80 FR 33299 - Independent Spent Fuel Storage Installation, Department of Energy; Fort St. Vrain

NUCLEAR REGULATORY COMMISSION

Federal Register Volume 80, Issue 112 (June 11, 2015)

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.

[Federal Register Volume 80, Number 112 (Thursday, June 11, 2015)]
[Notices]
[Pages 33299-33303]
From the Federal Register Online  [www.thefederalregister.org]
[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]
 BILLING CODE 7590-01-P