spent Fuel Dry Storage Systems," dated June 7, 1995.

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-
Assembly Sealed Basket," dated May 31, 1996.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Material Safety and Safeguards (NMSS) project manager.


signed by S.H. Weiss for                             signed by 

Marylee M. Slosson, Acting Director                 William F. Kane, Director 
Division of Reactor Program Management              Spent Fuel Project Office
Office of Nuclear Reactor Regulation                Office of Nuclear Material 
                                                      Safety and Safeguards

Technical contacts:  M. Waters, NMSS
                     301-415-3875
                     E-mail:  [email protected]

                     V. Hodge, NRR 
                     301-415-1861
                     E-mail:  [email protected]
```