[Federal Register Volume 70, Number 108 (Tuesday, June 7, 2005)]
[Rules and Regulations]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-11216]
Rules and Regulations
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Federal Register / Vol. 70, No. 108 / Tuesday, June 7, 2005 / Rules
NUCLEAR REGULATORY COMMISSION
10 CFR Part 72
List of Approved Spent Fuel Storage Casks: HI-STORM 100 Revision
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to revise the Holtec International HI-STORM 100 cask system
listing within the ``List of approved spent fuel storage casks'' to
include Amendment No. 2 to Certificate of Compliance (CoC) Number 1014.
Amendment No. 2 modifies the cask design to include changes to
materials used in construction, changes to the types of fuel that can
be loaded, changes to shielding and confinement methodologies and
assumptions, revisions to various temperature limits, changes in
allowable fuel enrichments, and other changes to reflect current NRC
staff guidance and use of industry codes, under a general license.
DATES: Effective Date: This final rule is effective June 7, 2005.
FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, telephone (301)
415-6219, e-mail [email protected], of the Office of Nuclear Material Safety
and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC
Section 218(a) of the Nuclear Waste Policy Act of 1982, as amended
(NWPA), requires that ``[t]he Secretary [of Energy] shall establish a
demonstration program, in cooperation with the private sector, for the
dry storage of spent nuclear fuel at civilian nuclear reactor power
sites, with the objective of establishing one or more technologies that
the [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, ``[t]he
Commission shall, by rule, establish procedures for the licensing of
any technology approved by the Commission under section 218(a) for use
at the site of any civilian nuclear power reactor.''
To implement this mandate, the NRC approved dry storage of spent
nuclear fuel in NRC-approved casks under a general license, publishing
a final rule in 10 CFR part 72 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 within 10 CFR part 72
entitled, ``Approval of Spent Fuel Storage Casks'' containing
procedures and criteria for obtaining NRC approval of dry storage cask
designs. The NRC subsequently issued a final rule on May 1, 2000 (65 FR
25241), that approved the Holtec International HI-STORM 100 cask design
and added it to the list of NRC-approved cask designs in Sec. 72.214
as CoC No. 1014.
On March 4, 2002, and as supplemented on October 31, 2002; August 6
and November 14, 2003; February 20, April 23, July 22, August 13,
October 14, and December 3, 2004, the certificate holder, Holtec
International, submitted an application to the NRC to amend CoC No.
1014 to modify the cask design to include changes to materials used in
construction, changes to the types of fuel that can be loaded, changes
to shielding and confinement methodologies and assumptions, revisions
to various temperature limits, changes in allowable fuel enrichments,
and other changes to reflect current staff guidance and use of industry
codes, under a general license. The specific changes requested in
Amendment No. 2 to CoC No. 1014 are listed in the Safety Evaluation
Report (SER). No other changes to the HI-STORM-100 cask system design
were requested in this application. The NRC staff performed a detailed
safety evaluation of the proposed CoC amendment request and found that
an acceptable safety margin is maintained. In addition, the NRC staff
has determined that there continues to be reasonable assurance that
public health and safety and the environment will be adequately
This rule revises the HI-STORM 100 cask design listing in Sec.
72.214 by adding Amendment No. 2 to CoC No. 1014. The amendment
consists of changes to the Technical Specifications (TS) as described
above. The particular TS which are changed are identified in the NRC
staff's SER for Amendment No. 2.
The NRC published a direct final rule (70 FR 9504; February 28,
2005) and the companion proposed rule (70 FR 9550) in the Federal
Register to revise the Holtec International HI-STORM 100 cask system
listing in 10 CFR 72.214 to include Amendment No. 2 to the CoC. The
comment period ended on March 30, 2005. One comment letter was received
on the proposed rule. The comments were considered to be significant
and adverse and warranted withdrawal of the direct final rule. A notice
of withdrawal was published in the Federal Register on May 12, 2005; 70
FR 24936. Additionally, the NRC staff amended the TS and the SER to
clarify the leak rate test requirement, as discussed in the response to
The NRC finds that the amended HI-STORM 100 cask system, as
designed and when fabricated and used in accordance with the conditions
specified in its CoC, meets the requirements of part 72. Thus, use of
the amended Holtec International HI-STORM 100 cask system, as approved
by the NRC, will provide adequate protection of public health and
safety and the environment. With this final rule, the NRC is approving
the use of the Holtec International HI-STORM 100 cask system under the
general license in 10 CFR part 72, subpart K, by holders of power
reactor operating licenses under 10 CFR part 50. Simultaneously, the
NRC is issuing a final SER and CoC that will be effective on June 7,
2005. Single copies of the CoC and SER are available for public
inspection and/or copying for a fee at the NRC Public Document Room,
11555 Rockville Pike, Rockville, MD. Copies of the public comments are
available for review in the
NRC Public Document Room, 11555 Rockville Pike, Rockville, MD.
Summary of Public Comments on the Proposed Rule
The NRC received one comment letter on the proposed rule from the
New England Coalition. A copy of the comment letter is available for
review in the NRC Public Document Room, 11555 Rockville Pike,
Rockville, MD. As stated in the proposed rule (70 FR 9550; February 28,
2005), the NRC considered this amendment to be a noncontroversial and
routine action. Therefore, the NRC published a direct final rule (70 FR
9504; February 28, 2005) concurrent with the proposed rule (70 FR 9550;
February 28, 2005). The NRC indicated that if it received a
``significant adverse comment'' on the proposed rule, the NRC would
publish a document withdrawing the direct final rule and subsequently
publish a final rule that addressed comments made on the proposed rule.
The NRC believes some of the issues raised by the commenter were
``significant adverse comments.'' Therefore, the NRC published a notice
withdrawing the direct final rule (70 FR 24936; May 12, 2005). This
subsequent final rule addresses the issues raised by the commenter that
were within the scope of the proposed rule.
Comments on Amendment 2 to the Holtec International HI-STORM 100 Cask
The commenter provided specific comments on the draft CoC, the NRC
staff's preliminary SER, the TS, and the applicant's Topical Safety
Analysis Report. As a result of public comments, both TS 3.1.1 and SER
section 8.4 were amended to clarify the leak rate test requirement.
Other sections of the SER were changed to conform with the
clarification of SER section 8.4. A review of the comments and the NRC
staff's responses follows:
Comment 1: The commenter stated that most changes in the CoC
amendment ``appear to diminish engineering conservation and increase
impact or risk.'' The commenter noted that ``while the changes appear
to be within the bounds of regulation, it is not apparent that NRC or
the CoC holder have demonstrated that diminished engineering
conservation and increased impact or risk are offset by gains and
benefits elsewhere.'' The commenter provided as examples of changes
which diminish engineering conservation ``incorporating the storage of
high burnup fuel and raising maximum permissible fuel cladding
temperatures per Proposed Change Number 15a in LAR 1014 to incorporate
a permissible spent fuel cladding temperature limit of 4000 [deg]C.''
Response: Amendments to a CoC are reviewed under the same criteria
as are used for the approval of the original CoC (10 CFR 72.246). The
applicant for an amendment must show that any changes meet all
applicable requirements to store spent fuel safely in the cask.
However, the applicant is not required to show that a change, which
might be viewed as reducing engineering conservatism, is offset by some
increased gain or benefit elsewhere as long as the change meets all
regulatory requirements for safety. The commenter acknowledges that all
the changes appear to be within the bounds of regulations. The NRC
staff specifically examined the effects of incorporating the storage of
high burnup fuel and incorporating a permissible single spent fuel
cladding temperature limit of 400 [deg]C. It should be noted that the
commenter made an error in stating that Amendment No. 2 raised
``permissible spent fuel cladding temperature limit'' to 4000 [deg]C.
The staff has reviewed the SER of Amendment No. 2 and found 5
references to the fuel temperature of 400 [deg]C on pages 4-2, 4-6, 8-
1(2), and 8-2. There was no mention of a 4000 [deg]C temperature in the
SER. The 570 [deg]C temperature was mentioned a number of times.
Consequently, the potential for a zirconium cladding exothermic
reaction would not be an issue at 400 [deg]C.
Comment 2: The commenter referred to an NRC staff statement that no
review of the existing CoC was repeated. The commenter believes this
may be an error if it also means that no review was undertaken to
ascertain if the changes affect conditions, assumptions, and other
inputs in determining compliance in the original application.
Response: The NRC staff did not state that no review of the
existing CoC was repeated. The SER states that the staff's evaluation
focused mainly on modifications requested in the amendment and did not
reassess previously approved portions of the CoC, TS, and the Final
Safety Analysis Report (FSAR), or those areas of the FSAR modified by
Holtec as allowed by 10 CFR 72.48.
Comment 3: The commenter referred to a specific section in the SER
which would allow ``storage of damaged fuel in the multipurpose
canister (MPC)-32 and damaged fuel and damaged fuel debris in the MPC-
32F. Additionally, include appropriate values for soluble boron for
MPC-32 and MPC-32F based on fuel assembly array/class, intact versus
damaged fuel, and initial enrichment.'' The commenter stated that a
definition of ``damaged fuel'' versus ``fuel debris'' including a
bounding description of ``damaged fuel'' and ``fuel debris'' should be
included. Damaged fuel could range from a rod that marginally failed a
leak test to a fuel fragment. Small, unclad bits of fuel would need to
be properly containerized and those containers certified to some
Response: The definitions of ``damaged fuel'' and ``fuel debris''
are given in section 1.0, Definitions, of Appendix B to the TS attached
to the CoC for Certificate Number 1014, Amendment No. 2. The
definitions contain commonly used terminology to distinguish between
these two classes of contents. The definitions are repeated here:
``DAMAGED FUEL ASSEMBLIES are fuel assemblies with known or
suspected cladding defects, as determined by a review of records,
greater than pinhole leaks or hairline cracks, empty fuel rod locations
that are not filled with dummy fuel rods, or those that cannot be
handled by normal means. Fuel assemblies that cannot be handled by
normal means due to fuel cladding damage are considered FUEL DEBRIS.''
``FUEL DEBRIS is ruptured fuel rods, severed rods, loose fuel
pellets or fuel assemblies with known or suspected defects which cannot
be handled by normal means due to fuel cladding damage.''
``Damaged fuel assemblies'' and ``fuel debris'' must be enclosed in
a specially designed ``damaged fuel container'' before being loaded
into the cask.
Comment 4: The commenter referred to a section in the SER that
stated that the change requested in this amendment affected the
inspection and leak testing of the final closure welds. The applicant
applied the criteria described in ISG-15, ``Materials Evaluation,'' and
ISG-18, ``The Design/Qualification of Final Closure Welds on Austenitic
Stainless Steel Canisters as Confinement Boundary for Spent Fuel
Storage and Containment Boundary for Spent Fuel Transportation,'' in
the amendment request. The commenter further stated that ISG-15
provides an NRC-approved alternative to the ASME Code for the
inspection of final closure welds for austenitic materials. The
inspection techniques described by ISG-15 will detect any such flaws
which could lead to a failure. In addition, ISG-18 states that when the
closure welds of austenitic stainless steel canisters are executed in
accordance with ISG-15, the staff concludes that no undetected flaws of
significant size will exist. Therefore, the NRC staff has reasonable
assurance that the inspection
demonstrates no credible leakage would occur from the final closure
welds of austenitic stainless steel canisters, and that ISG-18 removes
the need for a helium leak test of the final closure welds in
accordance with ANSI N14.5.
The commenter further stated that, in the past, inspection systems
have not been considered adequate for critical welds. A proof-system is
typically required due to the consequence of container leakage for
failure. The commenter believed it should be noted that helium is used
as a leak test agent due to its small size and inert properties. The
commenter did not credit that the inspection system referred to, or any
inspection system that could be used expeditiously, can detect flaws at
the molecular level. The commenter believed it is possible by this
revised process to approve welds that may have ordinarily failed a
helium leak test and stated this change could constitute a significant
reduction in the gas-tight certification of the containers.
Response: Dry storage casks use redundant means to achieve adequate
structural and confinement capability. First, the final closures
incorporate a double barrier. This is accomplished by the use of two
separate welded barriers. For the Holtec design, this is accomplished
by way of the structural lid and a separate closure ring that is welded
over the structural lid. If, in the unlikely event one of these welded
barriers should have a leak, the other would be capable of retaining
all the helium inside the storage canister.
With respect to testing of the various closure welds, a number of
independent tests are employed. During the welding of the structural
lid, Interim Staff Guidance (ISG)-15 specifies that a multi-pass liquid
penetrant test (PT) be employed. This means that a PT exam is performed
several times during the execution of the weld. The NRC staff guidance
calls for the initial weld pass (called root pass) to be examined.
Then, depending upon the results of a fracture mechanics evaluation or
net-section stress calculation, additional PTs are performed each time
a specified thickness of weld metal is deposited. Finally, the last
weld pass (cover pass) is examined by PT. If any flaws are detected by
any of these tests, the indicated flaw is removed by grinding. Then the
affected area is rewelded and retested. Any such rework is governed by
the provisions of the American Society of Mechanical Engineers (ASME)
Upon acceptance of the multiple PT exams, the structural lid weld
is pressure tested in accordance with the ASME Code. This pressure test
is performed at an elevated pressure that is above the design pressure
of the vessel. Holtec may use either water or helium for this pressure
Due to the large size of the structural lid weld (approximately 3/
4-inch thick or greater), it is extremely unlikely that a weld flaw
could exist that provided a leak path completely through the weld, and
that went undetected after multiple PT exams and the Code-required
pressure test. Because of the redundant nature of these independent
tests, the weld thickness, and staff and industry experience with heavy
section welds, it was deemed unnecessary to perform a helium leak test
on the structural lid weld.
After other loading operations are completed, the cask is filled
with helium and the helium pressure is adjusted to the design pressure.
Then the vent and drain valves (used for filling the vessel with
helium) are closed, and the valve access port is covered with a welded-
on closure plate. These final closure welds are both helium leak tested
and penetrant tested.
After successful completion of these required tests, the closure
ring, which provides a second confinement barrier, is welded on over
the structural lid, weld, and associated access port welds. This weld
is penetrant tested.
As a result of the comment regarding leak testing of the final
closure welds, NRC staff reviewed the TS and SER and clarified the
helium leak rate test requirements within these documents.
TS 3.1.1.C was modified to reflect the requirement to helium leak
rate test the vent and drain port cover plate welds. Section 8.4 of the
SER was added to clarify guidance, specifically that the vent and drain
port cover plate welds shall be helium leak rate tested but that it is
not necessary to helium leak rate test the lid-to-shell weld. Other
sections of the SER were revised accordingly to reflect this
The NRC staff finds that with the double confinement barriers and
the multiple tests employed to verify their quality and integrity, a
high level of assurance exists regarding the leak-tightness of the
Comment 5: The commenter referred to section 2.3.5 of the SER,
``Criticality.'' The design criterion for criticality safety is that
the effective neutron multiplication factor, including statistical
biases and uncertainties, does not exceed 0.95 under normal, off-
normal, and accident conditions. The commenter stated that 0.95 is
pretty close to <= 1 multiplication, or criticality. The commenter was
concerned that ``after pencil-whipping a design someone is willing to
work under a margin of error of 0.06.'' The commenter further stated
that the exact interior of the structure, the boron loading of the
Metamic neutron absorber, the exact position of the fuel (damaged or
otherwise) plus other factors, must be within a margin of error,
potentially, of 0.06. The commenter stated it was difficult to credit
that the fuel assemblies are packed so tight that they can be packed to
an MF of 0.94.
Response: A dry-storage cask design which maintains the effective
multiplication factor (keff) <= 0.95 at a 95-percent
confidence level when combined with the additional bounding assumptions
described below is considered by the NRC to provide reasonable
assurance that the cask and its contents will remain sufficiently
subcritical under all credible normal, off-normal, and accident
conditions. This acceptance criterion is specified in section 6.0,
subsection IV, of the ``Standard Review Plan for Dry Cask Storage
In addition to the administrative margin described above (i.e.,
when the final adjusted value of keff is at least 0.05 below
the critical value of 1.0), the applicant applied the following
bounding assumptions in its criticality analysis:
(1) No credit was taken for fuel burnup;
(2) The worst hypothetical combination of tolerances (i.e., those
value limits which maximized the multiplication factor) was assumed for
the basket structure and fuel assembly dimensions;
(3) Reduced credit from the minimum acceptable boron content in the
poison plates (25-percent reduction for Boral plates and 10-percent
reduction for the Metamic plates) was applied;
(4) Fuel related burnable neutron absorbers were neglected;
(5) Each fuel assembly was placed in its most reactive position
within its respective basket fuel cell;
(6) Neutron absorption in minor structural members and optional
heat conducting elements were neglected; and
(7) The flooding water (fresh or borated) was assumed to be at its
optimum density to maximize keff.
These bounding assumptions are consistent with NRC's guidance and
provide an additional margin of safety that encompasses any margin of
error in the nominal parameter values of the design and contents.
Comment 6: The commenter did not believe that the NRC staff
demonstrated consideration of a reasonably assumed error bandwidth
within each of the
seven coefficients (inputs) to the equation listed in Equation 126.96.36.199.
The commenter stated that the cumulative error potential is large
enough to have ``Biblical'' overtones, as in ``77 times 7.'' The
commenter also stated that one would like to assume that parallel
calculations were performed using traditional methods as a ``sanity
check.'' The commenter believed that with unique source-term analyses
and curve-fitting analyses designed by the applicant to drive the
coefficients, verification and validation information regarding this
burnup model is essential and should be included or referenced in the
Response: The comment expresses a concern regarding error in the
applicant's new methodology and the need for confirmatory analysis to
verify and validate the burnup equation and its coefficients. The
existing sections 5.0, 5.2.3, and 5.2.4 of the SER address this concern
and document that the NRC staff reviewed and explicitly considered the
applicant's methodology, the burnup equation, and its coefficients,
which include adjustments that account for error and uncertainty. As
part of its review, the staff performed confirmatory analyses, using
Computer Code SAS-2H, to test the validity of the burnup equation and
its associated coefficients. These calculations produced decay heats
that were in general agreement with the burnups and associated thermal
values applied in the burnup equation. The NRC staff did not identify
any significant errors in the new methodology, the burnup equation, and
its coefficients. The staff believes that its review of the new
methodology, including confirmatory calculations, provides reasonable
assurance that the shielding and thermal design is safe and satisfies
the regulations at 10 CFR part 72.
Comment 7: The commenter stated that NRC shot the SER through with
subjective language. The example given was ``The amendment request
addresses a slight increase of 10% in the off-normal internal design.''
The commenter objected to using the word ``slight'' and stated that
describing a 10% increase as slight is amateurish in regulatory
language or in any technical document and gives the appearance of
collusion, as if to help sell to the audience any changes that are less
conservative. The commenter questioned if a 10% reduction in the
allowable pressure would be described as huge.
Response: Section 3.0 of the SER provides an overview of the
structural evaluation. The full text of the third paragraph of that
section to which the commenter referred is as follows:
``The amendment request addresses a slight increase of 10% in the
off-normal internal design pressure, increases in the allowable
temperature of the structural materials and the creation of an eighth
type MPC unit: The MPC-32F. No changes were made to the drawings of the
various components that have been previously provided in Section 1.5 of
the FSAR since no material or design dimensions were revised.''
On page S-2 of the SER, the following is stated in Item 16:
``Increase off-normal design pressure from 100 psig to 110 psig and
increase the normal temperature limit for the overpack lid top plate
from 350-degrees F to 450-degrees F.'' This reflects the change
incorporated into the Amendment 2 documents.
Section 188.8.131.52 of the SER, ``Criteria for Multi-Purpose Dry
Storage Canisters,'' contains the following statements: ``The proposed
amendment revises the MPC off-normal internal pressure from 100 psig to
110 psig as noted in Table 2.2.1 of the FSAR * * *. No physical changes
were necessary to accommodate the revised pressure * * *.''
The technical document is quite clear in the fact that the increase
of 10 psig (an increase of 10 percent) has no impact on the physical
dimensions or design of the MPC pressure vessel. The reason for this is
that the physical dimensions of the MPC are not governed by the off-
normal internal pressure.
Comment 8: The commenter stated that there is an element of
vagueness in the SER that offers little guidance to a reader seeking to
confirm the degree of rigor to which the amendment application was
exposed. The NRC refers to many staff reviews of the licensee's
practices, but without specifics. In some cases, it is inferred that
the staff verified calculations; in others, that approval was cursory
because of similarities with other cask models. It is difficult to say
that early cask designs will be safe in the long term. One has to be
careful in approving a new design that is ``similar'' to the old one
when the old one has not yet met the test of time.
Response: NRC disagrees with the commenter that this amendment
application was not exposed to a sufficient degree of rigor. This
amendment request was under active review by the NRC staff for over
2.75 years. As discussed in the response to Comment 1,
amendments to a CoC are reviewed under the same criteria as are used
for the approval of the original CoC (10 CFR 72.246). Also, the
application for an amendment must show that any changes meet all
applicable requirements to store spent fuel safely in the cask. NRC's
review process is documented in NUREG-1536 entitled ``Standard Review
Plan for Dry Cask Storage Systems.'' NRC regulations permit applicants
to demonstrate compliance by various means, including certification
through testing, analyses, comparison to similar approved designs, or
combinations of these methods. Referencing previously reviewed
information that has not changed is acceptable. The SER documents the
NRC's review process and conclusions regarding the cask design's
ability to comply with part 72. Furthermore, this amendment will not
extend the CoC period. Therefore, it does not change the conclusion
reached previously regarding the safety of the cask with respect to
Comment 9: The commenter is concerned that the NRC review does not
extend beyond a review of the proposed theoretical model. The commenter
also stated that the application spoke very little about QA/QC with
respect to cask/canister materials and performance.
Response: The NRC conducts planned and reactive inspections of cask
vendors and their major fabricators on a continuing basis. The results
of these inspections, including any technical concerns of a licensing
nature, are shared internally with the NRC's Spent Fuel Project Office
staff, and are documented in publicly available inspection reports.
Quality assurance program implementation inspections were performed at
the Holtec corporate office in September 2004 (reference ML043080505)
and its fabricator, U.S. Tool & Die, in October 2004 (reference
ML043100408). No significant adverse findings with respect to quality
assurance/control issues were identified during those inspections.
Summary of Final Revisions
Section 72.214 List of Approved Spent Fuel Storage Casks
Certificate No. 1014 is revised by adding the effective date of
Amendment Number 2.
Good Cause To Dispense With Deferred Effective Date Requirement
The NRC finds that good cause exists to waive the 30-day deferred
effective date provisions of the Administrative Procedure Act (5 U.S.C.
553(d)). The primary purpose of the delayed effective date requirement
is to give affected persons, e.g., licensees, a reasonable time to
prepare to comply with or take other action with respect to the rule.
this case, the rule does not require any action to be taken by
licensees. The regulation allows, but does not require, use of the
amended Holtec International HI-STORM 100 cask system for the storage
of spent nuclear fuel. The Holtec International HI-STORM 100 cask
system, amended to include changes to materials used in construction,
changes to the types of fuel that can be loaded, changes to shielding
and confinement methodologies and assumptions, revisions to various
temperature limits, changes in allowable fuel enrichments, and other
changes to reflect current staff guidance and use of industry codes,
meets the requirements of 10 CFR part 72, and is ready to be used. A
number of utilities have an operational need to load the casks to
preserve full core off-load capability at their sites. The utilities
are preparing for refueling outages in Fall of 2005 and need to load
fuel into the storage casks in advance of the outages. The amended
Holtec International HI-STORM cask system, as approved by the NRC, will
continue to provide adequate protection of public health and safety and
Voluntary Consensus Standards
The National Technology Transfer Act of 1995 (Pub. L. 104-113)
requires that Federal agencies use technical standards that are
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 final rule, the NRC is revising the HI-STORM 100
cask system design listed in Sec. 72.214 (List of NRC-approved spent
fuel storage cask designs). This action does not constitute the
establishment of a standard that establishes generally applicable
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 are those that relate directly to
areas of regulation reserved to the NRC by the Atomic Energy Act of
1954, as amended (AEA), or the provisions of Title 10 of the Code of
Federal Regulations. Although an Agreement State may not adopt program
elements reserved to NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with the
particular State's administrative procedure laws but does not confer
regulatory authority on the State.
Finding of No Significant Environmental Impact: Availability
Under the National Environmental Policy Act of 1969, as amended,
and the NRC regulations in subpart A of 10 CFR part 51, the NRC has
determined that this rule is not a major Federal action significantly
affecting the quality of the human environment and, therefore, an
environmental impact statement is not required. This final rule amends
the CoC for the HI-STORM 100 cask system within the list of approved
spent fuel storage casks that power reactor licensees can use to store
spent fuel at reactor sites under a general license. The amendment
modifies the present cask system design to include changes to materials
used in construction, changes to the types of fuel that can be loaded,
changes to shielding and confinement methodologies and assumptions,
revisions to various temperature limits, changes in allowable fuel
enrichments, and other changes to reflect current NRC staff guidance
and use of industry codes, under a general license. The EA and finding
of no significant impact on which this determination is based are
available for inspection at the NRC Public Document Room, 11555
Rockville Pike, Rockville, MD. Single copies of the EA and finding of
no significant impact are available from Jayne M. McCausland, Office of
Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone (301) 415-6219, e-mail
Paperwork Reduction Act Statement
This final rule does not contain a new or amended information
collection requirement subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the
Office of Management and Budget, Approval Number 3150-0132.
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
OMB control number.
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, 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
contained in Sec. 72.214. On May 1, 2000 (65 FR 25241), the NRC issued
an amendment to part 72 that approved the HI-STORM 100 cask design by
adding it to the list of NRC-approved cask designs in Sec. 72.214. On
March 4, 2002, and as supplemented on October 31, 2002; August 6 and
November 14, 2003; February 20, April 23, July 22, August 13, October
14, and December 3, 2004, the certificate holder, Holtec International,
submitted an application to the NRC to amend CoC No. 1014 to modify the
present cask system design to include changes to materials used in
construction, changes to the types of fuel that can be loaded, changes
to shielding and confinement methodologies and assumptions, revisions
to various temperature limits, changes in allowable fuel enrichments,
and other changes to reflect current staff guidance and use of industry
codes, under a general license.
The alternative to this action is to withhold approval of this
amended cask system design and issue an exemption to each utility. This
alternative would cost both the NRC and the utilities more time and
money because each utility would have to pursue an exemption.
Approval of the final rule will eliminate this problem and is
consistent with previous NRC actions. Further, the final rule will have
no adverse effect on public health and safety. This final rule has no
significant identifiable impact or benefit on other Government
agencies. Based on this discussion of the benefits and impacts of the
alternatives, the NRC concludes that the requirements of the 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 thus, this action is
Regulatory Flexibility Certification
In accordance with 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 the licensing and operation of
nuclear power plants, independent spent fuel storage facilities, and
Holtec International. The companies that own these plants do not fall
within the scope of the definition of ``small entities'' set
forth in the Regulatory Flexibility Act or the Small Business Size
Standards set out in regulations issued by the Small Business
Administration at 13 CFR part 121.
The NRC has determined that the backfit rule (10 CFR 50.109 or 10
CFR 72.62) does not apply to this direct final rule because this
amendment does not involve any provisions that would impose backfits as
defined. Therefore, a backfit analysis is not required.
Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996, the NRC has determined that this action is not a
major rule and has verified this determination with the Office of
Information and Regulatory Affairs, Office of Management and Budget.
List of Subjects in 10 CFR Part 72
Administrative practice and procedure, Criminal penalties, Manpower
training programs, 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; and 5 U.S.C. 552 and 553; the NRC is adopting 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
1. The authority citation for part 72 continues to read as follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat.
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-
486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135,
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153,
10155, 10157, 10161, 10168); sec. 1704, 112 Stat. 2750 (44 U.S.C.
Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),
10168(c),(d)). Section 72.46 also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2244 (42 U.S.C. 10101,
10137(a), 10161(h)). Subparts K and L are also issued under sec.
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252
(42 U.S.C. 10198).
2. In Sec. 72.214, Certificate of Compliance 1014 is revised to read
Sec. 72.214 List of approved spent fuel storage casks.
* * * * *
Certificate Number: 1014.
Initial Certificate Effective Date: June 1, 2000.
Amendment Number 1 Effective Date: July 15, 2002.
Amendment Number 2 Effective Date: June 7, 2005.
SAR Submitted by: Holtec International.
SAR Title: Final Safety Analysis Report for the HI-STORM 100 Cask
Docket Number: 72-1014.
Certificate Expiration Date: June 1, 2020
Model Number: HI-STORM 100
* * * * *
Dated at Rockville, Maryland, this 25th day of May, 2005.
For the Nuclear Regulatory Commission.
Luis A. Reyes,
Executive Director for Operations.
[FR Doc. 05-11216 Filed 6-6-05; 8:45 am]
BILLING CODE 7590-01-P