This section of the FEDERAL REGISTER

contains regulatory documents having general

applicability and legal effect, most of which

are keyed to and codified in the Code of

Federal Regulations, which is published under

50 titles pursuant to 44 U.S.C. 1510.

The Code of Federal Regulations is sold by

the Superintendent of Documents. Prices of

new books are listed in the first FEDERAL

REGISTER issue of each week.

Rules and Regulations Federal Register

24623

Vol. 65, No. 82

Thursday, April 27, 2000

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 rim@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 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 § 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) 15×15

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 15×15 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 § 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

15×15 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 § 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

24624 Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules and Regulations

‘‘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

§ 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 § 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 § 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 § 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 §§ 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

§§ 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

§ 72.48 included certificate holders.

The NRC has not previously added

the effective date for a CoC to the list

contained in § 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 § 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

Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules and Regulations 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 § 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 § 72.214 and the CoC to ensure

consistency with the language contained

in § 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 § 72.248.

However, the FSAR revision number

and date of issuance will not be

included in the CoC because § 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 § 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 15×15 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 sitespecific 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

24626 Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules and Regulations

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 × 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 §§ 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

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

§ 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 § 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

7×10¥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 § 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 § 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

24628 Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules and Regulations

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 sitespecific 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 § 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 sitespecific 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

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

§ 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 § 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

24630 Federal Register / Vol. 65, No. 82 / Thursday, April 27, 2000 / Rules and Regulations

Milstein, Office of Nuclear Material

Safety and Safeguards, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, telephone (301) 415–8149,

email rim@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

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 § 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 § 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 § 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:

§ 72.214 List of approved spent fuel

storage casks.

* * * * *

Certificate Number: 1007.

Initial Certificate Effective Date: May

7, 1993.

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

DEPARTMENT OF TRANSPORTATION

Federal Aviation Administration

14 CFR Part 39

[Docket No. 99–NM–56–AD; Amendment

39–11700; AD 2000–08–14]

RIN 2120–AA64

Airworthiness Directives; Boeing

Model 747 Series Airplanes

AGENCY: Federal Aviation

Administration, DOT.

ACTION: Final rule.

SUMMARY: This amendment adopts a

new airworthiness directive (AD),

applicable to certain Boeing Model 747

series airplanes, that requires repetitive

inspections or checks to detect broken

H–11 steel bolts at the wing rear spar

side-of-body on the lower chord splice

plate and kick fitting; and corrective

actions, if necessary. This AD also

requires eventual replacement of the

existing bolts with new Inconel bolts,

which constitutes terminating action for

the repetitive inspections. This

amendment is prompted by a report of

broken bolts at the wing rear spar sideof-body on the lower chord splice plate.

The actions specified by this AD are

intended to prevent cracking of the bolts

due to stress corrosion, which could

result in reduced structural integrity of

the wing-to-body joint structure.

DATES: Effective June 1, 2000.

The incorporation by reference of

certain publications listed in the

regulations is approved by the Director

of the Federal Register as of June 1,

2000.

ADDRESSES: The service information

referenced in this AD may be obtained

from Boeing Commercial Airplane

Group, P. O. Box 3707, Seattle,

Washington 98124–2207. This

information may be examined at the

Federal Aviation Administration (FAA),

Transport Airplane Directorate, Rules

Docket, 1601 Lind Avenue, SW.,

Renton, Washington; or at the Office of

the Federal Register, 800 North Capitol

Street, NW., suite 700, Washington, DC.

FOR FURTHER INFORMATION CONTACT:

Tamara L. Anderson, Aerospace

Engineer, Airframe Branch, ANM–120S,

FAA, Transport Airplane Directorate,

Seattle Aircraft Certification Office,

1601 Lind Avenue, SW., Renton,

Washington 98055–4056; telephone

(425) 227–2771; fax (425) 227–1181.

SUPPLEMENTARY INFORMATION: A

proposal to amend part 39 of the Federal

Aviation Regulations (14 CFR part 39) to

include an airworthiness directive (AD)

that is applicable to certain Boeing

Model 747 series airplanes was

published in the Federal Register on

September 2, 1999 (64 FR 48120). That

action proposed to require repetitive

inspections or checks to detect broken

H–11 steel bolts at the wing rear spar

side-of-body on the lower chord splice

plate and kick fitting; and corrective

actions, if necessary.

Comments

Interested persons have been afforded

an opportunity to participate in the

making of this amendment. Due

consideration has been given to the

comments received.

Request To Reference Revised Service

Bulletin

One commenter requests that the FAA

revise paragraph (d) of the notice of

proposed rulemaking (NPRM) to allow

accomplishment of the terminating

action in accordance with either the

original issue of Boeing Alert Service

Bulletin 747–57A2309, dated February

25, 1999 (which is referenced as the

appropriate source of service

information in the NPRM), or Boeing

Service Bulletin 747–57A2309, Revision

1, dated December 22, 1999.

The FAA concurs with the

commenter’s request. Since the issuance

of the NPRM, the FAA has reviewed and

approved Boeing Service Bulletin 747–

57A2309, Revision 1. The procedures

specified in that service bulletin are

substantially similar to those in the

original issue. Among other things,

Revision 1 of the service bulletin

references kits with cadmium-plated

nuts instead of passivated nuts and

revises jacking instructions. The FAA

finds that use of either the original issue

or Revision 1 of the service bulletin is

acceptable for compliance with all

actions specified in this AD. Therefore,

the FAA is revising paragraphs (a), (b),

and (d), of this final rule to reference

Revision 1, as well as the original issue

of the service bulletin.

In addition, no new airplanes are

added to the effectivity listing in

Revision 1 of the service bulletin, but

the effectivity listing does show changes

in airplane operators. Therefore, for

clarity, the applicability statement of

this final rule has been revised to refer

to airplanes listed in Revision 1 instead

of the original issue of the service

bulletin.

Request To Revise Paragraph (c)

One commenter requests that

paragraph (c) of the proposed AD be

revised to refer not only to paragraph

(b), as specified in the proposal, but also

to paragraph (d)(1). The commenter

points out that paragraph (c) of the

proposal only refers to cracks found

during accomplishment of corrective

action required by paragraph (b), but

paragraph (d)(1) of the proposal also

refers to accomplishment of necessary

corrective actions in accordance with

paragraph (c). The commenter

recommends that paragraph (c) of this

AD be revised to read as follows: ‘‘If any

crack is detected during any corrective

action required by paragraph (b) of this

AD or during terminating action

required by paragraph (d)(1) of this AD.

* * *’’ The FAA concurs with the

commenter’s request, and has revised

paragraph (c) of this final rule

accordingly.

Request To Clarify Number of Fasteners

One commenter requests that the FAA

revise paragraphs (d)(1) and (d)(2) of the

proposed rule to accurately state the

correct number of fasteners for all

groups of airplanes listed in the service

bulletin. The commenter points out that

airplanes in Group 2 have only four

high strength H–11 steel bolts common

to the rear spar lower chord splice plate,

while airplanes in Groups 1, 3, 4, and

5 have eight high strength H–11 steel

bolts common to the rear spar lower

chord splice plate.

The same commenter requests that the

FAA revise paragraphs (d)(1) and (d)(2)

of the proposed AD to also explicitly

state that the number of high strength

H–11 steel bolts listed in those

paragraphs of the AD are the numbers

for each side. The commenter states that

this change is necessary for clarity.

The FAA partially concurs with the

commenter’s request. The FAA

acknowledges that airplanes in Group 2

have only four high strength H–11 steel

bolts common to the rear spar lower

chord splice plate, while airplanes in