List of Approved Spent Fuel Storage Casks: Holtec International HI-STORM 100 Cask System; Amendment No. 9, Revision 1, 371-378 [2015-33280]

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• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 3 (Wednesday, January 6, 2016)]
[Rules and Regulations]
[Pages 371-378]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-33280]



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Federal Register / Vol. 81, No. 3 / Wednesday, January 6, 2016 / 
Rules and Regulations

[[Page 371]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

[NRC-2015-0156]
RIN 3150-AJ63


List of Approved Spent Fuel Storage Casks: Holtec International 
HI-STORM 100 Cask System; Amendment No. 9, Revision 1

AGENCY: Nuclear Regulatory Commission.

ACTION: Direct final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its 
spent fuel storage regulations by revising the Holtec International 
(``Holtec,'' or ``the applicant'') HI-STORM 100 Cask System listing 
within the ``List of Approved Spent Fuel Storage Casks'' to include 
Amendment No. 9, Revision 1, to Certificate of Compliance (CoC) No. 
1014. Amendment No. 9, Revision 1, changes cooling time limits for 
thimble plug devices (TPDs), removes certain testing requirements for 
the fabrication of Metamic HT neutron-absorbing structural material, 
and reduces certain minimum guaranteed values (MGV) used in bounding 
calculations for this material. Amendment No. 9, Revision 1, also 
changes fuel definitions to classify certain boiling water reactor 
(BWR) fuel within specified guidelines as undamaged fuel.

DATES: The direct final rule is effective March 21, 2016, unless 
significant adverse comments are received by February 5, 2016. If the 
direct final rule is withdrawn as a result of such comments, timely 
notice of the withdrawal will be published in the Federal Register. 
Comments received after this date will be considered if it is practical 
to do so, but the NRC staff is able to ensure consideration only of 
comments received on or before this date. Comments received on this 
direct final rule will also be considered to be comments on a companion 
proposed rule published in the Proposed Rules section of this issue of 
the Federal Register.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0156. Address 
questions about NRC dockets to Carol Gallagher, telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Email comments to: Rulemaking.Comments@nrc.gov. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal 
workdays; telephone: 301-415-1677.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Robert D. MacDougall, Office of 
Nuclear Material Safety and Safeguards, telephone: 301-415-5175, email: 
Robert.MacDougall@nrc.gov; U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION: 

Table of Contents

I. Obtaining Information and Submitting Comments
II. Procedural Background
III. Background
IV. Discussion of Changes
V. Voluntary Consensus Standards
VI. Agreement State Compatibility
VII. Plain Writing
VIII. Environmental Assessment and Final Finding of No Significant 
Environmental Impact
IX. Paperwork Reduction Act Statement
X. Regulatory Flexibility Certification
XI. Regulatory Analysis
XII. Backfitting and Issue Finality
XIII. Congressional Review Act
XIV. Availability of Document

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0156 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0156.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://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 pdr.resource@nrc.gov. For 
the convenience of the reader, instructions about obtaining materials 
referenced in this document are provided in the ``Availability of 
Documents'' section.
     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.

B. Submitting Comments

    Please include Docket ID NRC-2015-0156 in the subject line of your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.

[[Page 372]]

    If you are requesting or aggregating comments from other persons 
for submission to the NRC, you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Procedural Background

    This rule is limited to the changes contained in Amendment No. 9, 
Revision 1, to CoC No. 1014 and does not include other aspects of the 
Holtec HI-STORM 100 Cask System design. The NRC is using the ``direct 
final rule'' procedure to issue this amendment because it represents a 
limited and routine change to an existing CoC and is expected to be 
noncontroversial. Adequate protection of public health and safety 
continues to be ensured. The amendment to the rule will become 
effective on March 21, 2016. If the NRC receives significant adverse 
comments on this direct final rule by February 5, 2016, the NRC will 
publish a Federal Register notice withdrawing the direct final rule, 
and will address the comments in a subsequent Federal Register notice 
for a final rule based on the companion proposed rule published in the 
Proposed Rule section of this issue of the Federal Register. Absent the 
need for significant modifications to the proposed revisions that would 
require republication, the NRC will not initiate a second comment 
period on this action.
    A significant adverse comment is a comment where the commenter 
explains why the rule would be inappropriate, including challenges to 
the rule's underlying premise or approach, or would be ineffective or 
unacceptable without a change. A comment is adverse and significant if:
    (1) The comment opposes the rule and provides a reason sufficient 
to require a substantive response in a notice-and-comment process. For 
example, a substantive response is required when:
    (a) The comment causes the NRC staff to reevaluate (or reconsider) 
its position or conduct additional analysis;
    (b) The comment raises an issue serious enough to warrant a 
substantive response to clarify or complete the record; or
    (c) The comment raises a relevant issue that was not previously 
addressed or considered by the NRC staff.
    (2) The comment proposes a change or an addition to the rule, and 
it is apparent that the rule would be ineffective or unacceptable 
without incorporation of the change or addition.
    (3) The comment causes the NRC staff to make a change (other than 
editorial) to the rule, CoC, or Technical Specifications.
    For detailed instructions on filing comments, please see the 
ADDRESSES section of this document.

III. Background

    Section 218(a) of the Nuclear Waste Policy Act (NWPA) of 1982, as 
amended, requires that ``the Secretary [of the U.S. Department of 
Energy] shall establish a demonstration program, in cooperation with 
the private sector, for the dry storage of spent nuclear fuel at 
civilian nuclear power reactor sites, with the objective of 
establishing one or more technologies that the [U.S. Nuclear 
Regulatory] Commission may, by rule, approve for use at the sites of 
civilian nuclear power reactors without, to the maximum extent 
practicable, the need for additional site-specific approvals by the 
Commission.'' Section 133 of the NWPA states, in part, that ``[t]he 
Commission shall, by rule, establish procedures for the licensing of 
any technology approved by the Commission under Section 219(a) [sic: 
218(a)] for use at the site of any civilian nuclear power reactor.''
    To implement this mandate, the Commission approved dry storage of 
spent nuclear fuel in NRC-approved casks under a general license by 
publishing a final rule to add a new subpart K in part 72 of title 10 
of the Code of Federal Regulations (10 CFR) entitled, ``General License 
for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 
18, 1990). This rule also established a new subpart L in 10 CFR part 72 
entitled, ``Approval of Spent Fuel Storage Casks,'' which contains 
procedures and criteria for obtaining NRC approval of spent fuel 
storage cask designs. The NRC subsequently issued a final rule (65 FR 
25241; May 1, 2000) that approved the HI-STORM 100 Cask System design 
and added it to the list of NRC-approved cask designs in 10 CFR 72.214, 
``List of approved spent fuel storage casks,'' as CoC No. 1014. Most 
recently, the NRC issued a final rule effective on March 11, 2014 (78 
FR 78165), that approved the HI-STORM 100 Cask System design amendment 
subject to this rulemaking and added it to the list of NRC-approved 
cask designs in 10 CFR 72.214 as CoC No. 1014, Amendment No. 9.

IV. Discussion of Changes

    On July 1, 2014, Holtec submitted a request to the NRC to revise 
CoC No. 1014 to supersede Amendment 9 with Amendment 9, Revision 1. 
Amendment No. 9, Revision 1, changes cooling time limits for TPDs, 
removes certain testing requirements for the fabrication of Metamic HT, 
and reduces certain MGVs used in bounding calculations for this 
material. Amendment No. 9, Revision 1, also changes fuel definitions to 
classify certain boiling water reactor (BWR) fuel within specified 
guidelines as undamaged fuel. The changes to the CoC and Technical 
Specifications (TS) Appendices are identified with revisions bars in 
the margin of each document.
    As a revision, the CoC and its associated TS will supersede the 
previous version of the CoC No. 1014, Amendment No. 9 CoC and its TSs 
in their entirety. A revision in lieu of a new amendment is justified 
on the grounds that:
     Equipment for CoC No. 1014, Amendment No. 9, cask systems 
has been placed in service by several general licensees, all of whom 
were made aware of Holtec's revision request and supported it;
     No new canisters are being requested to be added to CoC 
No. 1014, Amendment No. 9, cask systems;
     No new systems, components, or structures are requested to 
be added to CoC No. 1014, Amendment No. 9, cask systems;
     The requested changes have minor field and administrative 
implementation impacts on general licensees; and
     The requested changes are applicable to CoC No. 1014, 
Amendment No. 9, in their entirety.
    Each of the applicant's proposed changes is discussed below.

1. Reduced Cooling Time Limit for TPDS

    The TPDs are a form of non-fuel hardware inserted into guide tubes 
used in some pressurized water reactor (PWR) fuel assemblies and made 
radioactive by exposure to neutrons during reactor operation. 
Supporting its proposal to reduce the cooling time limits for TPDs, the 
applicant noted that TPDs are not considered in any of the thermal 
analyses of CoC No. 1014, Amendment No. 9, so that in order to comply 
with this amendment, general licensees must perform an evaluation under 
10 CFR 72.212 to ensure that maximum fuel storage decay heat limits are 
met. The applicant stated that, currently, cooling times for TPDs 
exposed to typical fuel burnups in a reactor core are long, preventing 
many TPDs from being stored in the dry multi-purpose canisters (MPC) 
that contain spent fuel and non-fuel hardware with ``activation 
products,'' or components or constituents made radioactive by

[[Page 373]]

exposure to neutrons in the reactor core. The applicant proposed to 
reduce the required cooling times so that general license users can 
have greater flexibility to store a larger population of TPDs.
    The principal activation product from the irradiation of TPDs in a 
reactor core is Cobalt-60 (Co-60), which has a half-life (the time it 
takes to lose half its radioactivity) of 5 years. The applicant 
calculated that the Co-60 source for a TPD with a five-year cooling 
time after exposure to a fuel burnup of 63,000 megawatt-days per metric 
ton of uranium (MWD/MTU) or less is 141 curies. The maximum Co-60 
activity of TPDs is 240 curies. The applicant selected 141 curies Co-60 
as the design basis Co-60 activity for each TPD, so that any TPD can be 
stored in a HI-STORM MPC so long as the TPD has a cooling time of 5 
years or greater after a burnup of 63,000 MWD/MTU or less, as required 
by the TSs.
    The applicant also calculated the dose rates from a HI-STORM 100 
overpack with an MPC for BWR and for PWR fuels using allowable burnup 
and cooling times from the proposed Revision 1 to CoC No. 1014, 
Amendment No. 9. These calculated dose rates were less than the 
allowable values in the TSs for the currently-approved Amendment No. 9.
    The NRC staff reviewed the applicant's proposed revisions to its 
final safety analysis report (FSAR) and finds that the proposed change 
would have no impact on a fuel rod's internal pressure or cladding 
temperatures. The NRC staff finds the storing of TPDs to be acceptable 
because, as non-fuel components, they present no risk of rupturing and 
releasing fission products, fission product gases, or any other 
material detrimental to the internals of the cask. Nor would the 
storage of TPDs prevent the retrieval of spent fuel from a cask. 
General licensees will, however, continue to be required under 72.212 
to evaluate and ensure that cell heat loads per canister remain below 
the applicable limits as listed in the FSAR and TSs prior to loading.

2. Removing or Revising Certain Metamic-HT Fabrication Testing 
Requirements

    Metamic-HT is a neutron-absorbing structural material used for 
internal components of MPCs, which hold spent fuel assemblies and other 
radioactive fuel components inside storage casks. The applicant 
proposed changing Metamic-HT fabrication testing requirements to: 
Remove testing using a 1-inch collimated neutron beam; remove Charpy V-
notch and lateral expansion testing; remove thermal conductivity 
testing; revise testing requirements for fuel basket welds; change re-
testing criteria when a component fails to meet an MGV by requiring 
only the failed property to be re-tested (not all MGVs); and add the 
ability to conduct 100% testing of an MGV property within a lot if a 
sample within the lot fails re-testing. According to the applicant, 
these changes are to improve Metamic-HT testing, or ease undue burden, 
because some testing requirements were overly conservative and created 
a lengthy testing process, while others did not affect the safety 
analysis.
    The requirement for the use of a 1-inch neutron beam is based on 
Interim Staff Guidance (ISG)-23, ``Application of ASTM Standard 
Practice C1671-07 when performing technical reviews of spent fuel 
storage and transportation packaging licensing actions.'' ISG-23 
concludes that a beam between 1 cm and 2.54 cm is acceptable for 
qualification and acceptance testing of neutron absorbing materials. 
The ISG also states, however, that ``a visual inspection should be 
conducted on all neutron absorbing materials intended for service,'' 
and that as part of that visual inspection, ``it is important to ensure 
that there are no defects that might lead to problems in service; such 
as delaminations or cracks that could appear on clad neutron absorbing 
materials.'' The staff finds that in this instance, a visual inspection 
of all neutron-absorbing materials intended for service, along with 
other fabrication testing measures called for in ISG-23, such as 
minimum plate thickness testing, will provide adequate assurance 
against significant defects in Metamic HT without the need for neutron 
beam testing.
    The Charpy V-notch test is a measure of a given material's 
toughness under impact loading to study temperature-dependent ductile-
to-brittle transitions. As temperature decreases, a metal's ability to 
absorb the energy of an impact--its ductility--decreases, and at some 
temperature, its ductility may suddenly drop almost to zero. This sharp 
transition to brittleness is essentially unidentified in metals with a 
face-centered cubic (FCC) crystal structure, however, and Metamic-HT is 
an aluminum composite with an FCC-based metal matrix. The staff 
therefore concludes that the Charpy V-notch test is not necessary for 
Metamic-HT.
    Proposing to remove the thermal conductivity testing requirement 
for Metamic-HT during fabrication, the applicant noted that there is 
little variability in this material's thermal conductivity when 
fabricated according to the manufacturing manual.
    The NRC staff evaluated the applicant's proposal and finds that the 
thermal conductivity of Metamic-HT is stable for normal operating 
temperatures (200 [deg]C to 500 [deg]C), so that removal of this 
testing requirement would have no impact on any of the previously 
approved NRC staff evaluations. The proposed change is therefore 
considered acceptable.
    The applicant also intends to employ a new qualified welding 
process called Friction Stir Welding (FSW), for external basket joints. 
Allowing the use of FSW of the Metamic HT basket does not change the 
safety basis as evaluated by the staff in HI-STORM 100, Amendment No. 
9, with respect to basket structural performance. Since the basket 
corners utilize the same welded joint configuration specified in 
amendment No. 9 and prior amendments, the primary consideration is that 
of weld process and qualification, rather than structural performance 
of the weld itself.
    Based on its review of the application, the staff determined that 
the methods used to qualify the weld joint were sufficiently robust to 
demonstrate a structural performance comparable to the welding method 
described in previous amendments. The loading conditions and the fully 
supported boundary conditions of the peripheral basket panels result in 
calculated joint stresses below their full capacity. The staff 
therefore concludes that this margin accounts for any differences in 
welding procedures, should they arise in the future. The staff's 
conclusions in this regard only apply to the basket corner welds and 
shim arrangement defined by this revision.

3. Changing Minimum Guaranteed Values for Metamic-HT Analyses

    Using the guidance of the American Society of Mechanical Engineers 
(ASME) Section II, Mandatory Appendix 5, ``Guideline on the Approval of 
New Materials Under the ASME Boiler and Pressure Vessel Code,'' Holtec 
determined the mechanical properties of Metamic-HT at ambient and 
various other higher and lower temperatures. It then analyzed its test 
data using statistical methods to determine minimum, average, and mean 
values of the material's structural properties. In addition, the 
applicant established a design value MGV for each of the various 
properties. An MGV is an arbitrary value for any given property below 
the lowest measured value from the test data. The MGV is then 
demonstrated or guaranteed to be exceeded for every manufactured lot of 
Metamic HT through lot testing.

[[Page 374]]

    The MGVs for Metamic-HT are used in calculations to demonstrate 
that structural components made with this material will satisfy 
engineering requirements, such as stress or deflection limits to ensure 
acceptable hardness of the component in service. Using MGV values 
produces a bounding calculation for any given engineering requirement.
    To support its proposal for reducing some of these MGVs, Holtec 
used differing MGV values in structural calculations for developing 
stress/strain curves from finite element analysis, a method of 
computing displacements, stresses, and strains at defined points along 
the length, width, or within a cross-section of a given component.
    Holtec's calculations determined that a positive margin of safety 
for basket performance criteria remains even with an average reduction 
of approximately 10 percent in MGVs for material yield stress, ultimate 
strength, and Young's modulus, a measure of a material's elasticity 
(ability to resume its original dimensions) under lengthwise tension or 
compression. The applicant also reported a calculated reduction of 20 
percent of the MGV for area criteria measured during a tensile test. 
Positive margins remain in the criteria for peak stress, maximum 
deflection, and crack propagation. These minimum values are guaranteed 
to be met by the imposition of a sampling test plan based on the 
standards for critical service parts. The applicant also proposed to 
add the ability to conduct 100 percent testing of an MGV property 
within a lot if a sample fails re-testing.
    This is the same change Holtec made to the HI-STORM 100 Flood/Wind 
(FW) Multipurpose MPC Storage System, CoC--No. 1032 using an acceptable 
evaluation that complied with 10 CFR 72.48, ``Changes, tests, and 
experiments.'' The NRC staff reviewed these results and finds the 
proposed changes acceptable, because an adequate safety margin remains 
for basket performance criteria even with the reduced MGVs.

4. New Spent Fuel Definitions

    Holtec proposed to add new definitions for ``undamaged fuel 
assembly,'' and ``repaired/reconstituted fuel assembly'' to provide 
further clarity for cask system users and greater consistency with NRC 
guidance for classifying fuel. In addition, the applicant says that 
these definitions will help some BWR users who have older, low-
enriched, channeled BWR fuel with potential cladding defects that these 
users want to load for dry storage without prior placement in a damaged 
fuel container. A discussion of the definition changes follows.
4.a. Definition of ``Undamaged Fuel Assembly''
    The applicant proposed the new definition for ``undamaged fuel 
assembly'' to read: ``a) a fuel assembly without known or suspected 
cladding defects greater than pinhole leaks or hairline cracks and that 
can be handled by normal means; or b) a BWR fuel assembly with an 
intact channel and a maximum average initial enrichment of 3.3 percent 
U-235 by weight (wt-percent) that has no known or suspected grossly 
breached spent fuel rods and can be handled by normal means.'' Under 
this definition, an ``undamaged fuel assembly'' may be a repaired and 
reconstituted fuel assembly.
    The applicant noted that with the currently approved definition, 
inspections to classify the fuel cladding of channeled BWR fuel as 
undamaged may be prohibitively costly and/or unjustifiable for 
maintaining worker radiation exposures as low as reasonably achievable. 
Holtec also noted, however, that a particular subset of older, less-
enriched fuel has been shown to remain subcritical even with 
significant cladding damage and rearrangement of the fuel rods inside 
the channel. If this fuel does not have gross cladding breaches 
(defined as breaches larger than pinhole leaks or hairline cracks), can 
be handled by normal means, and has enrichment less than or equal to 
3.3 weight-percent, Holtec asserted, the fuel does not require a 
damaged fuel container and is not limited to certain basket locations 
in the HI-STORM 100 Cask System's MPC model 68 designed for BWR fuel.
    Under the NRC's ISG-1, ``Classifying the Condition of Spent Nuclear 
Fuel for Interim Storage and Transportation Based on Function,'' 
undamaged fuel may contain some cladding defects if it is safeguarded 
from high temperatures and/or oxidation and does not contain gross 
cladding breaches. Because HI-STORM 100 Cask System MPCs are backfilled 
with helium and shown to keep peak fuel cladding temperatures below the 
limits in ISG 11, ``Cladding Considerations for the Transportation and 
Storage of Spent Fuel,'' the staff has determined that this fuel is 
protected during storage from temperatures that would lead to gross 
ruptures. Also, as long as the fuel meets ISG-1 and does not already 
contain a gross breach, the staff concludes that there are no means for 
the release of fuel fragments during storage. In addition, fuel that 
contains an assembly defect may be considered undamaged under ISG-1 if 
the fuel can still meet fuel-specific and system-related functions. The 
NRC staff will therefore also consider repaired and/or reconstituted 
assemblies meeting these functions as undamaged under the applicant's 
proposed revised definition.
4.b. Definition of ``Repaired/Reconstituted Fuel Assembly''
    As part of Amendment No. 9, Revision 1, Holtec proposed a new 
definition for a repaired or reconstituted fuel assembly as one that 
``contains dummy fuel rod(s) that displaces [sic] an amount of water 
greater than or equal to the original fuel rod(s) and/or which contains 
structural repairs so it can be handled by normal means.'' The 
applicant proposed this definition for clarification purposes and as a 
subset of the definition of ``undamaged fuel.'' It is a common practice 
to repair a nuclear fuel assembly by removing a damaged fuel rod and 
replacing it with a dummy rod to allow the assembly to be returned to 
the reactor core. The NRC has approved this use in specific 
applications, and has provided guidance to 10 CFR part 50 licensees to 
ensure that the repair is performed within the requirements of the 
licensee's 10 CFR part 50 TSs and does not create an unreviewed safety 
question. Because a repaired/reconstituted fuel assembly is restored to 
a condition within the bounds of its original design and safety 
analysis, the NRC staff finds this type of assembly to be a subset of 
``undamaged fuel,'' and concludes that the applicant's proposed 
definition is consistent with ISG-1 and therefore acceptable.

5. Conclusions

    As documented in its Safety Evaluation Report (SER), the NRC staff 
performed a detailed safety evaluation of this proposed CoC amendment 
request. There are no significant changes to cask design requirements 
in the proposed CoC amendment. Considering the specific design 
requirements for each accident condition, the design of the cask would 
prevent loss of containment, shielding, and criticality control. If 
there is no loss of containment, shielding, or criticality control, the 
environmental impacts would be not be significant. This amendment does 
not reflect a significant change in design or fabrication of the cask. 
In addition, any resulting occupational exposures or offsite dose rates 
from the implementation of Amendment No. 9, Revision 1, would remain 
well within 10 CFR part 20 radiation safety limits. Therefore, the 
proposed CoC changes will not result in any radiological or non-
radiological environmental impacts that significantly

[[Page 375]]

differ from the environmental impacts evaluated in the environmental 
assessment (EA) supporting the May 1, 2000, final rule approving the 
original HI-STORM 100 Cask System CoC. There will be no significant 
changes in the types or amounts of any effluent released, no 
significant increase in individual or cumulative radiation exposures, 
and no significant increase in the potential for or consequences of 
radiological accidents.
    This direct final rule revises the HI-STORM 100 Cask System listing 
in 10 CFR 72.214 by adding Amendment No. 9, Revision 1, to CoC No. 
1014. The revision consists of the changes previously described, as set 
forth in the revised CoC and TSs. The revised TSs are identified in the 
SER.
    The revised HI-STORM 100 Cask System design, when used under the 
conditions specified in the CoC, the TSs, and the NRC's regulations, 
will meet the requirements of 10 CFR part 72; therefore, adequate 
protection of public health and safety will continue to be ensured. 
When this direct final rule becomes effective, persons who hold a 
general license under 10 CFR 72.210 may load spent nuclear fuel into 
HI-STORM 100 Cask Systems that meet the criteria of Amendment No. 9, 
Revision 1, to CoC No. 1014 under 10 CFR 72.212.

V. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995 (Pub. 
L. 104-113) requires that Federal agencies use technical standards 
developed or adopted by voluntary consensus standards bodies unless the 
use of such a standard is inconsistent with applicable law or otherwise 
impractical. In this direct final rule, the NRC will revise the Holtec 
HI-STORM 100 Cask System design listed in Sec.  72.214, ``List of 
Approved Spent Fuel Storage Casks.'' This action does not constitute 
the establishment of a standard that contains generally applicable 
requirements.

VI. Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the Commission on June 30, 1997, 
and published in the Federal Register on September 3, 1997 (62 FR 
46517), this rule is classified as Compatibility Category ``NRC.'' 
Compatibility is not required for Category ``NRC'' regulations. The NRC 
program elements in this category relate directly to areas of 
regulation reserved to the NRC by the Atomic Energy Act of 1954, as 
amended, or the provisions of 10 CFR. Although an Agreement State may 
not adopt program elements reserved to the NRC, it may wish to inform 
its licensees of certain requirements using mechanisms consistent with 
the particular State's administrative procedure laws, but classifying 
an NRC rule as Category ``NRC'' does not confer regulatory authority on 
the State.

VII. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, and well-organized 
manner. The NRC has written this document to be consistent with the 
Plain Writing Act as well as the Presidential Memorandum, ``Plain 
Language in Government Writing,'' published June 10, 1998 (63 FR 
31883).

VIII. Environmental Assessment and Finding of No Significant 
Environmental Impact

A. The Action

    The action is to amend 10 CFR 72.214 to revise the Holtec HI-STORM 
100 Cask System listing within the ``List of Approved Spent Fuel 
Storage Casks'' to include Amendment No. 9, Revision 1, to CoC No. 
1014. Under the National Environmental Policy Act (NEPA) of 1969, as 
amended, and the NRC's regulations in subpart A of 10 CFR part 51, 
``Environmental Protection Regulations for Domestic Licensing and 
Related Regulatory Functions,'' the NRC has determined that this rule, 
if adopted, would not be a major Federal action significantly affecting 
the quality of the human environment and, therefore, an environmental 
impact statement (EIS) is not required. The NRC has made a finding of 
no significant impact on the basis of this EA.

B. The Need for the Action

    The need for this direct final rule is to allow users of HI-STORM 
100 Cask Systems under Amendment 9, Revision 1, to load for dry storage 
under a general license some PWR fuel assemblies with shorter cooling 
times for TPDs, and some BWR fuel assemblies that would otherwise have 
to remain in spent fuel storage pools. Specifically, Amendment No. 9, 
Revision 1, changes cooling time limits for TPDs, removes certain 
testing requirements for the fabrication of Metamic HT neutron-
absorbing structural material, and reduces certain MGVs used in 
bounding calculations for this material. Amendment No. 9, Revision 1, 
also changes fuel definitions to classify certain BWR fuel within 
specified guidelines as undamaged fuel, which could avert the worker 
radiation exposures that would otherwise be necessary to put this fuel 
into containers before loading them into MPCs.

C. Environmental Impacts of the Action

    On July 18,1990 (55 FR 29181), the NRC issued an amendment to 10 
CFR part 72 to provide for the storage of spent fuel under a general 
license in cask designs approved by the NRC. The potential 
environmental impact of using NRC-approved storage casks was initially 
analyzed in the EA for the 1990 final rule. The EA for this Amendment 
No. 9, Revision 1, tiers off of that EA for the July 18, 1990, final 
rule. Tiering on past environmental assessments is a standard process 
under NEPA. As stated in the Council on Environmental Quality's 40 
Frequently Asked Questions, the tiering process makes each EIS/EA of 
greater use and meaning to the public as the plan or program develops 
without duplication of the analysis prepared for the previous impact 
statement.
    Holtec HI-STORM 100 Cask Systems are designed to mitigate the 
effects of design basis accidents that could occur during storage. 
Design basis accidents account for human-induced events and the most 
severe natural phenomena reported for the site and surrounding area. 
Postulated accidents analyzed for an independent spent fuel storage 
installation, the type of facility at which a holder of a power reactor 
operating license would store spent fuel in casks in accordance with 10 
CFR part 72, include tornado winds and tornado-generated missiles, a 
design basis earthquake, a design basis flood, an accidental cask drop, 
lightning effects, fire, explosions, and other incidents.
    Considering the specific design requirements for each accident 
condition, the design of the cask would prevent loss of confinement, 
shielding, and criticality control. If there is no loss of confinement, 
shielding, or criticality control, the environmental impacts would be 
insignificant. This revision does not reflect a significant change in 
design or fabrication of the cask. There are no significant changes to 
cask design requirements in the proposed CoC revision. In addition, 
because there are no significant design or process changes, any 
resulting occupational exposures or offsite doses from the 
implementation of Amendment No. 9, Revision 1, would remain well within 
10 CFR part 20 radiation protection limits. Therefore, the proposed CoC 
changes will not result in any radiological or non-radiological 
environmental impacts that differ significantly from the environmental 
impacts evaluated in the EA supporting the July 18, 1990, final rule. 
There will

[[Page 376]]

be no significant change in the types or amounts of any effluent 
released, no significant increase in individual or cumulative radiation 
exposures, and no significant increase in the potential for or 
consequences of radiological accidents. The NRC staff documented these 
safety findings in the SER.

D. Alternative to the Action

    The alternative to this action is to deny approval of Amendment No. 
9, Revision 1, and end the direct final rule. Consequently, any 10 CFR 
part 72 general licensee that seeks to load spent fuel into a HI-STORM 
100 Cask System in accordance with the changes described in proposed 
Amendment No. 9, Revision 1, would have to request an exemption from 
the requirements of 10 CFR 72.212 and 72.214. Under this alternative, 
interested licensees would have to prepare, and the NRC would have to 
review, each separate exemption request, thereby increasing the 
administrative burden on the NRC and the costs to each licensee. The 
environmental impacts of this no-action alternative would therefore be 
the same as or more than those for the action itself.

E. Alternative Use of Resources

    Approval of Amendment No. 9, Revision 1, to CoC No. 1014 would 
result in no irreversible commitments of resources.

F. Agencies and Persons Contacted

    No agencies or persons outside the NRC were contacted in connection 
with the preparation of this EA.

G. Finding of No Significant Impact

    The environmental impacts of the action have been reviewed as 
required by the NRC's 10 CFR part 51 regulations. Based on the 
foregoing EA, the NRC concludes that this direct final rule entitled, 
``List of Approved Spent Fuel Storage Casks: Holtec International HI-
STORM 100 Cask System; Amendment No. 9, Revision 1,'' will not have a 
significant effect on the human environment. Therefore, the NRC has 
determined that an EIS for this direct final rule is not necessary.

IX. Paperwork Reduction Act Statement

    This rule does not contain any information collection requirements, 
and is therefore not subject to the Paperwork Reduction Act of 1995 (44 
U.S.C. 3501 et seq.).

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
Office of Management and Budget control number.

X. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
NRC certifies that this rule will not, if issued, have a significant 
economic impact on a substantial number of small entities. This direct 
final rule affects only nuclear power plant licensees and Holtec. These 
entities do not fall within the definition of small entities set forth 
in the Regulatory Flexibility Act or the size standards established by 
the NRC (10 CFR 2.810).

XI. Regulatory Analysis

    On July 18, 1990 (55 FR 29181), the NRC issued an amendment to 10 
CFR part 72 to provide for the storage of spent nuclear fuel under a 
general license in cask designs approved by the NRC. Any nuclear power 
reactor licensee can use NRC-approved cask designs to store spent 
nuclear fuel if it notifies the NRC in advance, the spent fuel is 
stored under the conditions specified in the cask's CoC, and the 
conditions of the general license are met. A list of NRC-approved cask 
designs is provided in 10 CFR 72.214. On May 1, 2000 (65 FR 25241), the 
NRC issued an amendment to 10 CFR part 72 that approved the HI-STORM 
100 Cask System design by adding it to the list of NRC-approved cask 
designs in 10 CFR 72.214.
    On July 1, 2014, Holtec submitted an application to revise the HI-
STORM 100 Cask System as described in Section III, ``Discussion of 
Changes,'' of this document.
    The alternative to this action is to withhold approval of Amendment 
No. 9, Revision 1, and to require any 10 CFR part 72 general licensee 
seeking to load spent nuclear fuel into a HI-STORM 100 Cask System 
under the changes described in Amendment No. 9, Revision 1, to request 
an exemption from the requirements of 10 CFR 72.212 and 72.214. Under 
this alternative, each interested 10 CFR part 72 licensee would have to 
prepare, and the NRC would have to review, a separate exemption 
request, thereby increasing the administrative burden upon the NRC and 
the costs to each licensee.
    Approval of the direct final rule is consistent with previous NRC 
actions. Further, as documented in the SER and the EA, the direct final 
rule will have no adverse effect on public health and safety or the 
environment. This direct final rule has no significant identifiable 
impact or benefit on other Government agencies. Based on this 
regulatory analysis, the NRC concludes that the requirements of the 
direct final rule are commensurate with the NRC's responsibilities for 
public health and safety and the common defense and security. No other 
available alternative is believed to be as satisfactory, and therefore, 
this action is recommended.

XII. Backfitting and Issue Finality

    For the reasons set forth below, the NRC has determined that the 
backfit rule (10 CFR 72.62) does not apply to this direct final rule, 
and therefore, a backfit analysis is not required.
    This direct final rule revises CoC No. 1014, Amendment No. 9, for 
the Holtec HI-STORM 100 Cask System, as currently listed in 10 CFR 
72.214, ``List of Approved Spent Fuel Storage Casks.'' Amendment No. 9, 
Revision 1, reduces cooling time limits for TPDs in some fuel 
assemblies, removes a thermal conductivity testing requirement for the 
fabrication of Metamic HT neutron-absorbing structural material, and 
reduces the MGVs used in bounding calculations for this material. 
Amendment No. 9, Revision 1, also changes fuel definitions to classify 
certain BWR fuel within specified guidelines as undamaged fuel.
    According to the certificate holder, casks have been manufactured 
under Amendment No. 9, the subject of this revision. Although Holtec 
(applicant, certificate holder) has manufactured some casks under the 
existing CoC No. 1014, Amendment No. 9, that is being revised by this 
direct final rule, Holtec, as the certificate holder, is not subject to 
backfitting protection under 10 CFR 72.62. Moreover, Holtec requested 
the change and requested to apply it to the existing casks manufactured 
under Amendment No. 9. Therefore, even if the certificate holder were 
deemed to be an entity protected from backfitting, this request 
represents a voluntary change and is not backfitting for Holtec.
    Under 10 CFR 72.62, general licensees are entities that are 
protected from backfitting, and in this instance, Holtec has provided 
casks under CoC No. 1014, Amendment No. 9, to general licensees at the 
Braidwood, Byron, Farley, Hatch, and Vogtle reactor facilities. General 
licensees are required, pursuant to 10 CFR 72.212, to ensure that each 
cask conforms to the terms, conditions, and specifications of a CoC, 
and that each cask can be safely used at the specific site in question. 
Because the casks purchased and delivered under CoC No. 1014 Amendment 
No. 9, must now be evaluated under 10 CFR 72.212

[[Page 377]]

consistent with the revisions in CoC No. 1014 Amendment 9, Revision 1, 
this change in the evaluation method and criteria constitutes a change 
in a procedure required to operate an independent spent fuel storage 
installation (ISFSI) and, therefore, would constitute backfitting under 
10 CFR 72.62(a)(2).
    In this instance, however, the affected general licensees 
voluntarily indicated their willingness to comply with the revised CoC. 
In order to provide these general licensees adequate time to implement 
the revised CoC, it now also incorporates a condition that provides 
general licensees 180 days from the effective date of Revision 1 to 
implement the changes authorized by this revision and to perform the 
required evaluation. Therefore, although the general licensees are 
entities that are protected from backfitting, this request represents a 
voluntary change and is not backfitting for the general licensees.
    In addition, the changes in CoC No. 1014, Amendment 9, Revision 1, 
do not apply to casks manufactured to the initial CoC 1014 or 
subsequent Amendments of CoC 1014. These changes therefore have no 
effect on current ISFSI general licensees using casks manufactured to 
the initial CoC 1014 or other amendments of CoC No. 1014. Thus, the NRC 
approval of CoC No. 1014, Amendment No. 9, Revision 1, does not 
constitute backfitting for general licensed users of the Holtec HI-
STORM 100 Cask System that were manufactured to the initial CoC No. 
1014 or to other amendments of CoC No. 1014, under 10 CFR 72.62, 10 CFR 
50.109(a)(1), or the issue finality provisions in 10 CFR part 52.
    For these reasons, no backfit analysis or additional documentation 
addressing the issue finality criteria in 10 CFR part 52 has been 
prepared by the NRC.

XIII. Congressional Review Act

    The Office of Management and Budget has not found this to be a 
major rule as defined in the Congressional Review Act.

XIV. Availability of Documents

    The documents identified in the following table are available to 
interested persons through one or more of the following methods, as 
indicated.

------------------------------------------------------------------------
                   Document                       ADAMS accession No.
------------------------------------------------------------------------
Proposed CoC 1014 Amendment No. 9, Revision 1  ML15156A941
Proposed CoC 1014 Amendment No. 9, Revision 1  ML15156A956
 Technical Specifications, Appendix A.
Proposed CoC 1014 Amendment No. 9, Revision 1  ML15156A970
 Technical Specifications, Appendix B.
Proposed CoC 1014 Amendment No. 9, Revision 1  ML15156A982
 Technical Specifications, Appendix A-100U.
Proposed CoC 1014 Amendment No. 9, Revision 1  ML15156B000
 Technical Specifications, Appendix B-100U.
Preliminary CoC 1014 Amendment No. 9,          ML15156B011
 Revision 1 Safety Evaluation Report.
Request for Revision Application dated July    ML14182A486
 1, 2014.
Notification by general licensees of           ML15240A233
 voluntary acceptance of Revision 1
 requirements dated August 28, 2015.
Interim Staff Guidance 1, Classifying the      ML071420268
 Condition of Spent Nuclear Fuel for Interim
 Storage and Transportation Based on Function.
Interim Staff Guidance 11, Revision 3,         ML033230335
 Cladding Considerations for the
 Transportation and Storage of Spent Fuel.
Interim Staff Guidance 23, Application of      ML103130171
 ASTM Standard Practice C1671-07 when
 performing technical reviews of spent fuel
 storage and transportation packaging
 licensing actions.
------------------------------------------------------------------------

    The NRC may post materials related to this document, including 
public comments, on the Federal rulemaking Web site at https://www.regulations.gov under Docket ID NRC-2015-0156. The Federal 
rulemaking Web site allows you to receive alerts when changes or 
additions occur in a docket folder. To subscribe: (1) Navigate to the 
docket folder (NRC-2015-0156); (2) click the ``Sign up for Email 
Alerts'' link; and (3) enter your email address and select how 
frequently you would like to receive emails (daily, weekly, or 
monthly).

List of Subjects in 10 CFR Part 72

    Administrative practice and procedure, Criminal penalties, 
Hazardous waste, Indians, Intergovernmental relations, Manpower 
training programs, Nuclear energy, Nuclear materials, Occupational 
safety and health, Penalties, Radiation protection, Reporting and 
recordkeeping requirements, Security measures, Spent fuel, 
Whistleblowing.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; the Nuclear Waste Policy Act of 1982, as 
amended; and 5 U.S.C. 552 and 553; the NRC adopts the following 
amendments to 10 CFR part 72:

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE

0
1. The authority citation for part 72 continues to read as follows:

    Authority: Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 
65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 
2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy 
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 
5841, 5842, 5846, 5851); National Environmental Policy Act of 1982, 
secs. 117(a), 132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 
U.S.C. 10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 
10168, 10198(a)); 44 U.S.C. 3504 note.

    Section 72.44(g) also issued under Nuclear Waste Policy Act 
secs. 142(b) and 148(c), (d) (42 U.S.C. 10162(b), 10168(c), (d)).
    Section 72.46 also issued under Atomic Energy Act sec. 189 (42 
U.S.C. 2239); Nuclear Waste Policy Act sec. 134 (42 U.S.C. 10154).
    Section 72.96(d) also issued under Nuclear Waste Policy Act sec. 
145(g) (42 U.S.C. 10165(g)).
    Subpart J also issued under Nuclear Waste Policy Act secs. 
117(a), 141(h) (42 U.S.C. 10137(a), 10161(h)).
    Subpart K also issued under sec. 218(a) (42 U.S.C. 10198).


0
2. In Sec.  72.214, Certificate of Compliance No. 1014 is revised to 
read as follows:


Sec.  72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1014.
    Initial Certificate Effective Date: May 31, 2000.
    Amendment Number 1 Effective Date: July 15, 2002.
    Amendment Number 2 Effective Date: June 7, 2005.
    Amendment Number 3 Effective Date: May 29, 2007.
    Amendment Number 4 Effective Date: January 8, 2008.
    Amendment Number 5 Effective Date: July 14, 2008.

[[Page 378]]

    Amendment Number 6 Effective Date: August 17, 2009.
    Amendment Number 7 Effective Date: December 28, 2009.
    Amendment Number 8 Effective Date: May 2, 2012, as corrected on 
November 16, 2012 (ADAMS Accession No. ML12213A170).
    Amendment Number 9 Effective Date: March 11, 2014, superseded by 
Amendment Number 9, Revision 1, on March 21, 2016.
    xxxx
    Amendment Number 9, Revision 1, Effective Date: March 21, 2016.
    Safety Analysis Report (SAR) Submitted by: Holtec International.
    SAR Title: Final Safety Analysis Report for the HI-STORM 100 Cask 
System.
    Docket Number: 72-1014.
    Certificate Expiration Date: May 31, 2020.
    Model Number: HI-STORM 100.
* * * * *

    Dated at Rockville, Maryland, this 22nd day of December, 2015.

    For the Nuclear Regulatory Commission.
Glenn M. Tracy,
Acting, Executive Director for Operations.
[FR Doc. 2015-33280 Filed 1-5-16; 8:45 am]
BILLING CODE 7590-01-P