[Federal Register Volume 65, Number 82 (Thursday, April 27, 2000)]
[Rules and Regulations]
[Pages 24623-24631]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-10392]



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Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules 
and Regulations

[[Page 24623]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG36


List of Approved Spent Fuel Storage Casks: PSNA VSC-24 Revision

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to revise the Pacific Sierra Nuclear Associates (PSNA) VSC-
24 cask system listing within the ``List of approved spent fuel storage 
casks'' to include Amendment No. 1 to the Certificate of Compliance. 
Amendment No. 1 will modify the present cask system design to permit a 
licensee to store burnable poison rod assemblies in the VSC-24 cask 
system with the spent fuel under a general license.

EFFECTIVE DATE: This final rule is effective on May 30, 2000.

FOR FURTHER INFORMATION CONTACT: Richard Milstein, telephone (301) 415-
8149, e-mail <a href="/cdn-cgi/l/email-protection#97e5fefad7f9e5f4b9f0f8e1"><span class="__cf_email__" data-cfemail="9deff4f0ddf3effeb3faf2eb">[email protected]</span></a>, 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 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 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 April 
7, 1993 (58 FR 17948) that approved the VSC-24 design and added it to 
the list of NRC-approved cask designs in Sec. 72.214 as Certificate of 
Compliance Number (CoC No.) 1007.

Discussion

    On December 30, 1998, the certificate holder (PSNA) submitted an 
application to the NRC to amend CoC No. 1007 to permit a Part 72 
licensee to store burnable poison rod assemblies (BPRAs) with Babcock & 
Wilcox (B&W) 15 x 15 spent fuel assemblies in the VSC-24 system. A BPRA 
is a reactor core component that is inserted inside a fuel assembly 
during core refueling. BPRAs provide a means of controlling reactor 
power distribution and do not contain fissile material. No other 
changes to the VSC-24 system design were requested in this application. 
The NRC staff performed a detailed safety evaluation of the proposed 
CoC amendment request and found that the addition of the BPRAs to the 
B&W 15 x 15 fuel does not reduce the VSC-24 safety margin. In addition, 
the NRC staff has determined that the storage of BPRAs in the VSC-24 
does not pose any increased risk to public health and safety.
    This final rule revises the VSC-24 design listing in Sec. 72.214 by 
adding Amendment No. 1 to CoC No. 1007. The amendment consists of 
changes to the Technical Specifications (TS) for the VSC-24 design that 
will permit a Part 72 licensee to store BPRAs with B&W 15 x 15 spent 
fuel assemblies in a VSC-24 system. The particular TS that are changed 
are identified in the NRC staff's Safety Evaluation Report (SER) for 
Amendment No. 1.
    The title of the safety analysis report (SAR) will be changed from 
``Safety Analysis Report for the Ventilated Storage Cask System'' to 
``Final Safety Analysis Report for the Ventilated Storage Cask 
System.'' This action is being taken to ensure that the SAR title is 
consistent with the approach taken in new Sec. 72.248, recently 
approved by the Commission (64 FR 53582; October 4, 1999). 
Additionally, other minor, nontechnical, changes have been made to CoC 
No. 1007 to ensure consistency with the NRC's new standard format and 
content for CoCs.
    The NRC finds that the amended PSNA VSC-24 system, as designed and 
when fabricated and used under the conditions specified in the CoC, 
meets the requirements of Part 72, Subpart L. Thus, use of the PSNA 
VSC-24 system, as approved by the NRC, will continue to provide 
adequate protection of public health and safety and the environment. 
With this final rule, the NRC is approving the use of Amendment No. 1 
to the PSNA VSC-24 system under the general license provisions in 10 
CFR part 72, subpart K [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 30, 2000. Single copies of the 
CoC and SER are available for public inspection and/or copying for a 
fee at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), 
Washington, DC 20003-1527.

Summary of Public Comments on the Proposed Rule

    The NRC received one comment letter on the proposed rule from a 
member of the public. A copy of the comment letter is available for 
review in the NRC Public Document Room. The NRC's response to the 
issues raised by the commenter are discussed below.
    As stated in the proposed rule (64 FR 51270), the NRC considered 
this rulemaking to add Amendment No. 1 to the VSC-24 system design to 
10 CFR 72.214 to be a noncontroversial and routine action. Therefore, 
the NRC published a direct final rule concurrent with the proposed 
rule. The NRC indicated that if it received a

[[Page 24624]]

``significant adverse comment'' on the proposed rule, the NRC would 
publish a notice withdrawing the direct final rule and subsequently 
publish a final rule that addressed comments made on the proposed rule. 
The NRC believes that at least one of the issues raised by the 
commenter was a ``significant adverse comment.'' Therefore, the NRC 
published a notice withdrawing the direct final rule (64 FR 72019; 
December 23, 1999). This subsequent final rule addresses the issues 
raised by the commenter that were within the scope of the proposed 
rule, including the issue that was determined to be a ``significant 
adverse comment.''

Comments on Amendment No. 1 to the VSC-24 System

    The comments and responses have been grouped into five subject 
areas: general, weight considerations, radiation protection, design, 
and miscellaneous issues. The commenter provided specific comments on 
the draft CoC, the NRC staff's preliminary SER, and the TS. To the 
extent possible, all of the comments on a particular subject are 
grouped together. The listing of the VSC-24 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 minor correction to the 
CoC was made in response to one of the comments, but no changes were 
made to the TS or SER. A review of the comments and the NRC staff's 
responses follow:
A. General
    Comment A.1: The commenter stated that the proposed action should 
be called an ``amendment'' rather than a ``revision'' of the List of 
Approved Spent Fuel Storage Casks.
    Response: The NRC disagrees with the comment. The NRC is issuing 
Amendment No. 1 to CoC No. 1007 to allow for the storage of BPRAs in 
the VSC-24 system; therefore, changes are required to both the CoC and 
the TS. Because each approved Part 72 CoC is listed under 10 CFR 
72.214, the NRC is also required to revise the language in Sec. 72.214 
to reflect the approval and applicability of Amendment No. 1. 
Therefore, to promote clarity the NRC is using both the term 
``amendment to CoC No. 1007'' and ``revision to Sec. 72.214'' in this 
rule.
    Comment A.2: The commenter stated that the Federal Register should 
not call the action a ``Direct Final Rule.'' Streamlining the 
rulemaking process in this manner de-emphasizes safety concerns. The 
commenter also disagreed with NRC's characterization of the amendment 
as being ``noncontroversial and routine'' because this is the first 
amendment to a dry cask generic CoC and it raised many concerns.
    Response: The NRC believed no new technical issues would arise from 
the storage of BPRAs coincident with spent fuel, because: (1) BPRAs are 
safely used within spent fuel in a reactor; (2) operating conditions 
inside a reactor are harsher than storage conditions inside a VSC-24 
system; and (3) the NRC has previously reviewed the technical issues 
associated with the operation and storage of BPRAs in dry casks. 
Additionally, the proposed rule to amend the VSC-24 design was not the 
first amendment to a Part 72 cask design. A proposed rule to amend the 
Transnuclear West cask design (CoC No. 1004) was published in the 
Federal Register before this proposed rule was published (see 64 FR 
41050; July 29, 1999). Consequently, the NRC considered the storage of 
BPRAs with spent fuel to be a noncontroversial and routine action. The 
NRC continues to believe that the use of the direct final rule process 
was appropriate. Furthermore, the NRC also believes that the public's 
opportunity to comment on the proposed amendment to the VSC-24 design 
was not adversely impacted by the use of the direct final rule process. 
The withdrawal of the direct final rule--in response to receipt of a 
significant adverse comment--and publication of this final rule 
containing responses to all public comments demonstrate the NRC's 
commitment to provide the public the opportunity to comment on direct 
final rules.
    Comment A.3: The commenter objected ``. . . to use of new Sec. 
72.48 as it muddies the waters as to all change processes and just adds 
confusion as to how to keep documents current and to who is supposed to 
do what and be liable for what.''
    Response: This comment on the revised Sec. 72.48 is beyond the 
scope of this rule which is focused solely on whether to amend the VSC-
24 cask design. The revision to Sec. 72.48 was addressed in a separate 
rulemaking (64 FR 53582; October 4, 1999).
    Comment A.4: The commenter asked for the regulatory justification 
for allowing the amendment of a CoC and renaming the SAR to FSAR (Final 
SAR). The commenter also asked why the VSC-24 CoC was not amended to 
include a process for making amendments. The commenter questioned why 
the ``effective date'' of the initial certificate was not included in 
the CoC ``to begin with'' which would have precluded the need to amend 
the CoC. The commenter questioned whether the VSC-24 has received 
``special treatment'' since other CoCs (e.g., NUHOMS CoC Condition 9) 
have to be changed. The commenter stated that the SAR should not be 
renamed an FSAR because it is not a ``final'' document if changes are 
continually allowed. The commenter further noted that the language in 
the CoC does not refer to the ``final'' SAR, nor does it contain the 
date or revision number of the SAR. This is inconsistent with NRC's 
objective to change the SAR to an FSAR.
    Response: As stated in the proposed rule, the authority to approve 
a CoC for a spent fuel storage cask design is contained in Sections 
218(a) and 133 of the NWPA. Inherent with the NRC's authority under the 
NWPA to approve a spent fuel storage cask design is the authority to 
amend a previously approved cask design. The NRC regulations on 
amending a Part 72 cask design are contained in Secs. 72.244 and 72.246 
(see 64 FR 53582). With respect to the comment to add language to the 
CoC to include a process for amending the cask design, this is 
unnecessary because of the regulations contained in Secs. 72.244 and 
72.246. Furthermore, Condition No. 9 of CoC No. 1004 for the NUHOMS-24P 
and -52B cask design is intended to allow that certificate holder to 
make minor changes to the cask design without obtaining prior NRC 
approval. It was not intended to define a process for submitting an 
amendment to the certificate. Furthermore, this provision is not 
necessary for the VSC-24 CoC because the recent change to Sec. 72.48 
included certificate holders.
    The NRC has not previously added the effective date for a CoC to 
the list contained in Sec. 72.214 because the NRC believed the public 
and industry had adequate information on the effective date for a new 
CoC in the Federal Register notice that published the final rule 
[approving a specific cask design]. However, with the issuance of 
amendments, the NRC determined that it is necessary to identify the 
effective date of a CoC amendment because the CoC amendment may require 
certain changes, or may not permit certain actions, for casks that were 
put in service before the effective date of the amendment. The use of 
an effective date in Sec. 72.214 for both the amendment and the 
original CoC will improve clarity and ensure that both the industry and 
public understand the standard to which a specific cask has been 
manufactured or loaded. For example, an amendment to a hypothetical 
cask design that changes a material specification or a welding detail 
in a fuel support basket would not automatically be applied to casks 
that

[[Page 24625]]

have been already fabricated, loaded with spent fuel, and sealed 
because this would impose an unreasonable burden on the licensees who 
are using the cask. For the VSC-24 design, the effective date of the 
amendment is listed in this notice. A licensee can not use a VSC-24 
cask under the Part 72 general license to store BPRAs before the 
effective date of Amendment No. 1.
    The NRC recently added a new regulation in Sec. 72.248 on the 
submission and updating of the FSAR for each approved cask design (see 
64 FR 53582). Consequently, the term FSAR is used in both Sec. 72.214 
and the CoC to ensure consistency with the language contained in 
Sec. 72.248. The NRC agrees with the commenter that the word ``Final'' 
was inadvertently omitted from the proposed CoC. However, the proposed 
rule text did include the term ``final safety analysis report.'' 
Therefore, the final CoC has been corrected to include the term ``Final 
Safety Analysis Report.''
    The date of the FSAR and the revision number will be included in 
the document itself, as required by Sec. 72.248. However, the FSAR 
revision number and date of issuance will not be included in the CoC 
because Sec. 72.248 requires the certificate holder to update the FSAR 
every two years. Therefore, the NRC has chosen to omit this information 
from the CoC to prevent confusion between the rule language and the 
current FSAR. The NRC also notes that the certificate holder is 
required by Sec. 72.248 to submit an updated ``FSAR'' within 90 days of 
the issuance of this amendment to reflect any changes made to the CoC 
or TS. For this certificate holder, this process will convert the 
current SAR into an FSAR.
    Comment A.5: The commenter stated that the original rulemaking 
[approving the VSC-24 design] should have addressed the changes since 
the desire for these changes (e.g., inclusion of BPRAs) were well known 
at the time. However, there was a ``big push'' allowed by the NRC to 
get the VSC-24 certified ``as is,'' so this action was not taken.
    Response: The specific design features of the VSC-24 system are 
within the purview of the applicant. The NRC's review of a cask design 
is intended to ensure that the submitted cask design provides 
reasonable assurance that public health and safety and the environment 
will be protected. As such, the NRC's review is limited to the cask 
design submitted by the applicant and does not consider potential 
future optional features or different designs. Rather, changes to the 
design (e.g., to store BPRAs) are considered by the NRC in subsequent 
amendments to the cask design, if and when they are submitted by the 
certificate holder.
    Comment A.6: The commenter noted that the casks used at Palisades 
were built ``by exemption'' before the design was certified.
    Response: Comments on previously built VSC-24 casks [e.g., those 
used at the Palisades Nuclear Power Plant] that do not identify any 
issues relative to the storage of BPRAs are beyond the scope of the 
proposed rule.
    Comment A.7: The commenter has favored the action the NRC is now 
taking, i.e., to ensure that changes to the cask design be reflected in 
the various documents including the CoC.
    Response: No response necessary.
    Comment A.8: The commenter urged the NRC staff to think creatively 
about different problems including the effects of added weight and 
added dose. The NRC staff should also ``visualize'' the potential for 
accidents by considering the entire process, from removal of BPRAs to 
their storage in Yucca Mountain.
    Response: The NRC staff has evaluated the storage of BPRAs within 
B&W 15 x 15 Mark B fuel assemblies for storage in the VSC-24 system, 
including added weight and dose, and found it acceptable. Unloading of 
fuel containing BPRAs is not expected to be any more challenging than 
unloading of fuel without BPRAs. Use of the VSC-24 at Yucca Mountain is 
beyond the scope of this rule.
    Comment A.9: The commenter disagreed with the assertion that it 
will cost utilities more time and money to pursue exemptions to permit 
storage of BPRAs. In the long run, these site-specific actions will be 
more effective than ``one big generic exemption'' because they will 
result in fewer inspections and enforcements.
    Response: The NRC disagrees with the comment. NRC regulates 
licensees by compliance with the Federal regulations rather than 
exemptions to the regulations. Multiple exemption requests for the same 
issue are a cost and resource burden to both NRC and licensees. In this 
case, since multiple licensees are expected to request storage of 
BPRAs, this provision is more effectively addressed by rulemaking to 
amend the CoC and TS.
    Comment A.10: The commenter recommended that the utilities should 
remove the BPRAs and dispose of them in separate containers as low 
level waste. Using [spent fuel storage] casks to dispose of BPRAs is a 
waste of cask space and repository space that should be used for high 
level waste.
    Response: The NRC disagrees with the comment. BPRAs are reactor 
core components that are inserted into fuel assemblies during core 
refueling. A BPRA is physically located within a fuel assembly; 
therefore, no additional space is required to store or dispose of a 
spent fuel assembly with a BPRA also stored within the spent fuel 
assembly. Thus the presence of BPRAs will not affect the number of 
spent fuel assemblies that can be stored in a spent fuel storage cask.
    Comment A.11: The commenter asked why no other agencies (e.g., DOE, 
NWTRB) were apparently contacted regarding the environmental 
assessment. Further, the commenter is concerned about the potential 
cumulative effect on the environment of many ``insignificant'' 
incremental changes.
    Response: The agencies mentioned by the commenter are notified of 
the proposed rule in the same manner as the public. Therefore, the NRC 
did not believe it was necessary to specifically solicit their input. 
Furthermore, the Environmental Assessment covering the proposed rule, 
as well as the Finding of No Significant Impact, prepared and published 
for this rulemaking, fully comply with NRC's environmental regulations 
in 10 CFR part 51. The Commission's environmental regulations in Part 
51 implement the National Environmental Policy Act and are consistent 
with the guidelines of the Council on Environmental Quality.
    Comment A.12: The commenter questioned if the use of Regulatory 
Guide 3.61 is appropriate for this amendment request since both the CoC 
and the SAR are being amended. Also, the commenter questioned the 
designation of LAR 98-01 [License Amendment Request] as a 
``supplemental document,'' and asks for whom (SNC, ANO) it is 
supplemental. The commenter also asked how NRC will assure that LAR 98-
01 will be considered with Rev.0 of the SAR.
    Response: Regulatory Guide 3.61, ``Standard Format and Content for 
a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask'' is 
incorporated into NUREG-1536, ``Standard Review Plan for Dry Cask 
Storage Systems.'' The NRC staff used the guidance in NUREG-1536 for 
this amendment. LAR 98-01 was referred to as a supplemental document in 
the SER because it must be considered with information provided in 
Revision 0 of the SAR. Revision 0 of the SAR will be revised to 
incorporate the information in LAR 98-01 in the FSAR submitted by the 
applicant upon completion of this rulemaking.
    Comment A.13: The commenter disagreed that unloading procedures

[[Page 24626]]

should ``be left up to licensees to do after the casks are certified.'' 
These procedures should be put in the PDR because they are of great 
interest and concern to the public. The commenter is specifically 
concerned about changes needed in the unloading procedures to address 
BPRAs.
    Response: The NRC disagrees with the comment. NRC reviews a 
licensee's programs for compliance with the regulations by inspecting 
the adequacy and implementation of licensee procedures. Licensees are 
not required to submit implementing procedures to NRC on the public 
docket. Each licensee is required to review the adequacy of its 
procedures as a result of changes to the cask design or operational 
parameters. Further, BPRAs are integral to the fuel assembly and few, 
if any, changes should be needed in the unloading procedures.
    Comment A.14: The commenter generally criticized industry's 
(Nuclear Energy Institute and the plants) waste management policy. 
Industry is interested in moving the waste into casks as fast as 
possible and shipping it to Nevada for disposal. The commenter 
expressed concern about the amounts of waste that are being generated, 
the potential need for more repositories, and the lack of sound science 
to justify the storage and disposal of waste.
    Response: These comments are beyond the scope of this rule, which 
is focused solely on whether to amend the VSC-24 cask design.
    Comment A.15: The commenter stated that the NRC should always look 
out for workers and the public because it is NRC's job.
    Response: The NRC agrees with the comment. The NRC's highest 
priority is to protect the health and safety of both the public and 
workers at nuclear facilities.
    Comment A.16: The commenter was sympathetic with the NRC staff 
which has had to deal with problems caused by licensees, vendors, and 
subcontractors.
    Response: No response necessary.
    Comment A.17: The commenter stated that vendors are not responsible 
enough in QA procedures and that licensees should be responsible.
    Response: The NRC staff disagrees with the comment. The CoC holder 
is required to have and implement a Quality Assurance (QA) program 
approved by the NRC as part of the CoC issuance process. This QA 
program must meet the requirements of 10 CFR part 72, subpart G for 
cask design and fabrication activities. The cask user is ultimately 
responsible for ensuring that the fabricator's QA programs comply with 
10 CFR part 72, subpart G. NRC inspects licensee performance and takes 
enforcement actions as appropriate.
B. Weight Considerations
    Comment B.1: The commenter stated that the added weight from the 
BPRAs poses a big concern and should not be allowed.
    Response: The NRC disagrees with the comment. The overall weight of 
the Multi-Assembly Sealed Basket (MSB), Ventilated Concrete Cask (VCC), 
and MSB Transfer Cask (MTC) with the BPRAs included remains below the 
weight discussed in the SAR. Revision 0 of the SAR specifies the 
maximum design weight of the MSB as 118,630 lbs. The weight of the MSB 
with BPRAs is 6130 pounds less than this maximum weight.
    Comment B.2: The commenter stated that the safety margin is being 
reduced because the [VCC maximum] 80-inch lift height is being reduced 
to 60 inches. This reduction (due to increased stress in vertical drop) 
will be difficult to enforce and will create confusion and future 
problems.
    Response: The NRC disagrees with the comment. The maximum lifting 
height of the VCC outside of the spent fuel pool building was reduced 
from 80 to 60 inches because all supporting calculations in the SAR 
were based on a 60-inch drop height. Consequently, previous use of an 
80-inch drop height was inappropriate. Therefore, this reduction in the 
administratively controlled lift height will effectively increase the 
safety margin since the maximum lift height will now be lower.
    Comment B.3: The commenter asked whether the additional 60 lbs. 
more weight per assembly means that there will be an additional 24  x  
60 = 1440 lbs. per cask, which seems like a significant increment. The 
commenter further asked if this additional weight would have an effect 
on the pad, the loading area floor, the pool liner, transporter, sling, 
etc.
    Response: The addition of a BPRA to a B&W Mark B 15X15 fuel 
assembly increases the weight of the fuel assembly from 1516 lbs. to 
1576 lbs. For a cask fully loaded with 24 fuel assemblies containing 
BPRAs, the cask weight would increase by 1440 lbs., approximately 4 
percent of the cask weight. This increase in weight was found by the 
NRC to be acceptable for complying with the normal use and accident 
conditions evaluated under the provisions of Part 72. Furthermore, each 
licensee using a VSC-24 cask is required by Secs. 50.59, 72.48, and 
72.212 to evaluate whether the additional weight of a cask will have an 
unacceptable adverse effect on structures, systems, or components, such 
as the ISFSI pad, the loading floor area, or the pool liner. The cask 
cannot be used if the licensee identifies an unacceptable adverse 
impact. [See also response to Comment No. B.1.]
    Comment B.4: The commenter stated that the proposed amendment 
reduces the VSC-24 safety margin and increases the risk to public and 
worker health and safety. The doses are larger, stresses are more, drop 
height is reduced, shielding on MTC is reduced, and weight is 
increased.
    Response: The NRC disagrees in part with the comment. The reduction 
in drop height for a loaded VCC increases the safety margin by ensuring 
that the VCC is not able to fall through more than 60 inches (rather 
than 80 inches) in the vertical orientation. Although the stresses 
associated with a vertical drop of the VCC increase 6 percent, these 
stresses comply with the ASME Code limits. Regarding the MTC, the 
shielding in the bottom doors of the MTC was reduced to compensate for 
the increased weight of the loaded MSB. The MTC weight reduction was 
required to maintain the lift load within a predetermined crane lift 
load capacity. Issues related to increased dose are discussed in 
response to Comment No. C.4.
C. Radiation Protection
    Comment C.1: The commenter stated that it is not acceptable to have 
an increase of 7.5 percent in offsite and direct skyshine dose rate to 
the public, even if the resulting doses are within the limits. The 
commenter questioned if the combined dose from ``a full cask array'' or 
``several full cask arrays'' would be acceptable to the public or to 
workers. For workers, in particular, the NRC needs to take into account 
the future cumulative effect of years of worker exposure resulting from 
inspections of the casks. The commenter disagreed that the projected 13 
percent increase in ``potential cask dose rates'' does not constitute 
an increased risk to health and safety. The commenter noted that the 
highest projected dose is at ``top center'' of the cask, and would like 
to know, since dosimeters are not located there, what the real dose 
would be (from a full cask array right above the casks on the pad) for 
a surveillance worker who needs to check outlets at the top of the 
casks.
    Response: The NRC disagrees with the comment. The increase in 
offsite dose at 1500 feet from an array of 68 VSC-24 casks with 5-year 
cooled spent fuel represents a conservative bounding estimate of the 
effect of BPRAs on offsite

[[Page 24627]]

doses. The actual offsite dose to the public from an Independent Spent 
Fuel Storage Installation (ISFSI) is affected by many factors, 
including the number of casks, specific placement of fuel assemblies 
within each cask, cask positioning, if the fuel is cooled beyond 5 
years, and the presence of natural shielding features such as earthen 
berms and buildings that are not credited in design safety offsite dose 
calculations. Each ISFSI licensee is required to demonstrate that 
offsite public annual whole body doses remain below the Sec. 72.104 
limit of 25 mrem/year.
    The NRC determined that the addition of BPRAs will result in an 
increase of approximately 7.5 percent in the calculated offsite direct 
and skyshine dose rate to the public as calculated and presented in 
Revision 0 of the SAR. The potential annual dose to the public at 1500 
feet from an array of 68 VSC-24s loaded with 5-year cooled spent 
nuclear fuel would increase from 0.039 mSv/year to 0.042 mSv/year (3.9 
mrem/year to 4.2 mrem/year), which remains well below the 0.25 mSv/year 
(25 mrem/year) limit in Sec. 72.104. The estimated annual occupational 
exposure for routine activities such as visual surveillance of cask air 
inlets/outlets and radiation protection surveys on a cask filled to 
design capacity would be 7 x 10-6 person-Sv/year/cask (0.0007 person-
rem/year/cask.) Based on these expected occupational activities, the 
NRC has reasonable assurance that individual exposures will be below 
the annual occupational limit of 0.05 Sv (5 rem) specified in 
Sec. 20.1201.
    Comment C.2: The commenter is concerned about where the dosimeters 
are placed in relation to the height of the casks. They should be 
placed at the ``top height'' where the dose is expected to be the 
highest. If the dosimeters are not placed in this position, the 
commenter would like an explanation.
    Response: ISFSI licensees are required by Sec. 72.104(a) to ensure 
that dose rates do not exceed 0.25 mSv/year (25 mrem/year) at the 
controlled area boundary. ISFSI licensees typically place radiation 
monitoring devices (dosimeters) at various locations around the ISFSI 
perimeter fence at approximately the chest height of an average worker 
standing at the ISFSI perimeter fence. This dosimetry is used to 
monitor the actual dose from the ISFSI and to determine the dose at the 
controlled area boundary. A dosimeter placed at the top of a cask would 
not provide useful information for the determination of dose to a 
member of the public or a worker. A worker that is within the ISFSI 
perimeter fence and performing an activity at the top of a cask would 
be subject to the licensees' 10 CFR part 20 Radiation Protection 
Program requirements, including controls to limit exposure and the 
placement (i.e., wearing) of personal dosimetry. [See also response to 
Comment No. C.1.]
    Comment C.3: The commenter questioned why the maximum increase of 
cask dose rate is evaluated at the air inlets rather than at the 
outlets and top of the cask where the highest dose rate is expected. 
Also, the commenter asked about the increase in reflected radiation 
``from cask to cask in full cask array,'' and if it is still correct to 
assume a center-to-center distance of 15 ft.
    Response: The maximum dose rate due to the inclusion of B&W 15x15 
BPRAs in the VSC-24 was calculated for all locations on and around the 
VSC-24 storage cask, including the air outlets and the top of the cask. 
Although the dose rates also increased at the air outlets and top of 
the cask, the SER specifically delineated the increase in dose rate at 
the air inlets because this was the largest percent increase and is a 
significant contributor to worker doses during required daily air 
inlet/outlet surveillance of the VSC-24. The NRC determined that the 
increase in reflected radiation from cask-to-cask in a full 68 cask 
array was insignificant and that the existing center-to-center cask 
distance of 15 feet was acceptable.
    Comment C.4: The commenter stated that to accommodate the added 
weight, changes have been made that reduce the safety margin and are 
inconsistent with ALARA. In particular, by reducing the MTC shielding, 
the potential occupational dose rate increases from 300 to 1932 mrem 
per hour. This should not be allowed because of the impact on workers. 
The commenter also questioned NRC's statement that workers are ``not 
expected'' to be in the area where they could receive an occupational 
dose of 1932 mrem/hr.
    Response: The NRC disagrees in part with the comment. Although 
there is some increase in the potential dose to workers, the likelihood 
of such an exposure is very low. Operations for loading the MSB, 
placing it into the MTC, and loading the MSB into the VCC from the MTC 
do not involve the presence of workers in or around the bottom of the 
MTC. Under the requirements for movement of heavy loads such as the 
MTC, personnel are prohibited from the area directly below the load 
when it is lifted or being moved. ALARA (``as low as reasonably 
achievable'') practices implemented by licensees include sound 
radiation protection principles and procedures for monitoring actual 
dose rates, using additional temporary shielding (when appropriate), 
and restricting the location and time of workers in various radiation 
fields to minimize doses.
    Comment C.5: The commenter asked how BPRAs in the cask and worker 
dose are affected by the fact that drain down is necessitated before UT 
[ultrasonic testing] of structural welds is finished.
    Response: Drain down of the cask has no effect on the BPRAs. [See 
also Comment No. D.4.] The issue of the effect of drain down on worker 
dose during the performance of UT on a structured weld is beyond the 
scope of the proposed rule.

D. Materials

    Comment D.1: The commenter stated that a big concern is materials' 
interactions. Consequently, it is important to know what materials are 
present in the BPRAs and what interactions (chemical and physical) they 
could have with the materials in a VSC-24. In particular, the commenter 
would like to know what coating will be used in the sleeves holding the 
BPRA assemblies, the proximity of the coating to the materials in the 
BPRA, and the dimensions and density of the BPRA material versus 
regular fuel rods. The commenter asked for a full description of all 
the materials that comprise a BPRA because such a description does not 
exist in the documentation reviewed.
    Response: BPRAs are composed of stainless steel hardware supporting 
sealed zircalloy rods containing aluminum oxide and boron carbide 
pellets. During normal nuclear power plant operation, some spent fuel 
assemblies operate with BPRAs inserted into their usually empty guide 
tubes. There are no coatings used in the zircalloy guide tubes of the 
B&W Mark B 15x15 fuel assemblies that would interact with the BPRA. No 
adverse interactions between the materials in a BPRA and the VSC-24 are 
expected. Description of a fuel assembly and a BPRA, including relevant 
dimensions, is contained within the SAR and its reference documents. 
These documents are available in the PDR.
    Comment D.2: The commenter questioned if ``all reactor BPRAs'' are 
the same (materials, size, weight, susceptibility to corrosion, cracks, 
pinhole leaks, etc.) and if they should be treated genericlly. Further, 
the commenter asked what criteria (i.e., TS) have been established for 
determining which BPRAs are to be allowed in the cask. This is based on 
concern over the storage of BPRAs that might be produced in the future. 
The commenter objected to the decision to accept BPRAs with cladding 
failures because of

[[Page 24628]]

concerns over depressurization including deterioration, collapse and 
``getting stuck,'' crumbling and clogging of spaces in other sleeves, 
reactions of decayed BPRAs with other cask materials (coatings).
    Response: The only BPRAs approved for storage under this rulemaking 
are those to be stored in B&W Mark B 15x15 fuel assemblies. BPRAs with 
cladding failures were analyzed and determined to be acceptable for 
loading in the VSC-24. A failed BPRA loaded in the VSC-24 would be 
depressurized and actually present a lower MSB accident pressure than 
that of an intact BPRA. Any release from a failed BPRA would not have 
an adverse effect on the internals of the MSB or the fuel assemblies 
stored in the MSB. [See also Comment Nos. D.1 and D.3.]
    Comment D.3: The commenter expressed concern about the possibility 
of leaks from a BPRA that is inserted inside a fuel assembly. Since 
BPRAs cannot be observed, the commenter wondered how leaks can be 
detected, how they react to vacuum drying of fuel rods, and if 
retainment of water (causing added weight and possible corrosion) could 
be a problem.
    Response: The NRC evaluated the postulated accident assuming all 24 
BPRAs in a VSC-24 MSB failed. This analysis showed that the maximum MSB 
pressure due to the simultaneous failure of all 24 BPRAs and all 24 
stored spent nuclear fuel assemblies resulted in MSB stresses that 
remained below the American Society of Mechanical Engineers (ASME) Code 
allowable values and therefore, would not affect the MSB confinement 
boundary. A failed BPRA would release helium gas, which is already 
present, to the MSB internals. A BPRA would not present more problems 
in vacuum drying the MSB than the spent fuel assembly itself.
    Comment D.4: The commenter asked how BPRAs change as they ``dry 
out'' and questioned whether any tests have been conducted regarding 
this issue. For example, could the materials lose their structural 
integrity which would cause a problem in unloading or shipping. This 
could be compounded by the effects of heat, radiation, and chemical 
reactions (e.g. with ``pool water chemicals'').
    Response: Vacuum drying will not reduce the structural integrity of 
a BPRA. The BPRA will continue to maintain the same structural 
integrity as the fuel assembly in which it is secured.
    Comment D.5: The commenter recommended that the next amendment 
should prohibit the use of ``flammable plastic tube'' and ``duct tape'' 
to prevent the release of hydrogen. In addition, the commenter 
recommended additional criteria that requires coatings that do not 
create hydrogen and stipulated the use of stainless steel. The 
commenter questioned how BPRAs could be affected by hydrogen 
generation.
    Response: Comments on future amendments are beyond the scope of the 
proposed rule. [See Comment No. D.1 on material composition of BPRAs.] 
Regarding the question of hydrogen generation, the NRC staff determined 
that the potential presence of hydrogen gas during VSC-24 loading 
activities has an insignificant effect on the BPRAs.
    Comment D.6: The commenter recommended the use of the term ``carbon 
steel,'' rather than ``steel'' when it is appropriate.
    Response: If there were different types of steel used in the VSC-24 
design, the NRC would agree with the comment. The NRC typically 
specifies the variety or grade of a steel when presenting information 
if there is a potential for misunderstanding. However, all of the steel 
used in the VSC-24 design is of the carbon steel variety. [See also 
Comment No. D.1.]
E. Design
    Comment E.1: The commenter stated that the amendment should be a 
site-specific design request and technical evaluation from Entergy for 
the Arkansas Nuclear One (ANO) ISFSI instead of a generic amendment. 
The commenter further stated that Entergy should be liable and 
responsible for future problems, but that apparently BNF [British 
Nuclear Fuel Limited] wants to be responsible. Although the NWPA calls 
for approval of generic cask designs ``to the maximum extent 
practicable,'' the commenter believes the current action ``calls for 
site-specific approval at each plant and is not practicable to be a 
generic amendment'' ``A generic cask CoC should not have to be amended 
to suit the site specific need of one licensee.'' In particular, the 
commenter is critical of the actions of ANO with respect to their use 
of the change process in Sec. 72.48, and stated that ANO should have 
gotten [applied for] a site specific license ``right from the 
beginning.''
    Response: The NRC does not agree that a site-specific approval is 
needed to store BPRAs in the VSC-24 cask design. The VSC-24 cask design 
was approved in a final rule (58 FR 17948; April 7, 1993) under the 
NRC's Part 72 regulations that implement Sections 218(a) and 133 of the 
NWPA. Section 218(a) directed the NRC to approve one or more spent fuel 
dry storage technologies for use at civilian nuclear power reactors 
``without, to the maximum extent practicable, the need for additional 
site-specific approvals by the Commission.'' Therefore, the NRC 
believes that the VSC-24 cask design, and any amendments to the cask 
design (i.e., storage of BPRAs), may be used by all Part 72 general 
licensees without obtaining an additional NRC site-specific approval. 
[See also response to Comment No. A.5.]
    The NRC understands that ANO is expected to be the first Part 72 
general licensee to utilize the provisions of Amendment No. 1 to store 
BPRAs in a VSC-24 cask. However, irrespective of which Part 72 general 
licensees may wish to use this provision to store BPRAs, the 
certificate holder is ultimately responsible for the cask design and 
for submitting any applications to amend the cask design. In submitting 
such an application, the certificate holder must demonstrate to the 
NRC's satisfaction that the proposed amendment will not adversely 
affect public health and safety and the environment.
    Comment E.2: The commenter questioned how the length of the B&W 
15x15 assemblies fit in with BPRAs. In particular, if the cask design 
and procedures must accommodate a difference in length, what are the 
ramifications? The commenter also questioned if there are any problems 
in unloading BPRAs and stated that, perhaps, there should be ``tests 
for BPRAs before the first loading at the plant.''
    Response: A BPRA is secured [located] within a fuel assembly so no 
additional space is required in a VSC-24 cask to store a spent fuel 
assembly with a BPRA. Consequently, handling operations such as loading 
or unloading of a spent fuel assembly containing a BPRA are not 
expected to present any more difficulty than for a spent fuel assembly 
without a BPRA. Licensee users are required to perform dry runs and 
training exercises of the cask loading and unloading activities before 
performing the actual operation.
    Comment E.3: The commenter recommended that the information on 
hydraulic roller skids and skid openings be removed [from the cask 
design] since nobody uses them.
    Response: The NRC disagrees with the comment. The applicant did not 
request an amendment to the information on the hydraulic roller skids 
and skid openings; therefore, this comment is beyond the scope of this 
rule and the information was not revised in this CoC amendment.
    Comment E.4: The commenter asked whether the basket supports have 
been

[[Page 24629]]

evaluated (over time and when dry) for extra weight, size, and stress.
    Response: The NRC reviewed the structural adequacy of the MSB 
including basket supports for the additional weight of the BPRAs and 
found that all stresses were less than the ASME Code allowable stress 
limits.
    Comment E.5: The commenter asked if the BPRAs can be drained 
effectively and if tests have been done to confirm this.
    Response: Vacuum drying the BPRA is not expected to present any 
more difficulty in vacuum drying the MSB than for the spent fuel 
assembly itself. The geometrical features of BPRAs that could retain 
water are equivalent to or less complex than the fuel assemblies 
themselves.
F. Miscellaneous
    Comment F.1: The commenter asked why the CoC, EA [Environmental 
Assessment], and SER inconsistently reference the certificate holder. 
Is it SNC or PSNA?
    Response: The entity that requested the CoC amendment was Sierra 
Nuclear Corporation (SNC). SNC is owned by Pacific Sierra Nuclear 
Associates (PSNA). PSNA is the registered owner of the VSC-24 design. 
The documents have been modified for consistency.
    Comment F.2: The commenter asked how a plant reports what is placed 
in each cask because this documentation may be crucial in the future.
    Response: The VSC-24 users are required to document pertinent 
information on each fuel assembly stored in the cask (including whether 
it contains a BPRA) under Secs. 72.76, 72.78, and 72.212(b)(8)(i). This 
information is required to be maintained by the licensee user until 
termination of the license.
    Comment F.3: The commenter asked about the process for notifying 
manufacturers, users, and potential users of problems in storing BPRAs 
in casks. This is important so that the same mistakes are not repeated. 
The commenter stated that the CoC holder should be held liable for not 
informing users of potential concerns.
    Response: Certificate holders are required by the recently revised 
Sec. 72.242(d) to notify the NRC of ``a design or fabrication 
deficiency, for any spent fuel storage cask which has been delivered to 
a licensee, when the design or fabrication deficiency affects the 
ability of structures, systems, and components important to safety to 
perform their intended safety function.'' (64 FR 56114; October 15, 
1999). The NRC expects that the certificate holder will provide a copy 
of this report to any affected licensees. If such a report is received 
by the NRC, the NRC can verify through inspections that all affected 
cask users are aware of the information.
    Comment F.4: The commenter stated that the term ``double-closure'' 
weld, used in the EA, is not correct. In the commenter's opinion, it is 
not possible to count the shield lid as a closure weld because it is 
not UT tested. The CoC should be amended to say that there is only one 
closure weld (i.e., the structural lid weld).
    Response: The NRC disagrees with the comment. VSC-24 cask users are 
required to perform nondestructive examination of both the shield lid 
to MSB shell weld and the structural lid to MSB shell weld. Both of 
these welds are considered closure welds. The CoC and TS require cask 
users to perform liquid penetrant examination of both of these welds.
    Comment F.5: The commenter stated that the sabotage evaluations for 
dry casks are outdated and need to be redone because of the increased 
threat of terrorist activity.
    Response: This comment is beyond the scope of the current rule.
    Comment F.6: The commenter asked why the name of the valve 
manufacturer has now been deleted from the amendment and believed this 
should have been done long ago.
    Response: The NRC agrees with the comment. The name of the valve 
manufacturer is not required for operational activities of the VSC-24 
and has been deleted.
    Comment F.7: The commenter questioned whether there will be 
specific ``checks,'' documented in procedures, for boron concentration 
to eliminate potential confusion if a plant uses VSC casks to store 
both BPRAs and non-BPRAs.
    Response: The storage of BPRAs in the VSC-24 cask does not require 
a change in the boron concentration of the water inside the MSB. 
Technical Specification 1.2.6 controls the boron concentration inside 
the MSB during loading and unloading operations.
    Comment F.8: The commenter stated that ``dry runs don't seem to be 
effective in troubleshooting,'' and asked what other actions need to be 
taken.
    Response: Changes to the requirement to conduct dry runs of cask 
operations are beyond the scope of the proposed rule.
    Comment F.9: The commenter asked what ``wet helium'' is and how 
tests can be conducted for it.
    Response: The NRC does not recognize the term ``wet helium,'' as 
used by the commenter; consequently, this comment is not addressed.

Summary of Final Revisions

Section 72.214  List of Approved Spent Fuel Storage Casks

    Certificate No. 1007 is revised by adding the effective date of the 
initial certificate, the effective date of Amendment Number 1, and 
revising the title of the SAR submitted by PSNA to ``Final Safety 
Analysis Report for the Ventilated Storage Cask System.''

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, 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 Commission's 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 PSNA VSC-24 CoC, and accordingly revises the VSC-
24 system listing within the list of approved spent fuel storage casks 
in Sec. 72.214. Power reactor licensees can use these approved casks to 
store spent fuel at reactor sites without additional site-specific 
approvals from the Commission. The amendment modifies the present cask 
system design to permit a Part 72 licensee to store BPRAs in the VSC-24 
system design along with the spent fuel. The environmental assessment 
and finding of no significant impact on which this determination is 
based are available for inspection at the NRC Public Document Room, 
2120 L Street NW. (Lower Level), Washington, DC. Single copies of the 
environmental assessment and finding of no significant impact are 
available from Richard

[[Page 24630]]

Milstein, Office of Nuclear Material Safety and Safeguards, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 
415-8149, email <a href="/cdn-cgi/l/email-protection#99ebf0f4d9f7ebfab7fef6ef"><span class="__cf_email__" data-cfemail="6012090d200e12034e070f16">[email protected]</span></a>.

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

    If a means used to impose an information collection does not 
display a currently valid Office of Management and Budget control 
number, the NRC may not conduct or sponsor, and a person is not 
required to respond to, the information collection.

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 would revise the PSNA VSC-24 
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.

Regulatory Analysis

    On July 18, 1990 (55 FR 29181), the NRC issued an amendment to 10 
CFR part 72. The amendment provided for the storage of spent nuclear 
fuel in cask systems with the designs approved by the NRC under a 
general license. Any nuclear power reactor licensee can use cask 
systems with designs approved by the NRC to store spent nuclear fuel if 
it notifies the NRC in advance, the spent fuel is stored under the 
conditions specified in the cask's CoC, and the conditions of the 
general license are met. A list of NRC-approved cask designs is 
contained in Sec. 72.214. On April 7, 1993 (58 FR 17948), the NRC 
issued an amendment to Part 72 that approved the VSC-24 design, added 
it to the list of NRC-approved cask designs in Sec. 72.214, and issued 
CoC No. 1007. On December 30, 1998, the certificate holder (PSNA), 
submitted an application to the NRC to amend CoC No. 1007 to permit a 
Part 72 licensee to store BPRAs with B&W 15x15 spent fuel assemblies in 
the VSC-24 system.
    This final rule will permit the storage of certain reactor core 
components (i.e., BPRAs) that do not contain fissile material in the 
VSC-24 system. The alternative to this action is to withhold approval 
of this amended cask system design and issue an exemption to each 
general license that proposes to use the casks to store BPRAs. This 
alternative would cost both the NRC and the utilities more time and 
money because each utility would have to submit a request for an 
exemption and NRC would have to review each request.
    Approval of the final rule will eliminate the problem described 
above and is consistent with previous Commission actions. Further, the 
final rule will have no adverse effect on public health and safety. 
This final rule has no significant identifiable impact on or benefit to 
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 Commission'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.

Small Business Regulatory Enforcement Fairness Act

    Under 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.

Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
Commission certifies that this rule will not, if promulgated, 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 PSNA. 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 
in the backfit rule. Therefore, a backfit analysis is not required.

List of Subjects in 10 CFR Part 72

    Administrative practice and procedure, Hazardous waste, Nuclear 
materials, Occupational safety and health, Penalties, Radiation 
protection, Reporting and recordkeeping requirements, Security 
measures, Spent fuel, and 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 AND HIGH-LEVEL RADIOACTIVE 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. 10d-
48b, sec. 7902, 10b Stat. 31b3 (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).
    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. Section 72.214, Certificate of Compliance No. 1007 is revised to 
read as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1007.
    Initial Certificate Effective Date: May 7, 1993.

[[Page 24631]]

    Amendment Number 1 Effective Date: May 30, 2000.
    SAR Submitted by: Pacific Sierra Nuclear Associates.
    SAR Title: Final Safety Analysis Report for the Ventilated Storage 
Cask System.
    Docket Number: 72-1007.
    Certificate Expiration Date: May 7, 2013.
    Model Number: VSC-24.
* * * * *

    Dated at Rockville, Maryland, this 12th day of April, 2000.

For the Nuclear Regulatory Commission.
Frank J. Miraglia, Jr.,
Acting Executive Director for Operations.
[FR Doc. 00-10392 Filed 4-26-00; 8:45 am]
BILLING CODE 7590-01-P


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