[Federal Register Volume 70, Number 215 (Tuesday, November 8, 2005)]
[Notices]
[Pages 67757-67761]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-22199]


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NUCLEAR REGULATORY COMMISSION


Notice of Availability of Interim Staff Guidance Documents for 
Fuel Cycle Facilities

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of availability.

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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].

SUPPLEMENTARY INFORMATION:

I. Introduction

    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.

II. Summary

    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].

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         Interim staff guidance                ADAMS  Accession No.
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FCSS Interim Staff Guidance-08, Revision  ML052650305
 0.
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    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 
Phenomena Hazards

Prepared by Division of Fuel Cycle Safety and Safeguards, Office of 
Nuclear Material Safety and Safeguards

Issue

    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.

Introduction

    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.

Discussion

    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 
(or

[[Page 67758]]

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 
administrative controls.
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    \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.
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    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.

Seismic Hazards

    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 
include:
    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 
event,

[[Page 67759]]

thus limiting any leakage of UF6 to the outside.

Flooding Hazards

    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 
individual basis.
    In evaluating the effects of flooding on existing facilities, the 
following flood-related hazards should be considered:
River Flooding
     Inundation and hydrostatic loading
     Dynamic forces
     Wave action
     Sedimentation and erosion
     Ice loading
Upstream Dam Failures
     Inundation and hydrostatic loading
     Dynamic forces
     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
     Dynamic forces
     Wave action
    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 
``highly unlikely.''
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    \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 
unlikely event.
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    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

[[Page 67760]]

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 
facility structure.
    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 
facilities.

Regulatory Basis

    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 
parameters.
    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.

Recommendation

    This guidance should be used to supplement NUREG-1520, Chapter 3, 
Integrated Safety Analysis.

References

    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.

[[Page 67761]]

    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 
``unlikely'' events.

Seismic Evaluation

    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 
events.

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.

Flooding Evaluation

    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]
BILLING CODE 7590-01-P