List of Approved Spent Fuel Storage Casks: HI-STORM 100 Revision, 32977-32982 [05-11216]

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• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 108 (Tuesday, June 7, 2005)]
[Rules and Regulations]
[Pages 32977-32982]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-11216]



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Federal Register / Vol. 70, No. 108 / Tuesday, June 7, 2005 / Rules 
and Regulations

[[Page 32977]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AH64


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 jmm2@nrc.gov, of the Office of Nuclear Material Safety 
and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.

SUPPLEMENTARY INFORMATION:

Background

    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.

Discussion

    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 
protected.
    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 
Comment 4.
    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

[[Page 32978]]

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 
System

    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 
degree.
    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

[[Page 32979]]

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) 
Code.
    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 
test.
    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 
clarification.
    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 
confinement boundary.
    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 
Systems,'' NUREG-1536.
    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

[[Page 32980]]

seven coefficients (inputs) to the equation listed in Equation 2.1.9.3. 
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 
SER.
    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 3.1.2.1 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 
time.
    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. 
In

[[Page 32981]]

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 
the environment.

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 
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, 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 
jmm2@nrc.gov.

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 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 
recommended.

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

[[Page 32982]]

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 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.

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).

    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).


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.
    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 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