[Federal Register Volume 60, Number 216 (Wednesday, November 8, 1995)]
[Notices]
[Pages 56359-56361]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-27624]
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NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; Boraflex Degradation in Spent
Fuel Pool Storage Racks (M91447)
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of opportunity for public comment.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue
a generic letter concerning Boraflex degradation in spent fuel pool
storage racks. The purpose of the proposed generic letter is to request
that licensees who use Boraflex as a neutron absorber in their spent
fuel storage racks (1) assess the capability of the boraflex to
maintain a 5 percent subcriticality margin and (2) submit a plan of
action if this subcriticality margin cannot be maintained by the
Boraflex material because of current or projected degradation. The NRC
is seeking comment from interested parties regarding both the technical
and regulatory aspects of the proposed generic letter presented under
the Supplementary Information heading.
The proposed generic letter was endorsed by the Committee to Review
Generic Requirements (CRGR) on September 26, 1995. The relevant
information that was sent to the CRGR will be placed in the NRC Public
Document Room. The NRC will consider comments received from interested
parties in the final evaluation of the proposed generic letter. The
NRC's final evaluation will include a review of the technical position
and, as appropriate, an analysis of the value/impact on licensees.
Should this generic letter be issued by the NRC, it will become
available for public inspection in the NRC Public Document Room.
DATES: Comment period expires December 8, 1995. Comments submitted
after this date will be considered if it is practical to do so, but
assurance of consideration cannot be given except for comments received
on or before this date.
ADDRESSEES: Submit written comments to Chief, Rules Review and
Directives Branch, U.S. Nuclear Regulatory Commission, Mail Stop T-6D-
69, Washington, DC 20555-0001. Written comments may also be delivered
to 11545 Rockville Pike, Rockville, Maryland, from 7:30 am to 4:15 pm,
Federal workdays. Copies of written comments received may be examined
at the NRC Public Document Room, 2120 L Street, N.W. (Lower Level),
Washington, D.C.
FOR FURTHER INFORMATION CONTACT: Laurence I. Kopp (301) 415-2879.
SUPPLEMENTARY INFORMATION:
NRC Generic Letter 95-XX: Boraflex Degradation in Spent Fuel Pool
Storage Racks (M91447)
Addressees
All holders of operating licenses for nuclear power reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this
generic letter to request that each addressee that uses Boraflex as a
neutron absorber in its spent fuel storage racks (1) assess the
capability of the Boraflex to maintain a 5 percent subcriticality
margin and (2) submit to the NRC a plan describing its proposed actions
if this subcriticality margin cannot be maintained by Boraflex material
because of current or projected future Boraflex degradation.
Background
Degradation of Boraflex has been previously addressed by the NRC in
Information Notice (IN) 87-43, ``Gaps in Neutron-Absorbing Material in
High-Density Spent Fuel Storage Racks,'' September 8, 1987, IN 93-70,
``Degradation of Boraflex Neutron Absorber Coupons,'' September 10,
1993, and IN 95-38, ``Degradation of Boraflex Neutron Absorber in Spent
Fuel Storage Racks.'' The Electric Power Research Institute (EPRI) has
been studying the phenomenon of Boraflex degradation for several years
and recently issued EPRI TR-103300, ``Guidelines for Boraflex Use in
Spent-Fuel Storage Racks,'' December 1993, identifying two issues with
respect to using Boraflex in spent fuel storage racks. The first issue
related to gamma radiation-induced shrinkage of Boraflex and the
potential to develop tears or gaps in the material. This phenomenon is
typically accounted for in criticality analyses of spent fuel storage
racks. The second issue concerned long-term Boraflex performance
throughout the intended service life of the racks as a result of gamma
irradiation and exposure to the wet pool environment.
Description of Circumstances
Palisades Nuclear Power Station
During the removal of several Boraflex surveillance coupons from
the Palisades spent fuel pool in August 1993, a loss of as much as 90
percent of the Boraflex was observed and has been attributed to
exposure to high-level gamma radiation in conjunction with interaction
with the pool water. The Boraflex in these coupons was sandwiched and
bolted between two stainless steel strips, allowing a relatively large
area of Boraflex to be exposed to the pool water environment and flow.
Neutron attenuation testing (blackness tests) of the actual Palisades
storage racks indicated that because of the relatively watertight
Boraflex panel enclosures, there was no similar degradation.
South Texas Project
The results of blackness tests performed in August 1994 at South
Texas indicated that the Boraflex was degraded, as evidenced by gaps
and/or localized washout of the boron content in 20 of the 37 storage
cells tested. Of the eight cells that had been designated to receive an
accelerated gamma dose, five cells exhibited substantial degradation
(0.91 to 1.37 m [3 to 4.5 ft]). The licensee postulated that the
degradation mechanism was washout-accelerated dissolution of the
Boraflex caused by pool water flow through the panel enclosures. As a
justification for continued operation, the licensee has placed
restrictions on the use of the
[[Page 56360]]
degraded storage cells to ensure compliance with the required
subcriticality margin. In addition, a long-term neutron absorption
panel management plan is being developed, as well as a dose-to-
degradation correlation that will aid in establishing restrictions for
the use of the spent fuel racks.
Fort Calhoun Station
As part of the Fort Calhoun Station rerack project, the old spent
fuel storage racks containing Boraflex were removed and disassembled in
December 1994 to determine the condition of the Boraflex. The new
storage racks do not contain Boraflex. The licensee inspected two cells
from the removed Boraflex racks which had experienced the highest gamma
flux since 1983. Only 40 percent of the Boraflex remained in one of the
panels from these cells while another panel in the same cell exhibited
no loss of Boraflex. An adjacent cell had a panel which had some
Boraflex loss but subsequent attenuation and density tests confirmed
that the average boron-10 areal density still exceeded the material
minimum certifications. No other storage cells exhibited as significant
a loss of Boraflex. The licensee has determined that there was
sufficient Boraflex in the walls of each cell to meet the minimum
requirements in the design-basis criticality analysis.
Discussion
Experimental data from test programs, including blackness tests
performed at various boiling-water reactor (BWR) and pressurized-water
reactor (PWR) spent fuel storage pools, confirmed that when Boraflex is
exposed to gamma radiation, the material may shrink by as much as 3 to
4 percent. Shrinkage saturates at an integrated gamma exposure of about
1 to 2 x 10\10\ cGy (1 to 2 x 10\10\ rad). The application of realistic
assumptions based on these tests has demonstrated that the reactivity
effects of Boraflex shrinkage and gaps are very small and can generally
be accommodated within the existing design basis of most storage racks
(EPRI TR-101986, ``Boraflex Test Results and Evaluation,'' February
1993).
Data from laboratory tests and spent fuel pool silica measurements
have identified a second factor that could affect storage rack service
life: the potential gradual release of silica from Boraflex following
gamma irradiation and long-term exposure to the wet pool environment.
When Boraflex is subjected to gamma radiation in the pool's aqueous
environment, the silicon polymer matrix becomes degraded and silica
filler and boron carbide are released. Since irradiated Boraflex
typically contains 46 percent of silica, 4 percent of polydimethyl
siloxane polymer and 50 percent of boron carbide by weight, the
presence of silica in the pool indicates depletion of boron carbide
from Boraflex. The loss of boron carbide from Boraflex is characterized
by slow dissolution of the silicon polymer from the surface of the
Boraflex and a gradual thinning of the material. In a typical spent
fuel pool, the irradiated Boraflex represents a significant source of
silica (several thousand kilograms) and is the most likely source of
pool silica contamination. The boron carbide loss, of course, can
result in a significant increase in the reactivity of the storage
racks. An additional consideration is the potential for silica transfer
through the fuel transfer canal into the reactor core during refueling
operations and its effect on the fuel clad heat transfer capability.
EPRI TR-103300 has identified several factors that influence the
rate of silica release from Boraflex. The access of water to and around
the Boraflex panels is perhaps the most significant factor influencing
the rate of silica dissolution from Boraflex. Because of the different
rack designs, this water access will vary from plant to plant. The rate
of dissolution also increases with higher pool temperature and gamma
exposure, suggesting that pool temperatures be maintained as low as
practical and that freshly discharged fuel assemblies should not be
placed in the same storage cells at each refueling outage. Once silica
reaches an equilibrium value, the rate of dissolution essentially
stops. However, when water purification systems are used to remove
silica from the pool water, the solubility equilibrium becomes
unbalanced and panel dissolution resumes.
Because Boraflex is used in spent fuel storage racks for
nonproductive absorption of neutrons, a reduction in the amount of
Boraflex could result in an increase in the reactivity of the spent
fuel pool configuration, which may approach, or even exceed, the
current NRC acceptance criterion of keff no greater than 0.95. The
NRC has established this 5 percent subcriticality margin to comply with
General Design Criterion (GDC) 62 of Appendix A to Part 50 of Title 10
of the Code of Federal Regulations (10 CFR Part 50), which requires the
prevention of criticality in fuel storage and handling. Those plants
that have installed storage racks containing Boraflex have the 5
percent subcriticality margin included in the plant technical
specifications and/or a written commitment to meet this subcriticality
margin, as reflected in the plant updated final safety analysis report
(FSAR). The technical specifications for most other operating power
reactors also include this 5 percent subcriticality requirement.
Safety Assessment
On the basis of test and surveillance information from plants that
have detected areas of Boraflex degradation, no safety concern exists
that warrants immediate action. Boraflex dissolution appears to be a
gradual and localized effect forewarned by relatively high silica
levels in the pool water. Because of the safety margin present in spent
fuel storage pools, compliance with the required subcriticality margin
(or conformance with the same margin to which licensees have committed
in their updated FSARs) can be expected to be maintained during the
initial stage of Boraflex degradation. This safety margin is due to the
5 percent subcriticality margin assumed in the analysis, the generally
lower reactivity of stored fuel than that assumed in the safety
analysis, and, in the case of PWRs, the presence of borated water in
the pool. However, to verify compliance with both the regulatory
requirements of GDC 62 and the 5 percent subcriticality margins, either
contained in the technical specifications or committed to in the
updated FSARs, and to maintain an appropriate degree of defense-in-
depth measures, the NRC staff has concluded that it is appropriate for
licensees to submit the following information.
Requested Information
All licensees of power reactors with spent fuel pool storage racks
containing the neutron absorber Boraflex are requested to provide a
description of the physical condition of the Boraflex, including any
deterioration, on the basis of current as well as future projected
accumulated gamma exposure and possible water ingress to the Boraflex
and state whether a subcritical margin of 5 percent can be maintained
for the life of the racks in unborated water. All licensees are further
requested to submit to the NRC a description of any proposed actions to
monitor or confirm that this 5 percent subcriticality margin can be
maintained for the lifetime of the storage racks and describe what
corrective actions will be taken in the event it cannot be maintained.
Licensees should describe the results from any previous blackness tests
and state whether blackness testing will be periodically performed. Any
abnormal pool silica levels should also be described. All licensees are
requested to submit the information to the NRC to
[[Page 56361]]
ensure that the onsite storage of spent fuel is in compliance with GDC
62 for the prevention of criticality in fuel storage and handling and
with the 5 percent subcriticality margin position of the NRC staff to
assure compliance with GDC 62.
Required Response
All addressees are required to submit a written response to the
information requested above within 120 days of the date of this generic
letter. If an addressee chooses not to respond to specific questions,
an explanation of the reason and a description of any proposed
alternative course of action should be provided, as well as the
schedule for completing the alternative course of action (if
applicable), and the safety basis for determining the acceptability of
the planned alternative course of action.
Address the required written reports to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, D.C. 20555, under
oath or affirmation under the provisions of Section 182a, Atomic Energy
Act of 1954, as amended, and 10 CFR 50.54(f). In addition, submit a
copy to the appropriate regional administrator.
Backfit Discussion
This generic letter only requires information from the addresses
under the provisions of Section 182a of the Atomic Energy Act of 1954,
as amended, and 10 CFR 50.54(f). Therefore, the staff has not performed
a backfit analysis. The information requested will enable the NRC staff
to determine whether licensees are complying with the current licensing
basis for the facility with respect to GDC 62 for the prevention of
criticality in fuel storage and handling and 5 percent subcriticality
margins either contained in the technical specifications, or committed
to in the updated FSARs, of plants containing Boraflex in the spent
fuel storage racks. The staff is not establishing a new position for
such compliance in this generic letter. Therefore, this generic letter
does not constitute a backfit and no documented evaluation or backfit
analysis need be prepared.
Federal Register Notification
(To be completed after the public comment period.)
Paperwork Reduction Act Statement
The information collections contained in this request are covered
by the Office of Management and Budget clearance number 3150-0011,
which expires July 31, 1997. The public reporting burden for this
collection of information is estimated to average 150 hours per
response, including the time for reviewing instructions, searching
existing data sources, gathering and maintaining the data needed, and
completing and reviewing the collection of information. Send comments
regarding this burden estimate or any other aspect of this collection
of information, including suggestions for reducing this burden, to the
Information and Records Management Branch, (T-6F33), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and to the Desk
Officer, Office of Information and Regulatory Affairs, NEOB-10202
(3150-0011), Office of Management and Budget, Washington, DC 20503.
Dated at Rockville, Maryland, this 2nd day of November, 1995.
For the Nuclear Regulatory Commission.
Dennis M. Crutchfield,
Director, Division of Reactor Program Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 95-27624 Filed 11-7-95; 8:45 am]
BILLING CODE 7590-01-P
