[Federal Register Volume 70, Number 215 (Tuesday, November 8, 2005)]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-22199]
NUCLEAR REGULATORY COMMISSION
Notice of Availability of Interim Staff Guidance Documents for
Fuel Cycle Facilities
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of availability.
FOR FURTHER INFORMATION CONTACT: James Smith, Project manager,
Technical Support Group, Division of Fuel Cycle Safety and Safeguards,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20005-0001. Telephone: (301) 415-
6459; fax number: (301) 415-5370; e-mail: [email protected].
The Nuclear Regulatory Commission (NRC) continues to prepare and
issue Interim Staff Guidance (ISG) documents for fuel cycle facilities.
These ISG documents provide clarifying guidance to the NRC staff when
reviewing licensee integrated safety analysis, license applications or
amendment requests or other related licensing activities for fuel cycle
facilities under subpart H of 10 CFR part 70. FCSS-ISG-08 has been
issued and is provided for information.
The purpose of this notice is to provide notice to the public of
the issuance of FCSS-ISG-08, Revision 0, which provides guidance to NRC
staff to address accident sequences that may result from natural
phenomena hazards relative to license application or amendment request
under 10 CFR Part 70, Subpart H. FCSS-ISG-08, Revision 0, has been
approved and issued after a general revision based on NRC staff and
public comments on the initial draft.
III. Further Information
The document related to this action is available electronically at
the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this site, you can access the NRC's Agencywide
Documents Access and Management System (ADAMS), which provides text and
image files of NRC's public documents. The ADAMS ascension number for
the document related to this notice is provided in the following table.
If you do not have access to ADAMS or if there are problems in
accessing the document located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or
by e-mail to [email protected].
Interim staff guidance ADAMS Accession No.
FCSS Interim Staff Guidance-08, Revision ML052650305
This document may also be viewed electronically on the public
computers located at the NRC's PDR, O 1 F21, One White Flint North,
11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction
contractor will copy documents for a fee. Comments on these documents
may be forwarded to James Smith, Project Manager, Technical Support
Group, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20005-0001. Comments can also be submitted by telephone,
fax, or e-mail which are as follows: Telephone: (301) 415-6459; fax
number: (301) 415-5370; e-mail: [email protected].
Dated at Rockville, Maryland this 27th day of October 2005.
For the Nuclear Regulatory Commission.
Melanie A. Galloway,
Chief, Technical Support Group, Division of Fuel Cycle Safety and
Safeguards, Office of Nuclear Material Safety and Safeguards.
Attachment--FCSS Interim Staff Guidance-08, Revision 0, Natural
Prepared by Division of Fuel Cycle Safety and Safeguards, Office of
Nuclear Material Safety and Safeguards
Additional guidance is required to address accident sequences that
may result from natural phenomena hazards in the context of a license
application or an amendment request under Title 10 Code of Federal
Regulations (10 CFR) part 70, subpart H.
This Interim Staff Guidance (ISG) provides additional guidance for
reviewing the applicant's (or licensee's) evaluation of natural
phenomena hazards up to and including ``highly unlikely'' events for
both new and existing facilities.
The performance requirements of 10 CFR 70.61 for facilities
processing special nuclear materials require that individual accident
sequences resulting in high consequences to workers and the public be
``highly unlikely'' and that sequences resulting in intermediate
consequences to these receptors be ``unlikely.'' Although the threshold
levels that differentiate high consequence events from intermediate
consequence events are established in the regulations, the definitions
of ``highly unlikely'' and ``unlikely'' are not. Definitions of these
terms must be described in the integrated safety analysis (ISA) summary
submitted by applicants and licensees according to 10 CFR 70.65(b)(9)
and subjected to staff approval. Further description of the acceptance
criteria for the definitions of these terms can be found in Chapter 3
of NUREG-1520, ``Standard Review Plan for the Review of a License
Application for a Fuel Cycle Facility.''
The implementation of these requirements may vary somewhat due to
different definitions of likelihood proposed by different applicants
licensees).\1\ The consequence thresholds of the performance
requirements (except for chemical releases) are specified
quantitatively in the regulation. The regulation and its performance
requirements pertain to existing facilities as well as proposed
facilities and apply to man-made external hazards and natural phenomena
hazards as well as process hazards. However, new facilities and new
processes at existing facilities must also address 10 CFR 70.64
requirements which includes the baseline design criterion for natural
phenomena hazards (10 CFR 70.64(a)(2)). This baseline design criterion
requires that ``the design must provide for adequate protection against
natural phenomena with consideration of the most severe documented
historical events for the site.'' The Statement of Considerations
(Reference 2) describes the application of the baseline design criteria
as consistent with good engineering practice, which dictates that
certain minimum requirements should be applied to design and safety
considerations. The baseline design criteria must be applied to the
design of new facilities and new processes at existing facilities, but
does not require retrofits to existing facilities or existing processes
(e.g., those housing or adjacent to the new processes). Also included
in 10 CFR 70.64(b) is a requirement for incorporation of defense-in-
depth in design and a requirement to prefer engineered controls over
\1\ For natural phenomena, deterministically defined events such
as the probable maximum flood (PMF) or safe shutdown earthquake
(SSE) which are used as reactor design bases can also be applied to
10 CFR Part 70 facilities as ``highly unlikely'' events. The actual
probability (or likelihood) of such events may be difficult to
define quantitatively and varies from site to site.
New structures associated with facilities being reviewed, such as
the gas centrifuge facilities and the Mixed Oxide Fuel Fabrication
Facility (MOX), will be designed and constructed to meet the seismic
regulatory requirements. Hence, these facilities and additional new
facilities to be licensed under 10 CFR part 70 are not expected to
present designs with seismic deficiencies. New facilities can also be
expected to be above a ``highly unlikely'' flood such as the PMF and
can be expected to withstand tornado winds and missiles, if necessary.
Most structures at existing nuclear fuel cycle facilities are built
to a model building code which includes meeting a design basis
earthquake having an exceedance probability of 2 x 10-3 per
year to less than 10-3 per year (Department of Energy (DOE)
Standard-1020-2002, Appendix C). Existing facilities are generally
sited above the 100-year floodplain and are designed for wind as well
as snow and ice loading as specified in applicable building codes.
Extreme natural events such as ``highly unlikely'' floods and/or
earthquakes have not been calculated for many existing sites, and it
would be expensive and time consuming to do so.
The staff believes that many existing facilities can be shown to be
in compliance with, or, at least, near compliance with, the performance
requirements of the regulation by accounting for conservatisms in the
seismic, flooding, and wind design of the facility. In addition,
relatively minor engineered improvements and administrative measures
may further enhance safety, at least with respect to the public and
other off-site receptors.
Potential damage to and/or failure of items relied on for safety
(IROFS) due to ground movement and/or the seismic response of adjacent
or interior IROFS must be considered in the ISA/ISA Summary accident
sequence evaluations. Damage or failures that also should be considered
1. Seismic-induced failure of a facility component which is not an
IROFS but which can fall and damage an IROFS, for example, a heavy load
drop from a crane on a container.
2. Displacement of adjacent IROFS during a seismic event, causing
them to pound together.
3. Displacement of adjacent components resulting in failure of
connecting pipes or cables resulting in flooding, fires, and/or
releases of radiological or chemical materials.
Seismic event evaluations must also consider potential multiple
failure of IROFS. For example, multiple failures of tanks.
DOE has also recognized the difference between earthquake design
probability and the probability that a safety component cannot perform
its function. To quantify this difference, DOE has developed a risk
reduction factor, R, as the ratio between the seismic hazard exceedance
probability and the performance goal probability. Conservatism in
nuclear facility design arising from factors such as use of prescribed
analysis methods, specification of material strengths, and limits on
inelastic behavior explains at least part of this apparent reduction in
actual risk. This risk reduction factor is discussed in Appendix C of
DOE-STD-1020-2002 (Reference 3).
For a consequence to occur to the public or external site workers,
licensed material or hazardous chemicals that could affect the safety
of licensed material must be released through at least one, and often
two, confinement barriers, for example:
1. Storage containers, gloveboxes, tanks, or handling devices,
2. Ventilation system dynamic confinement and filtration, and/or
3. Building structural shell.
Criticalities, on the other hand, may result from the introduction
of a moderator or loss of safe geometric control of confined materials.
By using risk reduction factors calculated for a facility and its
specific components and/or making estimates of the degree of failure by
comparison with the observed behavior of similarly constructed
buildings during severe earthquakes, reasonable scenarios can be
postulated. These scenarios may not release all the material at risk or
present an unimpeded leak path to receptors. For example, some
facilities might be able to show that even with an earthquake that is
``highly unlikely'' only certain types of containers or confinement
systems are likely to be breached. If the amount of material contained
in such containers is variable, then that probabilistic component may
be factored into the overall likelihood of the accident sequence. If
employing some of these mitigating considerations to the analysis
requires reliance on special containers or procedures, then additional
IROFS may also be needed. Another factor to be considered is the likely
rate of release based on the damage sustained. For example, some
facilities may lose dynamic confinement but maintain building
integrity. In some processes, radiological and/or chemically hazardous
material is held inside its primary containment at subatmospheric
pressure. In these cases, even though the primary containments are
inside a structure designed to withstand less than a ``highly
unlikely'' earthquake, the subatmospheric conditions may be sufficient
to limit both facility worker and off-site doses in the event of a
greater earthquake. For example, an earthquake that results in limited
subatmospheric containment losses may allow adequately trained workers
to evacuate and/or take mitigative actions. The buildings containing
cylinders of liquid UF6 at gas centrifuge facilities are
designed for a ``highly unlikely'' earthquake. In addition, some
buildings at one of the proposed facilities are equipped with a
seismically-activated interlock (an IROFS) that will shut off the
buildings' heating, ventilation, and air conditioning system during an
thus limiting any leakage of UF6 to the outside.
Most fuel cycle licensees do not require large quantities of
cooling water and, therefore, do not need to be located near large
bodies of water. A site licensed under 10 CFR Part 70 does not need to
meet prescriptive flood protection requirements but does have to meet
the performance requirements for all credible events including
flooding. A site meeting the flood protection requirements of a
commercial reactor should be considered as being designed or located
adequately to withstand a ``highly unlikely'' flooding event. NUREG-
1407, ``Procedural and Submittal Guidance for the Individual Plant
Examination of External Events for Severe Accident Vulnerability,''
Section 2.4, states that the design basis flood (which for river sites
is the probable maximum flood) as described in Regulatory Guide 1.59,
``Design Basis Flooding for Nuclear Power Plants,'' is estimated to
have an exceedance frequency of less than 10-5 per year.
Sites that do not meet this level of protection can still meet the 10
CFR 70.61 performance requirements but have to be considered on an
In evaluating the effects of flooding on existing facilities, the
following flood-related hazards should be considered:
Inundation and hydrostatic loading
Sedimentation and erosion
Upstream Dam Failures
Inundation and hydrostatic loading
Erosion and sedimentation
Precipitation/Local Storm Runoff
Inundation (local ponding) and hydrostatic loading
Dynamic loads (flash flooding)
Tsunami, Seiche, Hurricane Storm Surge
Inundation and hydrostatic loading
Methods for determining these flooding and water-related effects
for reactor sites are described in American National Standards
Institute/American Nuclear Society 2.8, ``Determining Design Basis
Flooding at Power Reactor Sites.'' These methods can be applied to 10
CFR 70.61 analyses with less conservatism in some of these parameters.
A standard siting requirement for residential and commercial
developments is to be above the 100-year floodplain. For large river
basins, warning time and time to secure materials and evacuate
personnel will probably be available. For small streams there may be
relatively little warning in regard to thunderstorms and localized
rainfall. In such cases, rapid actions may be the only administrative
protection available. In evaluating the effectiveness of proposed
protection, the effects of inundation, hydrostatic loading, erosion,
and sedimentation will need to be evaluated. At a minimum, this would
require that criticality events be prevented and materials remain
confined within site structures.
At some sites, a delineation of the 500-year floodplain may also be
available. If the site is above the 500-year floodplain, flooding may
be considered an unlikely event \2\ depending on the quality of the
estimate. In this category, criticality events should still be
prevented, but the breaching of a limited number of material containers
may be allowable under the performance requirements (up to 25 rem for
the public, up to 100 rem for workers, and a specified release limit)
for events, that in terms of likelihood, are between ``unlikely'' and
\2\ Even if the licensee defines unlikely as less than
10-3 per year for the process sequences in the ISA
Summary, the conservative assumptions inherent in most flood plain
hydrologic studies such as those performed for Federal Emergency
Management Agency flood insurance rate maps should justify the
consideration of flooding above the 500-year floodplain as an
In addition to the facility's location relative to the 100-year or
500-year floodplains, the effects of local intense precipitation and
snow load should be considered. Local intense precipitation especially
with snow can result in roof collapse and localized site flooding.
Normally, protection from local precipitation and snow is relatively
easy to achieve in roof design and local site drainage design.
Wind and Tornado Loading
Wind design for an existing facility if prescribed by an applicable
building code would have an annual exceedance probability of greater
than or equal to 2 - 10-2. At such relatively high
probabilities, tornado design criteria are not specified. However,
depending on the geographical location of the facility, the effects of
a tornado with an annual exceedance probability of 10-5 or
greater may need to be considered.
Wind forces on walls of structures should be determined using
appropriate pressure coefficients, gust factors, and other site-
specific adjustments. If the wind is likely to blow inside the
structure, either through design or wind-driven missile vulnerability,
the effects of wind on internal IROFS requires consideration. If the
winds are from a tornado, the effects of the atmospheric pressure
change (APC) associated with the tornado must be considered. Normally,
ventilation systems are most vulnerable to APC but windows, buried
tanks, and sand filters can also be affected.
For straight winds, hurricanes, and weak tornadoes, missile
criteria as specified in Table 3-3 of DOE-STD-1020-2002 (Reference 3)
may be considered. The missile specified is a 15 pound, 2 inches by 4
inches plank at a specified elevation and impact velocity. For
facilities which may be subjected to more severe tornado missiles, the
guidance in Tables 3-4 and/or 3-5 of DOE-STD-1020 may be followed. For
the tornado, a 3,000 pound automobile rolling and tumbling on the
ground should also be considered. For such evaluations, the probability
of the entire sequence should be considered, and missile criteria from
either Tables 3-4 or 3-5 of DOE-STD-1020 may be used as appropriate.
Considerations for Existing Processes at Existing Facilities
Existing processes at existing facilities are not required to
address 10 CFR 70.64 baseline design criteria. They must still meet the
performance requirements of 10 CFR 70.61 including accidents caused by
natural phenomena, for which the staff may require additional IROFS to
meet the performance requirements. For existing facilities, additional
administrative controls/IROFS can be used to meet the performance
requirements without the need for design features normally required by
accepted engineering practice. For plants where near compliance can be
obtained and complete compliance will be relatively costly, an
exemption to the regulation may be requested.
As discussed earlier, many existing 10 CFR Part 70 facilities are
not designed for an earthquake beyond that specified in applicable
building codes. Although this design may provide fairly good seismic
protection to the structure, it may not protect internal equipment.
Also, an existing facility may not be designed to any specific seismic
criteria in which case its ability to withstand earthquakes can only be
estimated based on comparison with similar structures or through
complex structural analysis. In such cases, licensees may add
additional IROFS to meet the performance requirements. An example where
such IROFS (procedures and upgrades) may be effectively implemented
could be a facility where the consequences of a release of licensed
material to the public in a seismic event would be from fires and/or
explosions. In this case, fixes such as seismically qualified flammable
gas shutoff valves or electrical shutoffs might provide a large
decrease in potential seismic consequences.
In regard to flooding, flood elevations beyond that of the 100-year
flood may not have been determined for the site. For sites in close
proximity to a river, these determinations could be expensive and time
consuming. For these cases, flood warning time may allow measures such
as moving material at risk and/or blocking doors and openings in the
Improving a facility's ability to withstand high winds, rain and
snow loads, and exterior fires can likewise be improved with a
combination of administrative procedures and engineered improvements.
Removing material at risk from under walls or roofs that are not
seismically designed can reduce potential releases in case of collapse
from winds or roof loads.
Exemptions to the regulation may still be required for existing
facilities even with administrative and engineered improvements. In
regard to consequences to the public, complete compliance with 10 CFR
70.61 using realistic assumptions should be the goal if obtainable.
Compliance with 10 CFR 70.61 regarding consequences to facility workers
may require a request for an exemption once personnel protective
equipment, emergency procedures, and worker training is accounted for.
In evaluating a request for an exemption to the regulation, the
expected operational life of the facility should also be factored into
the determination of risk.
Considerations for New Processes at Existing Facilities
The design of new processes at existing facilities must address
natural phenomena hazards in accordance with 10 CFR 70.64 (a)(2) as
well as the performance requirements of 10 CFR 70.61. Nevertheless, new
processes at existing facilities may have the same problems in
demonstrating compliance with 10 CFR 70.61 in regard to accident
sequences initiated by natural phenomena as existing facilities based
on the design and/or siting of the original structures. In the case of
new processes, the Nuclear Regulatory Commission staff should expect
compliance with the performance requirements of 10 CFR 70.61 to the
extent possible given the existing facility design and location. New
processes at existing facilities also must meet the requirements of 10
CFR 70.64(b) which requires defense-in-depth and a preference for
engineered controls over administrative controls. However, structural
improvements, permanent flood barriers, and other engineered
improvements which could be considered retrofits cannot be required by
the staff for application to existing structures. New structural
features within existing structures to prevent breaches in containment
in the event of natural phenomena hazards may be considered, however.
An example might be a seismically-designed vault to hold radioactive
materials associated with a new process. In regard to new processes,
engineered controls, where feasible, are preferred over administrative
procedures that might otherwise be proposed for an existing process
with a limited operational lifetime. Such engineered improvements may
not be required for licensing but could be scheduled to replace
administrative procedures or other long-term compensatory measures on a
timely basis after the start of operations. The object is to encourage
engineered safety in new processes compared to equivalent existing
process while recognizing the restraints of the existing structures and
location. Although primarily aimed at reducing risk to the public, the
emphasis on engineered safety may also be applied to worker
consequences in a way consistent with what has been accepted at other
10 CFR 70.61 specifies performance requirements associated with
risks identified by an ISA.
10 CFR 70.64 specifies requirements for new facilities or new
processes at existing facilities including baseline design criteria
(a)(2), ``Natural Phenomena Hazards.''
Technical Review Guidance
In reviewing the applicant's evaluation of the effects of natural
phenomena on its facility, it should be recognized that estimates of
``unlikely'' and ``highly unlikely'' natural phenomena such as the PMF
or SSE may not exist for the particular site. Hence, extrapolation and/
or transposition of extreme event estimates made for other relatively
nearby facilities (such as power reactor sites) should be allowed where
feasible and technically justifiable. In addition, sophisticated
probabilistic tools such as Bayesian analysis or Monte Carlo sampling
methods need not be employed to improve the estimate of likelihoods of
natural phenomena event sequences unless desired by the applicant (or
licensee). For the purpose of determining appropriate values of extreme
events, deterministic events such as the probable maximum flood or safe
shutdown earthquake can be used in place of purely probabilistically
determined ``highly unlikely'' events and may be preferable, depending
on the quality of historical data. Where extreme events need to be
coupled with other probability-driven mechanisms such as the release
fraction or transport pathway, already low likelihood combinations do
not have to be made even less likely with the use of conservative
For existing facilities, due credit should be given to analysis
assumptions and administrative controls, emergency procedures, and
active engineered controls that do not change the design bases of the
facility structures to natural phenomena. If the ISA/ISA Summary
demonstrates that the existing facility is near compliance (within an
order of magnitude of a likelihood threshold or within 50 percent of
meeting a consequence threshold, but not both), an exemption to the
regulation may be considered.
An example evaluation for an amendment request is provided in the
appendix to this ISG.
This guidance should be used to supplement NUREG-1520, Chapter 3,
Integrated Safety Analysis.
U.S. Code of Federal Regulations, Title 10, Energy, Part 70,
``Domestic Licensing of Special Nuclear Material.''
U.S. Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1520,
``Standard Review Plan for the Review of a License Application for a
Fuel Cycle Facility.'' NRC: Washington, D.C. March 2002.
Nuclear Regulatory Commission (U.S.), Washington, D.C.
``Domestic Licensing of Special Nuclear Material; Possession of a
Critical Mass of Special Nuclear Material.'' Federal Register: Vol.
65, No. 181, pp. 56211-562331. September 18, 2000.
NUREG-1407, ``Procedural and Submittal Guidance for the
Individual Plant Examination of External Events (IPEEE) for Severe
Accident Vulnerabilities.'' NRC: Washington, D.C. June 1991.
Regulatory Guide 1.59, Revision2, ``Design Basis Flooding for
Nuclear Power Plants.'' NRC: Washington, D.C. August 1977.
U.S. Department of Energy (U.S.) (DOE). DOE-Standard-1020-2002,
``Natural Phenomena Hazards Design and Evaluation Criteria for
Department of Energy Facilities.'' DOE: Washington, D.C. 2002.
American National Standards Institute/American Nuclear Society
(ANSI/ANS). ANS-2.8, ``Determining Design Basis Flooding at Power
Reactor Sites.'' ANSI/ANS: July 1992.
Dated: October 28, 2005.
Robert C. Pierson,
Director, Division of Fuel Cycle Safety and Safeguards, NMSS.
APPENDIX--Example Natural Phenomena Hazard Review for Compliance with
10 CFR 70.61
This example review is for an amendment to authorize operations
in a blended low-enriched uranium oxide conversion building (OCB).
The site is located near a river and is just above the 100-year
flood plain of a nearby creek. The Effluent Process Building (EPB)
was also part of the amendment but was not evaluated because the
quantities of radioactive material or hazardous chemicals (that come
under NRC regulation) contained in the EPB are not considered
sufficient to exceed the 10 CFR 70.61 consequence threshold for
The OCB is of reinforced concrete construction and is
constructed to seismic criteria contained in the Standard Building
Code (SBC-1999) which is equivalent to being designed for an
earthquake having a probability of exceedance of approximately 4 X
10-4 per year. Using Appendix C of DOE-STD-1020-2002, a
risk reduction factor of 4 was determined by U.S. Nuclear Regulatory
Commission (NRC) staff, giving the structure a likelihood of
significant damage from an earthquake of 10-4 per year or
less. Hence, the collapse or loss of building integrity from an
earthquake may be considered to be ``highly unlikely'' as the
probabilistic value of ``highly unlikely'' indicated by the
applicant was a probability of exceedance of 10-4 to
10-5 per year. Within the building, the material at risk
consists of low enriched uranyl nitrate liquid, ammonium diuranate
slurry, and uranium dioxide powder. All of these materials are
expected to be within containers and spillage during a seismic event
is expected to be minimal. Since the building is expected to retain
its integrity, the leak path factor will be relatively low even
without dynamic confinement from the ventilation system. Facility
workers are expected to take actions to limit personal intake of
radionuclides. The staff concludes that the OCB complies with the
performance requirements of 10 CFR 70.61 with regard to seismic
High Winds Evaluation
The OCB structure is also designed for wind loads in accordance
with the SBC-1999, and the probability of a tornado impacting the
facility is less than 10-5 per year. Therefore, the
facility needs only to be evaluated in regard to the effects of wind
loads and missiles, but not for tornadoes. The reinforced concrete
exterior walls of the OCB are considered by NRC staff to be adequate
to withstand high wind velocities as well as missiles (from DOE-STD-
1020-2002) that should be assumed for such events. A collapse of
building walls due to wind forces such that radioactive material
would escape is considered to be ``highly unlikely'' by NRC staff.
In addition, the meteorological conditions likely to result in
severe winds may be forecast in advance, and protective measures
taken. The staff concludes that the OCB complies with the
performance requirements of 10 CFR 70.61 with regard to wind events.
The lowest floor in the OCB is 15 feet above the 100-year flood
plain from an adjacent creek. From a review of the topography of the
site area, it appears that flooding of the site could occur, most
likely, from flooding on the nearby river with coincident flooding
on the adjacent creek which could back up through the railroad
culvert. This event is expected to have warning time and may overtop
the railroad embankment to the north of the facility and flood parts
of the town nearby. However, the facility is sufficiently removed
from the main channel of the river such that flood-induced scouring
and erosion would not be expected. In addition, the hydrostatic
loading from the flood on the exterior walls of the OCB would not be
expected to cause collapse. The primary concern is inundation which
could float unsecured containers within the OCB but not remove them
from the facility. A criticality event can not be excluded, but
could occur only in the flooded and, therefore, evacuated section of
the plant and would not affect facility workers. In addition, the
warning time would allow the movement of material to reduce the
likelihood of a flood-induced criticality. The staff concludes that
the OCB complies with the performance requirements of 10 CFR 70.61
with regard to flooding.
[FR Doc. 05-22199 Filed 11-7-05; 8:45 am]
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