[Federal Register Volume 72, Number 80 (Thursday, April 26, 2007)]
[Rules and Regulations]
[Pages 20712-20716]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-8033]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 72
RIN 3150-AH98
List of Approved Spent Fuel Storage Casks: HI-STORM 100 Revision
3
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations by revising the Holtec International HI-STORM 100 cask
system listing within the ``List of approved spent fuel storage casks''
to include Amendment No. 3 to Certificate of Compliance Number 1014.
Amendment No. 3 revises Technical Specification (TS) 3.1.3, to
eliminate cooling of the Multi-Purpose Canister (MPC) cavity prior to
reflood with water, as part of cask unloading operations; TS 3.3.1, to
allow linear interpolation between minimal soluble boron
concentrations, for certain fuel enrichments in the MPC-32/32F;
Appendix B, Section 1 to the CoC, to make modifications to the
definitions of fuel debris, damaged fuel assembly, and non-fuel
hardware; and Appendix B, Section 2 to the CoC, to permit the storage
of pressurized water reactor fuel assemblies with annular fuel pellets
in the top and bottom 12 inches of the active fuel length. Other
changes are made to incorporate minor editorial corrections. This final
rule allows the holders of power reactor operating licenses to store
spent fuel in this approved cask in accordance with the revised
conditions, under the NRC's general license provisions.
DATES: The final rule is effective on May 29, 2007.
ADDRESSES: Publicly available documents related to this rulemaking may
be viewed electronically on the public computers located at the NRC's
Public Document Room (PDR), Room O1F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland. The PDR reproduction contractor
will copy documents for a fee. Selected documents can be viewed and
downloaded electronically via the NRC's rulemaking Web site at http://ruleforum.llnl.gov.
Publicly available documents created or received at the NRC are
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this site, the public can gain
entry into the NRC's Agencywide Document Access and Management System
(ADAMS), which provides text and image files of NRC's public documents.
If you do not have access to ADAMS or if there are any problems in
accessing the documents located in ADAMS, contact the NRC PDR Reference
staff at (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Federal
and State Materials and Environmental Management Programs, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
6219, e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
Background
Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA), as
amended, requires that ``[t]he Secretary [of the Department of Energy
(DOE)] 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 [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 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 by
publishing a final rule in 10 CFR Part 72 entitled ``General License
for Storage of Spent Fuel at
[[Page 20713]]
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 spent fuel storage cask designs. The NRC
subsequently issued a final rule on May 1, 2000 (65 FR 25241) 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 as Certificate of
Compliance Number (CoC No.) 1014.
Discussion
On November 7, 2005, and as supplemented on April 30, 2006, the
certificate holder, Holtec International, submitted an application to
the NRC to amend the HI-STORM 100 cask system. The application
requested changes to eliminate cooling of the MPC cavity prior to
reflood with water as part of cask unloading operations; changes to
allow linear interpolation between minimal soluble boron concentrations
for certain fuel enrichments in the MPC-32/32F; modifications to the
definitions of fuel debris, damaged fuel assembly, and non-fuel
hardware; changes to permit the storage of pressurized water reactor
fuel assemblies with annular fuel pellets in the top and bottom 12
inches of the active fuel length; and other changes to incorporate
minor editorial corrections. No other changes to the HI-STORM 100 cask
system 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 protected.
The NRC published a direct final rule (71 FR 60659; October 16,
2006) and the companion proposed rule (71 FR 60672) in the Federal
Register to amend the HI-STORM 100 cask system listing in 10 CFR 72.214
to include the changes requested by Holtec International as Amendment
No. 3 to CoC No. 1014. The comment period ended on November 15, 2006.
One comment letter was received on the proposed rule. The comments
contained within the letter 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 December 27, 2006
(71 FR 77586). Additionally, the NRC staff is amending the TS to remove
non-fuel hardware from the definition of fuel debris, as discussed in
the response to Comment C.1 in the preamble. The safety evaluation
report (SER) has been modified to describe the NRC's staff's
determination.
The NRC finds that the Holtec International HI-STORM cask system,
as designed and when fabricated and used in accordance with the
conditions specified in its CoC, meets the requirements of 10 CFR Part
72. Thus, use of the Holtec International HI-STORM 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 May
29, 2007. 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.
Discussion of Amendments by Section
Section 72.214 List Of Approved Spent Fuel Storage Casks
Certificate No. 1014 is revised by adding the effective date of
Amendment Number 3.
Summary of Public Comments on the Proposed Rule
The NRC received one comment letter on the proposed rule from
Public Citizen and the Nuclear Information and Resource Service. Copies
of the public comment letter are available for review in the NRC's
Public Document Room, O-1F21, One White Flint North, 11555 Rockville
Pike, Rockville, Maryland.
Comments on the Holtec HI-STORM 100 Cask System Revision 3
The commenters provided specific comments on Holtec's TS. To the
extent possible, the comments on a particular subject are grouped
together. The listing of the Holtec HI-STORM 100 cask system within 10
CFR 72.214, ``List of approved spent fuel storage casks,'' has not been
changed as a result of the public comments. A review of the comments
and the NRC staff's responses follow:
A. Holtec's Proposal To Eliminate Cooling of the MPC Cavity Prior to
Reflood With Water as Part of Cask Unloading Operations
Comment A.1: The commenters stated that if adequate cooling is not
done prior to reflooding with water during cask unloading, the casks
could experience brittle fracturing caused by a sudden temperature
change from hot to cold. The fracturing could be in addition to the
brittle fracturing already introduced into the casks by forced cooling
during their original manufacture. The commenters stated that forced
cooling violates NRC regulations and applicable ASME and ANSI codes.
Response: The Holtec spent fuel canisters are fabricated from
austenitic stainless steel. This is an extremely tough material with
excellent ductility at all temperatures. Also, this material does not
have a ductile-to-brittle transformation temperature that is typical of
some other types of steel. Hence, this material and the casks which are
fabricated from it are not susceptible to any kind of brittle fracture
as suggested by the comment. For the design environmental temperatures,
the range varies from -40 degrees F to 775 degrees F for the MPC, and
this range of temperatures formed the design bases for the MPC of the
cask system. The structural analyses performed for the cask system
considered this range of temperatures.
There are no heating or cooling rate restrictions imposed by any
regulatory or code requirement for this material or for this
application.
Comment A.2: The commenters stated that during welding, the
strength of the material decreases dramatically with the increased
temperature of the material. After welding, Federal regulations require
cooling at 100 degrees F without forced cooling. They further stated
that if the material does not cool properly, voids inside the heated
zones caused by welding could remain and cause cracking in the future,
and that these cracks may not be detected by testing that is performed
immediately after cooling. The commenters believed that the potential
delayed cracking is the reason why Federal regulations require specific
tests to assess whether the material's strength, which is reduced by
welding, is returned to its original design strength. The commenters
believed that such cracking is also why forced cooling, such as
immersion in water baths or forced air fan cooling, is not allowed by
NRC regulations and applicable ASME and ANSI codes.
[[Page 20714]]
Response: As stated in the response to Comment A.1, above, there is
no regulatory or code requirement or restriction for heating or cooling
rates for austenitic stainless steel, nor is there any need to impose
such requirements. Further, cooling rates as alluded to by the
commenters only apply during post-weld heat treatment (PWHT). PWHT is
not required by the ASME code for this material, nor is it desirable
because of the deleterious effect the PWHT temperatures would have upon
the fuel payload.
The part of the cask which is welded while the cask is in the
loading pool is some distance from the surface of the loading pool
during welding of the closure lid. Any potential ``forced cooling''
effect by the pool water would be negligible compared to the normally
occurring cooling effect which arises from the thermal mass of the
structural lid which is being welded. Likewise, the inert gas purge
which is employed during welding is just sufficient to displace any
hydrogen which may evolve from the fuel payload. It also provides a
backing gas to protect the root pass of the weld from oxidation. It is
insufficient to provide any significant cooling effect. To provide any
significant cooling would require a gas flow such that welding would
not be possible.
No credible delayed cracking mechanism exists for this material,
unlike the situation for other types of steel. Given this, the
excellent ductility of the material, and the lack of any kind of
ductile-to-brittle transformation for the material, no suggested
``brittle fracturing'' mechanism is credible.
Comment A.3: The commenters stated that nine quality assurance (QA)
violations affecting Holtec casks at the U.S. Tool and Die factory in
Pittsburgh, PA, were identified by a former senior lead QA inspector
for Commonwealth Edison/Exelon and his QA team in June and July 2000.
The major QA violations included regulatory code violations, weld
flaws, design flaws, and manufacturing flaws that call into question
the structural integrity of the Holtec shipping containers, especially
under transport accident conditions. The commenters stated that an NRC
Region III dry cask inspector shared the concerns about the QA
violations, and that despite this, NRC failed to address these issues.
Response: Region III forwarded the allegations raised by the former
senior lead QA inspector for Commonwealth Edison/Exelon and his QA team
in June and July 2000 to the former Spent Fuel Project Office (SFPO) at
NRC Headquarters in memoranda dated February 20, 2002, and April 19,
2002. SFPO staff reviewed his allegations and concluded that there were
no safety-significant problems with Holtec's QA program, and more
importantly, that there were no identified defects in any casks
previously manufactured. When the former senior lead QA inspector for
Commonwealth Edison/Exelon asserted that NRC did not adequately address
his issues, the NRC 's independent Office of the Inspector General
(OIG) conducted an investigation. The OIG report, dated July 27, 2004
(available on the NRC website: http://www.nrc.gov/reading-rm.html),
concluded that: (1) The NRC staff did not fail to provide adequate
oversight of Holtec and U.S. Tool and Die; (2) the NRC appropriately
inspected those companies, found deficiencies, and verified that
corrective actions were taken; (3) NRC's handling of the allegations
from the former senior lead QA inspector for Commonwealth Edison/Exelon
was appropriate; and (4) the NRC conducted a timely inspection and had
a valid basis to determine that no safety significant problems existed.
B. Holtec's Proposal To Allow Linear Interpolation Between Minimal
Soluble Boron Concentrations for Certain Fuel Enrichments in the MPC-
32/32F
Comment: The commenters stated that boron concentrations must be
maintained very carefully given the risk of inadvertent criticality due
to the fissile materials (such as U-235 and Pu-239) still present in
the irradiated fuel. They stated that the NRC should not allow
rollbacks on criticality safety regulations.
Response: During the review of the proposed amendment the staff
carefully considered the additional risk of an inadvertent criticality
given a corresponding reduction in the soluble boron levels based on
enrichment. The original requirement to load any fuel over 4.1 weight
percent uranium-235 as if it were 5.0 weight percent uranium-235 fuel
was extremely conservative. Based on the study performed in the license
amendment request, staff finds that linear interpolation of the soluble
boron levels is conservative in this instance and continues to provide
an ample margin of safety against inadvertent criticality.
C. Holtec's Proposal To Modify the Definitions of Fuel Debris, Damaged
Fuel Assembly, and Nonfuel Hardware
Comment C.1: The commenters stated that fuel debris and damaged
fuel assemblies are among the most risky high-level radioactive waste
to handle, store, transport, and dispose of, because the integrity of
the fuel cladding has been ruined. They stated that radioactive
particles and gases and entire nuclear fuel pellets are able to escape
the fuel rods, worsening contamination of the Holtec inner canister and
cask systems. They believed this could increase radiation doses for
nuclear workers and the public as well as increase criticality risks in
certain accident scenarios such as underwater submersions. Thus, the
commenters believe that the definitions of these terms should not be
modified.
Response: In its review of the final rule that added the Holtec HI-
STORM 100 cask system to the listing in 10 CFR 72.214 (65 FR 25241; May
1, 2000), the staff found that fuel debris, as defined in that
amendment, can be stored safely in the HI-STORM 100 cask system. The
basis for the staff's finding is explained in the SER for that final
rule. The current amendment does propose, however, to expand the
definition of fuel debris to include additional materials.
In response to the comment, staff reevaluated this proposal and
determined that expanding the definition of fuel debris to include
containers or structures that are supporting intact or damaged fuel
assembly parts is acceptable, as stated in the SER. However, staff
determined that expanding the definition of fuel debris to include non-
fuel hardware in order to permit storage of non-fuel hardware
separately from (i.e., not within) a fuel assembly was not acceptable,
and modified the Technical Specifications to remove non-fuel hardware
from the definition of fuel debris. The SER has been modified to
describe the staff's determination.
Comment C.2: The commenters stated that the nonfuel hardware is a
hazardous material due to the radioactive contamination and radioactive
activation it has experienced and presents a danger to workers and the
public.
Response: In its review of Amendment 1 to the HI-STORM 100 cask
system (67 FR 46369; July 15, 2002), the staff found that non-fuel
hardware, as defined in that amendment, can be stored safely in the HI-
STORM 100 cask system. The basis for the staff's finding is explained
in the SER for that previous amendment. The current amendment proposes
to add neutron source assemblies (NSA) to the definition of allowable
non-fuel hardware and limits the number and the locations of NSAs to
one per MPC stored in one of the four center-most fuel basket
positions. Also, the staff found in its review that the shielding
source term for an NSA is bounded by
[[Page 20715]]
the shielding source terms of the cask contents approved in the
previous amendment. Thus, the staff finds the cask system can safely
store non-fuel hardware as defined in the current amendment.
D. Holtec's Proposal To Permit the Storage of Pressurized Water Reactor
(PWR) Fuel Assemblies With Annular Fuel Pellets in the Top and Bottom
12 Inches of the Active Fuel Length
Comment: The commenters expressed concern that permitting the
storage of PWR fuel assemblies with annular fuel pellets in the top and
bottom 12 inches of the active fuel length would risk increasing doses
to nuclear workers and the public during cask loading, handling,
storage, transport, and disposal operations. They stated that this
storage should not be allowed by NRC.
Response: The current amendment proposes to modify the allowable
PWR contents to included PWR assemblies containing annular fuel pellets
in the top and bottom 12 inches of the active fuel length. NRC staff
considered the difference between annular and solid fuel pellets in
this part of the fuel from two aspects--source term and shielding--and
concluded that the effect would not be noticeable. The annular pellet
would produce a smaller source term than the solid pellet, since there
is less fuel in the annular pellet, though the difference would be
small, considering the lower burnup that the ends of the active fuel
experience and the fact that the majority of fissions occur in the
outer portions of a fuel pellet. Also, while solid pellets may be more
effective than annular pellets as shielding, the amount of shielding
provided by the MPC lid and the cask lid would make this effect small.
Thus, the staff finds that the cask system can safely store PWR
assemblies with annular pellets in the top and bottom 12 inches of the
active fuel length.
Summary of Final Revisions
In Appendix B to the CoC, Section 1.0, Definitions, the TS has been
revised in response to Comment C.1. to remove non-fuel hardware from
the definition of fuel debris. The SER has also been revised to
document this change.
Voluntary Consensus Standards
The National Technology Transfer and Advancement 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 10 CFR 72.214 (List of NRC-
approved spent fuel storage cask designs). This action does not
constitute the establishment of a standard that contains generally
applicable requirements.
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 (AEA), as amended, 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, if adopted, would not be 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.
Amendment No. 3 modifies the present cask system design by revising TS
3.1.3 to eliminate cooling of the MPC cavity prior to reflood with
water as part of cask unloading operations; TS 3.3.1 to allow linear
interpolation between minimal soluble boron concentrations for certain
fuel enrichments in the MPC-32/32F; Appendix B, Section 1 to the CoC,
to make modifications to the definitions of fuel debris, damaged fuel
assembly, and non-fuel hardware; and Appendix B, Section 2 to the CoC,
to permit the storage of pressurized water reactor fuel assemblies with
annular fuel pellets in the top and bottom 12 inches of the active fuel
length. Other changes are made to incorporate minor editorial
corrections.
The environmental assessment (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
[email protected].
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.
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, 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 10 CFR 72.214. On May 1, 2000 (65 FR 25241), the NRC
issued an amendment to 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 November 7, 2005, and as supplemented on April 30,
2006, the certificate holder, Holtec International, submitted an
application to the NRC to amend the HI-STORM 100 cask system. The
amendment revises TS 3.1.3 to eliminate cooling of the MPC cavity prior
to reflood with water as part of cask unloading operations; TS 3.3.1 to
allow linear interpolation between minimal soluble boron concentrations
for certain fuel enrichments in the MPC-32/32F; Appendix B, Section 1
to the CoC, to make modifications to the
[[Page 20716]]
definitions of fuel debris, damaged fuel assembly, and non-fuel
hardware; and Appendix B, Section 2 to the CoC, to permit the storage
of pressurized water reactor fuel assemblies with annular fuel pellets
in the top and bottom 12 inches of the active fuel length. Other
changes are made to incorporate minor editorial corrections. The
alternative to this action is to withhold approval of this amended cask
system design. Withholding approval, in the absence of any safety
reason for doing so, would not comply with the requirements of sections
218(a) and 133 of the Nuclear Waste Policy Act.
Approval of the final rule 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 recommended.
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 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.
Backfit Analysis
The NRC has determined that the backfit rule (10 CFR 50.109 or 10
CFR 72.62) does not apply to this final rule because this amendment
does not involve any provisions that would impose backfits as defined.
Therefore, a backfit analysis is not required.
Congressional Review Act
Under the Congressional Review 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.
0
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
0
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.
3504 note); sec. 651(e), Pub. L. 109-58, 119 Stat. 806-10 (42 U.S.C.
2014, 2021, 2021b, 2111).
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, 2224 (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).
0
2. In Sec. 72.214, Certificate of Compliance 1014 is revised to read
as follows:
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.
Amendment Number 3 Effective Date: May 29, 2007.
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: June 1, 2020.
Model Number: HI-STORM 100.
* * * * *
Dated at Rockville, Maryland, this 13th day of April, 2007.
For the Nuclear Regulatory Commission.
Luis A. Reyes,
Executive Director for Operations.
[FR Doc. E7-8033 Filed 4-25-07; 8:45 am]
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