[Federal Register Volume 80, Number 112 (Thursday, June 11, 2015)]
[Notices]
[Pages 33299-33303]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-14291]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 72-09; NRC-2015-0150]
Independent Spent Fuel Storage Installation, Department of
Energy; Fort St. Vrain
AGENCY: Nuclear Regulatory Commission.
ACTION: Exemption; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing an
exemption in response to a March 19, 2015 request, as supplemented
April 3, and June 1, 2015, from the Department of Energy (DOE or the
licensee). The exemption seeks to delay the performance of an O-ring
leakage rate test specified in Technical Specification (TS) 3.3.1 of
Appendix A of Special Nuclear Material License No. SNM-2504, and to
delay the performance of an aging management surveillance described in
the Fort St. Vrain (FSV) Final Safety Analysis Report (FSAR) to check
six Fuel Storage Containers (FSCs) for hydrogen buildup, both until
June, 2016.
DATES: Notice of issuance of exemption given on June 11, 2015.
ADDRESSES: Please refer to Docket ID NRC-2015-0150 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly-available information related to this document
using any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0150. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. For
the convenience of the reader, the ADAMS accession numbers are provided
in a table in the ``Availability of Documents'' section of this
document. Some documents referenced are located in the NRC's ADAMS
Legacy Library. To obtain these documents, contact the NRC's PDR for
assistance.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Chris Allen, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone: 301-415-6877; email:
[email protected].
I. Background
DOE is the holder of Special Nuclear Material License No. SNM-2504
which authorizes receipt, possession, storage, transfer, and use of
irradiated fuel elements from the decommissioned FSV Nuclear Generating
Station in Platteville, Colorado, under part 72 of title 10 of the Code
of Federal Regulations (10 CFR).
II. Request/Action
According to TS 3.1.1 in Appendix A of License No. SNM-2504, the
FSC seal leakage rate shall not exceed 1 x 10\3\ reference cubic
centimeters per second (ref-cm\3\/s). Surveillance Requirement (SR)
3.3.1.1 calls for one FSC from each vault to be leakage rate tested
every five years. The last leakage rate test was performed in June,
2010; the next leakage rate test is scheduled to be completed by June,
2015. In addition, as part of the aging management program implemented
when the license was renewed in 2011, Chapter 9 of the FSV FSAR
provides the licensee will check six FSCs for hydrogen buildup by June,
2015. This provision regards the potential for hydrogen generation. The
date of sampling was chosen to be consistent with the FSC seal leakage
rate testing schedule. No FSCs have been sampled for hydrogen since
being
[[Page 33300]]
placed into storage. DOE requests an exemption to delay performance of
both the FSC O-ring leakage rate test requirement and the FSAR aging
management activity described above by one year.
III. Discussion
Under 10 CFR 72.7, the Commission may, upon application by any
interested person or upon its own initiative, grant exemptions from the
requirements of 10 CFR part 72 when the exemption is authorized by law,
will not endanger life or property or the common defense and security,
and is otherwise in the public interest. In addition to the requirement
from which DOE requested exemption, the NRC staff determined that an
exemption from 10 CFR 72.44(c)(1) would also be necessary to implement
DOE's exemption proposal. Section 72.44(c)(1) requires, in part,
compliance with functional and operational limits to protect the
integrity of waste containers and to guard against the uncontrolled
release of radioactive material.
Authorized by Law
This exemption would delay performance of an FSC O-ring leakage
rate test required by TS 3.3.1 of Appendix A of Special Nuclear
Material License No. SNM-2504, and an FSAR aging management
surveillance to check six FSCs for hydrogen buildup by June, 2015 by
one year. Condition 9 of SNM-2504 states, in part, that authorized use
of the material at the FSV ISFSI shall be ``in accordance with
statements, representations, and the conditions of the Technical
Specifications and Safety Analysis Report.'' Condition 11 of SNM-2504
also directs the licensee to operate the facility in accordance with
the Technical Specifications in Appendix A.
The provisions in 10 CFR part 72 from which DOE requests an
exemption, as well as the provisions considered by the NRC staff,
require the licensee to follow the technical specifications and the
functional and operational limits for the facility. Section 72.7 allows
the NRC to grant exemptions from the requirements of 10 CFR part 72.
Issuance of this exemption is consistent with the Atomic Energy Act of
1954, as amended, and not otherwise inconsistent with NRC regulations
or other applicable laws. Therefore, the exemption is authorized by
law.
Will Not Endanger Life or Property or the Common Defense and Security
As discussed below, the NRC staff has evaluated the proposed
exemption request, and found that it would not endanger life or
property, or the common defense and security.
Potential Corrosion
The FSV ISFSI Aging Management Program described in Section 9.8 of
the FSV ISFSI FSAR provides for sampling one FSC in each vault for
hydrogen no later than June, 2015. The intent of the test was to
identify any potential corrosion on the interior of the FSCs. The
applicant stated its position as to why hydrogen buildup has not
occurred, and thus why there are no safety implications with delaying
the test for one year, including:
1. The fuel was stored in dry helium prior to placement in the
FSCs;
2. General corrosion, as opposed to galvanic corrosion, was
determined by the licensee to be the only corrosion mechanism of
concern for the canister; and
3. The expected corrosion reactions would not generate significant
quantities of hydrogen since the pH of any water inside the FSCs was
expected to be neutral (i.e., not acidic).
In addition to reviewing information in the exemption request, the
NRC staff also reviewed information associated with the 2011 license
renewal applicable to this request. From its review of the license
renewal documents, the NRC staff identified the following information
pertinent to its review of DOE's exemption request: Corrosion
originating on the FSC interior surfaces was evaluated in Engineering
Design File 9166 (EDF-9166) (ADAMS Accession No. ML15132A638). The EDF
9166 assumed that 775.6 grams of water was present in each FSC. The
analysis assumed uniform corrosion of all interior FSC surfaces
resulting in a loss of material of 0.0014 inches. Crevice and galvanic
corrosion were also assumed for the FSC bottom plate resulting in a
loss of thickness of 0.0576 inches. In both cases, the licensee's
analyses determined that the remaining material thicknesses for all
interior FSC surfaces were greater than the required minimum thickness
for the FSCs to maintain confinement of the radioactive material.
As referenced in the application, a surface coating had been
applied to the interior FSC surfaces, but the NRC staff also found that
the licensee's statement that general corrosion, and not galvanic
corrosion, was the only corrosion mechanism of concern for the FSCs is
not consistent with information in the FSV FSAR. For instance, Chapter
4, section 4.2.3.2.3 of the FSV ISFSI FSAR considered the potential for
galvanic corrosion with the carbon steel FSC acting as the anode and
the graphite fuel acting as the cathode. In addition, the NRC staff
determined that EDF-9166 may not have fully considered all possible
reactions. For instance, EDF-9166 only considered galvanic corrosion
between the fuel blocks and the bottom of the FSC, and it assumed
material loss from corrosion was distributed over the entire internal
surface area of the FSC. The NRC staff notes that small portions of
carbon steel, resulting either from coating defects during the surface
coating application or from nicks and scratches during fabrication or
loading, could act as localized sites of galvanic corrosion when
exposed to water in the FSC. Therefore, the NRC staff finds that the
applicant may have incorrectly assumed that corrosion is uniformly
distributed to all FSC interior surfaces instead of being localized
where protective coating is not present. Nevertheless, the NRC staff
finds that through wall corrosion remains unlikely even if localized
corrosion occurs at areas of coating defects or damage because the
amount of water present is limited, because water is a low conductivity
electrolyte, and the voluminous iron hydroxide formed by the corrosion
reactions would stifle the corrosion process prior to significant
localized loss of thickness of the FSC.
The corrosion processes discussed above would generate hydrogen as
a result of reduction reactions on the graphite surfaces. For these
reduction reactions to occur, a liquid medium must be present in the
FSCs. Information contained in the application indicated that, if the
temperature of the graphite fuel blocks exceeded 200 [deg]F [93 [deg]C]
due to off-normal or accident conditions, any water in the graphite
fuel blocks could be forced out of the fuel blocks resulting in as much
as 77.6 grams of water being inside an FSC. The EDF-9166, which stated
that it increased this amount of water by a factor 10, contains a
corrosion analysis that identified oxygen reduction as the most likely
reduction reaction in the system. This reduction reaction does not
generate hydrogen. Although the possible generation of hydrogen as a
result of other reactions is described in EDF-9166, the applicant did
not evaluate the amount of hydrogen that may be produced.
In addition to reviewing information submitted by DOE in the
exemption request, the NRC staff identified several possible reactions
to assess the potential for hydrogen generation from corrosion
reactions. These include the corrosion of iron, the formation of iron
corrosion products, the oxidation of iron corrosion products, and the
reduction reactions
[[Page 33301]]
for oxygen, water and hydrogen ions. These reactions are listed below.
[GRAPHIC] [TIFF OMITTED] TN11JN15.003
The reduction of hydrogen ions (Eq. 6) occurs primarily in acidic
solutions. The reduction of oxygen (Eq. 2) is the likely reduction
reaction in a system with air. Since the environment inside the FSCs is
air, the reduction of oxygen (Eq. 2) is applicable. If the oxygen in
the air is completely consumed, then the corrosion reaction can proceed
until water is consumed via the water reduction reaction (Eq. 5).
Using the equations above, the NRC staff performed the following
analysis assuming the complete consumption of oxygen and water in
corrosion product formation and reduction reactions. It is uncertain if
the complete consumption of the reactants is a reasonable assumption
due to the use of the surface coating. Therefore, it is unknown how
much of the carbon steel is available for corrosion product formation
and the reduction reactions. Thus, assuming complete consumption of
oxygen and water provides a conservative estimate of the amount of
hydrogen that may be formed.
The free volume inside an FSC is estimated to be 230 liters. At 200
[deg]F [93 [deg]C], the temperature at which water, if present, could
be released from the graphite, a mole of air, the gas inside an FSC,
occupies 30 liters. Since air contains 21 percent oxygen by volume, the
free volume of the FSC may be expected to contain 1.61 moles of
O2 and 6.05 moles of N2. There are 4.3 moles of
water in 77.6 grams of water. The reduction of oxygen (Eq. 2) requires
2 moles of water for each mole of oxygen. Reduction of 1.61 moles of
O2 requires 3.22 moles of H2O leaving 1.08 moles
of water unreacted. If the remaining 1.08 moles of water is reduced
(Eq. 5), then 0.54 moles of hydrogen would be produced. The volume
occupied by 0.54 moles of H2 at 200 [deg]F [93 [deg]C] is
16.2 liters. This results in a volume fraction of 16.2/230 = 0.07 or 7
percent H2.
Although the analysis above does not consider either the formation
of water as a result of decomposition of the surface coating on the
interior surfaces of the FSC or hydrogen formation from the small
amount of grease used on the metallic O-rings, it shows that, if all of
the water present is released from the graphite and subsequently
consumed in corrosion reactions, there is a possibility of generating a
significant amount of hydrogen. It also shows that, if the amount of
water assumed by DOE in EDF-9166 were present in the FSCs, the amount
of hydrogen would be even greater.
However, the NRC staff notes the following facts relative to the
possibility of either an explosive or combustible mixture of gases
inside an FSC at the FSV ISFSI. Based upon the above reactions, oxygen,
which is a necessary ingredient in explosive and combustible gas
mixtures, would not be present within the FSC interior free volume
because the reduction reactions would have completely consumed it.
There are no credible sources of ignition during normal fuel storage
operations for the following reasons. First, sparks caused by metal to
metal interaction are not produced because the FSCs are stationary.
Second, Chapter 3 of the FSV FSAR identified the maximum FSC gas
temperature as approximately 165 [deg]F (74 [deg]C). The NRC staff
notes that this gas temperature is far below the estimated minimum
auto-ignition temperature of hydrogen gas in air of 752 [deg]F (400
[deg]C). Since the maximum temperature in Chapter 3 of the FSV FSAR was
used in support of the license renewal, the NRC staff further notes
that the maximum
[[Page 33302]]
temperature inside the FSC is now even lower considering the fuel has
been in storage for 24 years. Finally, Chapter 4 of the FSV FSAR states
the licensee will, prior to either handling of a loaded FSC or removal
of the lid bolts, implement the following procedural controls:
1. Analyze the gas environment in the FSCs;
2. Determine if flammable levels of hydrogen are present; and
3. As necessary, either evacuate or purge the FSC with air to
assure hydrogen concentrations are below flammable levels.
Therefore, NRC staff concludes that a fire or explosion due to the
presence of hydrogen is very unlikely, and does not present a
significant safety issue if the exemption request is granted.
Consequently, delaying the analysis of the gases inside the FSC from 24
to 25 years would not result in an increase in the probability of
either a hydrogen ignition event during storage or failure of the FSC
integrity due to corrosion. The NRC staff also finds that, as long as
operational controls that eliminate ignition sources and requirements
for gas sampling prior to handling or removal of lid bolts are
maintained and followed, hydrogen ignition events associated with
handling FSCs will not occur.
Leakage Rate
Limiting Condition of Operation 3.3.1 in Appendix A of License No.
SNM-2504 states that the FSCs seal leakage rate shall not exceed 1 x
10[middot]3 ref-cm\3\/s. SR 3.3.1.1 calls for one FSC from
each vault to be leakage rate tested every 5 years. The basis for SR
3.3.1.1 is that performance of a leakage rate test of at least six FSC
closures every 5 years provides reasonable assurance of continued
integrity. The leakage rate test was originally performed in 1991 after
loading and subsequent leakage rate tests were performed in 1996, 2001,
2005, and 2010. None of the prior leakage rate tests exceeded the
requirement of 1 x 10[middot]3 ref-cm\3\/s.
DOE evaluated the potential impact of this exemption request in
accordance with the confinement requirements described in the FSV FSAR.
DOE classified the failure of the FSC redundant metal O-ring seals as a
low probability event, and stated Chapter 8, section 8.2.15 of the FSV
FSAR identified no credible failure mechanisms for the FSC O-rings. DOE
also estimated average and maximum O-ring seal leakage rates would be
3.75 x 10-4 and 6.76 x 10-4 ref-cm\3\/s,
respectively and documented these calculations in EDF-10727
(ML15104A064). Both seal leakage rate values are below the allowed
leakage rate of 1 x 10-3 ref-cm\3\/s required by TS 3.3.1.
DOE identified O-ring failure as a potential failure mode that would
allow leakage in excess 1 x 10[middot]3 ref-c cm\3\/s;
however, DOE provided no specific details of potential O-ring failure
mechanisms.
The NRC staff notes typical failure modes for O-ring seals include:
1. Corrosion of the O-ring;
2. Corrosion of the O-ring flange sealing surface (area in contact
with the O-ring); and
3. Creep or relaxation of the O-ring.
The O-rings are described in DOE's exemption request, as
supplemented on June 1, 2015 (ADAMS Accession No. ML15153A280), as
silver plated alloy X-750 in the work hardened condition. The O-rings
are installed with a grease/lubricant to facilitate sealing and prevent
damage to the O-rings during lid installation and compression of the O-
rings. The presence of the grease, the materials of construction, and
the limited amount of water in the vicinity of the O-rings reduce the
possibility of corrosion of the O-rings and the O-ring seal area on the
FSC.
The NRC staff reviewed the test methods, the test pressures
generated by previous leakage rate tests, and the correlations between
the leakage rate and the pressure drop across the seals used in EDF-
10727 to estimate the O-ring seal leakage rates. The NRC staff finds
that DOE used appropriate data and mechanistic relationships between
the rate and the test pressure to predict June, 2017 FSC O-ring seal
leakage rates. The staff determined that both the average and maximum
estimated 2017 leakage rates of 3.75 x 10-4 and 6.76 x
10-4 ref-cm\3\/s are acceptable and are below the required
limit of 1 x 10-3 ref-cm\3\/s.
The NRC staff also reviewed both Chapter 8, section 8.2.15 of the
FSV FSAR and DOE's analytical results of the consequences associated
with a radiological release from an FSC, and confirmed that even if the
leakage rate of 1 x 10-3 ref-cm\3\/s is grossly exceeded:
1. The radiological consequences at the controlled area boundary
would be within the requirements of 10 CFR 72.106;
2. The radiological release caused by a leakage rate greater than 1
x 10-3 ref-cm\3\/s past the redundant seals would be bounded
by the maximum credible accident in the FSV FSAR; and
3. The failure of the redundant metallic seals (loss of
confinement) can be considered a low probability event during the
entire storage period.
Based on the findings above, NRC staff concludes that granting
DOE's exemption to delay performance of the FSC O-ring leakage rate
test in accordance with TS 3.1.1 and performance of the aging
management surveillance to sample six FSCs for hydrogen until June
2016, would not endanger public health and safety or the common defense
and security.
Otherwise in the Public Interest
As described in the application, delaying the FSC O-ring leakage
rate test and FSAR aging management surveillance for one year would
allow DOE to more effectively prioritize important activities at the
FSV site. It would also reduce the administrative burden both on the
licensee and on the NRC staff in the performance of the test.
Therefore, issuance of the proposed exemption is otherwise in the
public interest.
Environmental Consideration
The NRC staff evaluated whether there would be any significant
environmental impacts associated with the issuance of the requested
exemption. The NRC staff determined that this proposed action fits a
category of actions which do not require an environmental assessment or
environmental impact statement. Specifically, the exemption meets the
categorical exclusion in 10 CFR 51.22(c)(25).
Granting an exemption from the requirements of 10 CFR 72.44(c)(1),
and 10 CFR 72.44(c)(3) involves inspection and surveillance
requirements associated with both the FSC O-ring leakage rate test
required per TS 3.3.1 and the FSAR aging management surveillance of
FSCs for hydrogen. A categorical exclusion for inspection and
surveillance requirements is provided under 10 CFR 51.22(c)(25)(vi)(C)
if the criteria in 10 CFR 51.22(c)(25)(i) through (v) are also
satisfied. In its review of the exemption request, the NRC staff
determined that, under 10 CFR 51.22(c)(25): (i) Granting the exemption
does not involve a significant hazards considerations, because granting
the exemption neither reduces a margin of safety, creates a new or
different kind of accident from any accident previously evaluated, nor
significantly increases either the probability or consequences of an
accident previously evaluated; (ii) granting the exemption would not
produce a significant change in either the types or amounts of any
effluents that may be released offsite, because the requested exemption
neither changes the effluents nor produces additional
[[Page 33303]]
avenues of effluent release; (iii) granting the exemption would not
result in a significant increase in either occupational radiation
exposure or public radiation exposure, because the requested exemption
neither introduces new radiological hazards nor increases existing
radiological hazards; (iv) granting the exemption would not result in a
significant construction impact, because there are no construction
activities associated with the requested exemption; and; (v) granting
the exemption would not increase either the potential or consequences
from radiological accidents such as a gross leak from an FSC, or the
potential for hydrogen buildup or consequences from radiological
accidents, because the exemption neither reduces the ability of the FSC
to confine radioactive material nor creates new accident precursors at
the FSV ISFSI. Accordingly, this exemption meets the criteria for a
categorical exclusion in 10 CFR 51.22(c)(25)(vi)(C).
IV. Conclusions
Accordingly, the NRC has determined that, under 10 CFR 72.7, this
exemption is authorized by law, will not endanger life or property or
the common defense and security, and is otherwise in the public
interest. Therefore, the Commission hereby grants DOE an exemption from
10 CFR 72.44(c)(1) and 10 CFR 72.44(c)(3) to delay by one year the
scheduled June, 2015 leakage rate test under SR 3.3.1.1 for one FSC
from each vault to be leakage rate tested every five years, and to
delay by one year the scheduled June, 2015 hydrogen buildup test
described in Chapter 9 of the FSV FSAR. These tests shall be completed
no later than June, 2016. This exemption is effective as of June 4,
2015.
Dated at Rockville, Maryland, this 4th day of June, 2015.
For the Commission.
Mark Lombard,
Director, Division of Spent Fuel Management, Office of Nuclear Material
Safety and Safeguards.
[FR Doc. 2015-14291 Filed 6-10-15; 8:45 am]
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