Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24855

5. In § 46.13, paragraphs (a)(2) and

(a)(5) are revised to read as follows:

§ 46.13 Address, ownership, changes in

trade name, changes in number of

branches, changes in members of

partnership, and bankruptcy.

* * * * *

(a) * * *

(2) Any change in officers, directors,

members, managers, holders of more

than 10 percent of the outstanding stock

in a corporation, with the percentage of

stock held by such person, and holders

of more than 10 percent of the

ownership stake in a limited liability

company, and the percentage of

ownership in the company held by each

such person;

* * * * *

(5) When the licensee, or if the

licensee is a partnership, any partner is

subject to proceedings under the

bankruptcy laws. A new license is

required in case of a change in the

ownership of a firm, the addition or

withdrawal of partners in a partnership,

or in case business is conducted under

a different corporate charter, or in case

a limited liability company conducts

business under different articles or

organization from those under which

the license was originally issued.

* * * * *

Dated: April 21, 2000.

Robert C. Keeney,

Deputy Administrator, Fruit and Vegetable

Programs.

[FR Doc. 00–10481 Filed 4–27–00; 8:45 am]

BILLING CODE 3410–02–P

NUCLEAR REGULATORY

COMMISSION

10 CFR Part 72

RIN 3150–AG 30

List of Approved Spent Fuel Storage

Casks: TN–68 Addition

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is amending its

regulations to add the Transnuclear TN–

68 cask system to the list of approved

spent fuel storage casks. This

amendment allows holders of power

reactor operating licenses to store spent

fuel in the Transnuclear TN–68 cask

system under a general license.

DATES: The final rule is effective May

30, 2000.

FOR FURTHER INFORMATION CONTACT:

Gordon Gundersen, telephone (301)

415–6195, e-mail, GEG1@nrc.gov of the

Office of Nuclear Material Safety and

Safeguards, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001.

SUPPLEMENTARY INFORMATION:

Background

Section 218(a) of the Nuclear Waste

Policy Act of 1982, as amended

(NWPA), requires that ‘‘[t]he Secretary

[of Energy] shall establish a

demonstration program for the dry

storage of spent nuclear fuel at civilian

nuclear power reactor sites, with the

objective of establishing one or more

technologies 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, publishing a final rule,

in 10 CFR part 72 entitled ‘‘General

License for Storage of Spent Fuel at

Power Reactor Sites’’ (55 FR 29181, July

18, 1990). This rule also established a

new Subpart L within 10 CFR part 72

entitled, ‘‘Approval of Spent Fuel

Storage Casks’’ containing procedures

and criteria for obtaining NRC approval

of dry storage cask designs.

Discussion

This rule will add the Transnuclear

TN–68 cask system to the list of NRC

approved casks for spent fuel storage in

10 CFR 72.214. Following the

procedures specified in 10 CFR 72.230

of Subpart L, Transnuclear submitted an

application for NRC approval with the

Safety Analysis Report (SAR) entitled

‘‘Final Safety Analysis Report for the

TN–68 Dry Storage Cask,’’ dated January

23, 1998. The NRC evaluated the

Transnuclear submittal and issued a

preliminary Safety Evaluation Report

(SER) and proposed Certificate of

Compliance (CoC) for the Transnuclear

TN–68 cask system. The NRC published

a proposed rule in the Federal Register

(64 FR 45920; August 23, 1999) to add

TN–68 cask system to the listing in 10

CFR 72.214. The comment period ended

on November 8, 1999. Three comment

letters were received on the proposed

rule.

Based on NRC review and analysis of

public comments, the NRC staff has

modified, as appropriate, its proposed

CoC, including its appendices, the

Technical Specifications (TSs), and the

Approved Contents and Design Features

for the Transnuclear TN–68 cask

system. The NRC staff has also modified

its preliminary SER.

The NRC finds that the Transnuclear

TN–68 cask system, as designed and

when fabricated and used in accordance

with the conditions specified in its CoC,

meets the requirements of 10 CFR part

72. Thus, use of the Transnuclear TN–

68 cask system, as approved by the

NRC, will provide adequate protection

of public health and safety and the

environment. With this final rule, the

NRC is approving the use of the

Transnuclear TN–68 cask system under

the general license in 10 CFR part 72,

subpart K, by holders of power reactor

operating licenses under 10 CFR part 50.

Simultaneously, the NRC is issuing a

final SER and CoC that will be effective

on May 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.

Summary of Public Comments on the

Proposed Rule

The NRC received three comment

letters on the proposed rule. The

commenters included an industry

representative, an individual member of

the public, and a utility. Copies of the

public documents are available for

review in the NRC Public Document

Room, 2120 L Street, NW (Lower Level),

Washington DC.

Comments on the Transnuclear TN–68

Cask System

The comments and responses have

been grouped into eight subject areas:

General, materials, crud, miscellaneous

issues, technical specifications,

comments on applicant’s SAR,

accidents, and radiation protection. To

the extent possible, all of the comments

on a particular subject are grouped

together. A review of the comments and

the NRC staff’s responses follow:

A. General Comments

Comment A–1: One commenter

requested that the general comments

submitted by the commenter on the TN–

32 rule apply to this rule as well.

Response: Comments that were

general enough to apply to both the TN–

32 and the TN–68 casks, were addressed

in the response to the comments on the

TN–32 rule (65 FR 14790, March 20,

2000). Specific comments are addressed

in this rulemaking for the TN–68 cask.

24856 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

Comment A–2: One commenter stated

that the environmental assessment (EA)

is ‘‘tiered’’ on documents having little to

do with the dry casks of today and that

an Environmental Impact Statement

(EIS) for each generic design should be

done.

Response: The NRC disagrees with the

comment. The EA and Finding of No

Significant Impact (FONSI) for this rule

are limited in scope to the TN–68 in a

generic setting. The NRC has given

specific consideration to environmental

impacts of dry storage and has not

found any new information affecting the

conclusion that these impacts are

expected to be extremely small and not

environmentally significant. Therefore,

the NRC is not convinced that

meaningful new environmental insights

would be gained by performing an

environmental impact analysis for each

new cask that is certified. The EA

covering the proposed rule, as well as

the FONSI prepared and published for

this final rule, 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 (NEPA) and give proper

consideration to the guidelines of the

Council of Environmental Quality

(CEQ). The EA and FONSI prepared for

the TN–68, as required by 10 CFR part

51, conform to NEPA procedural

requirements. Tiering on past EISs and

EAs is a standard process under NEPA.

As stated in CEQ’s 40 Frequently Asked

Questions, the tiering process makes

each EIS/EA of greater use and meaning

to the public as the plan or program

develops, without duplication of the

analysis prepared for the previous

impact statement.

Comment A–3: One commenter stated

that decommissioning, transport, and

disposal of fuel from these casks have

not been adequately analyzed.

Response: The CoC for the TN–68 is

for the storage of spent fuel.

Decommissioning, transport, and

disposal of fuel from the casks is beyond

the scope of this rule.

Comment A–4: One commenter stated

that the environmental impacts would

not be the same for a general license and

a site-specific license.

Response: The NRC disagrees with the

comment. Each cask is designed and

fabricated to specific design criteria

whether it is licensed for site-specific or

general use. The process for determining

the environmental impact varies, but the

cask must satisfy the same technical

requirements. There are no significant

environmental impacts using a spent

fuel dry storage cask under either a sitespecific or a general license.

Comment A–5: One commenter stated

that previous fabricators of casks have

not realized that the casks are made to

store nuclear spent fuel and the quality

of their work can affect the health and

safety of the public. The commenter

asked why the NRC is ‘‘opening up’’ the

approval process to lower standards by

fabricators, material suppliers, and

inspectors.

Response: The NRC disagrees with

this comment. All licensees/CoC

holders must have a quality assurance

(QA) program that has been approved by

the NRC as part of the licensing or CoC

issue process. This QA program must

meet the requirements of 10 CFR 72.148

and 72.154 in regards to the selection of

fabricators. The licensee/CoC holder is

required to assure that all regulations

and certificate conditions applicable to

the cask are met. In addition, the

licensee/CoC holders and fabricators are

subject to NRC inspections to verify

compliance.

Comment A–6: One commenter stated

that the design should be built and

tested before certification and that NRC

approving a design without a test is

wrong, and asked if the NRC is going to

allow the first cask to be tested by a

utility.

Response: The NRC disagrees with the

comment. The TN–68 cask design has

been reviewed by the NRC. The basis of

the safety review and findings are

clearly identified in the SER and CoC.

Testing is normally only required when

the analytic methods have not been

validated or assured to be appropriate

and/or conservative. In place of testing,

the NRC staff finds acceptable analytic

conclusions that are based on sound

engineering methods and practices. As

detailed in the SER, the NRC staff has

reviewed the analyses performed by TN

and found them acceptable.

Comment A–7: One commenter noted

a lack of confidence that the vendor

knows what it is doing when it is

permitted by the NRC to make a best

effort in the realm of testing and

verification of weld quality.

Response: In fabrication, the specific

nondestructive examination desired or

otherwise required for a particular weld

sometimes cannot be performed due to

joint geometry or part configuration. As

used here, the term ‘‘best effort’’ means

the joint will be examined using other

acceptable methods suitable for the

application under the American Society

of Mechanical Engineers (ASME) code.

Specifically, on the weld of the bottom

inner plate to the confinement shell

where the weld cannot be examined by

ultrasonic testing (UT), the weld will be

examined by radiographic testing (RT)

and either penetrant testing (PT) or

magnetic particle testing (MT) under

ASME Subsection NB requirements.

Comment A–8: One commenter stated

that everything in the cask should be

identified on the cask label in case

documents are lost or destroyed.

Response: The NRC disagrees with

this comment. NRC regulations do not

require the identification of cask

contents on permanent markings affixed

to the cask. The need for labeling was

evaluated during the rulemaking that

established Subpart L in 10 CFR part 72

entitled ‘‘Approval of Spent Fuel

Storage Casks’’ (55 FR 29193; July 18,

1990). The NRC notes that § 72.212(b)(8)

requires that each general licensee

accurately maintain a record for each

cask that lists the spent fuel stored in

the cask. The record must be maintained

by the cask user until decommissioning

of the cask is complete. Also, § 72.72

requires that records of spent fuel in

storage must be kept in duplicate, with

the duplicate set sufficiently remote

from the original records that a single

event would not destroy both sets of

records.

Comment A–9: One commenter asked

if the ‘‘less than 1 gram-mole/cask’’

recommendation listed on Page 8–2 of

the SER came from PNL–6365,

‘‘Evaluation of Cover Gas Impurities and

Their Effects on the Dry Storage of LWR

Spent Fuel,’’ R.W. Knoll and E.R.

Gilbert, Pacific Northwest Laboratory,

Richland, Washington, November 1987;

what kind of dry storage PNL evaluated;

and what dry storage casks were in use

before 1987? The commenter then

added a recommendation that the

reference be updated.

Response: The less than 1 gram-mole/

cask limit is from the cited reference.

The investigators evaluated four cask

designs loaded with spent fuel, the MC–

10, TN–24P, Castor-V/21, and MSF IV.

Further details are contained in the

report. Dry storage casks in use before

1987 were the Castor V/21, the MC–10,

and the NUHOMS–7P. The NRC

considers this reference material to be

acceptable and that it does not need to

be updated.

Comments A–10: One commenter

recommended that detailed site-specific

unloading procedures should not be

developed by licensees. Instead, the

NRC should fully inspect the

procedures and place them in the PDR

before any cask loading is done at the

plant. The commenter also suggested

that contamination control measures

should be carefully thought out to

adequately address the presence of fuel

crud, and suggested that the generic

review should pay more attention to a

detailed plan for emergency cask

unloading including how contamination

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24857

is controlled, especially crud, and how

effluents are released.

Response: The NRC disagrees with

this comment. The TN–68 Storage Cask

System Design operating descriptions

and analysis have been reviewed and

accepted by the NRC. The NRC staff

concluded in the SER that there was

reasonable assurance that the cask

unloading operations could be safely

performed by qualified personnel using

detailed procedures developed by the

cask user at an ISFSI site. Cask general

licensees must be licensed under 10

CFR 50. These licensees have sufficient

infrastructure, experience, and

processes in place to develop adequate

detailed unloading procedures without

prior NRC review. Detailed site-specific

procedures for performing unloading

operations, including contamination/

effluent control measures, are required

to be developed and demonstrated at

each facility that uses the TN–68.

Comment A–11: One commenter

stated that the use of a proprietary

neutron shielding material is not in the

interest of the public health and safety,

and that the best neutron shielding

material should be identified and

available for use by all vendors and

licensees.

Response: This comment is beyond

the scope of this rule. The applicant’s

proposed materials have been found by

the NRC staff to be acceptable. The

critical attributes of the material are not

proprietary and are specified in the CoC

and SER.

Comment A–12: One commenter

stated that the public would be better

served if one design would be approved

for casks rather than the large number

that is being approved based on utilities

choosing the least expensive designs.

Response: This comment is beyond

the scope of this rule. NWPA gives NRC

authority to approve multiple cask

designs.

Comment A–13: One commenter

asked where the decontaminated TN–68

components would be stored and where

the remaining low-level waste would be

disposed.

Response: This comment is beyond

the scope of this rule. Disposal of lowlevel waste is covered by 10 CFR parts

20 and 61.

Comment A–14: One commenter

stated that NRC is approving generic

designs which allow site specific

changes by utilities that use the casks

and that this makes it difficult to

establish a standardized, integrated total

waste system for the United States. The

commenter further stated that approval

of generic designs is creating vendor

competition to rapidly develop cheap

designs with current materials instead

of competition to create the best and

safest designs. The commenter asked

how many designs does the NRC plan

to allow in the industry and how will

approving a large number affect

shipping and final disposal of spent

nuclear fuel.

Response: This comment is beyond

the scope of this rule. NWPA gives NRC

authority to approve multiple cask

designs.

Comment A–15: One commenter

stated that NRC documents are long,

repetitive, and hard to understand. The

commenter also stated that the more

people who go over these documents

and ask questions, the better.

Response: The NRC agrees that

documents should be easy to

understand. Because the documentation

necessary to license a storage cask is

tiered and must be comprehensive to

document the NRC staff’s evaluation

and findings, the documentation may be

extensive. The NRC documents are

available for public comment.

Comment A–16: One commenter

disagreed that sabotage scenarios have

been fully evaluated, and stated that

sabotage evaluation for site-specific

parameters should be updated.

Response: The NRC disagrees with the

comment. The NRC reviewed potential

issues related to possible radiological

sabotage of storage casks at reactor site

ISFSIs in the 1990 rule that added

Subparts K and L to 10 CFR part 72

(55FR 29181; July 18, 1990). The NRC

still finds the results of the 1990 rule

current and acceptable. Spent fuel in the

ISFSI is required to be protected against

radiological sabotage using provisions

and requirements as specified in 10 CFR

72.212(b)(5). Each Part 72 licensee is

required by § 73.51 or § 73.55 to develop

a physical protection plan for the ISFSI

and to install a physical protection

system that provides high assurance

against unauthorized activities that

could constitute an unreasonable risk to

the public health and safety. Each ISFSI

is periodically inspected by NRC, and

the licensee conducts periodic patrols

and surveillances to ensure that

physical protection systems are

operating within their design limits.

B. Materials

Comment B–1: One commenter asked

what is a torispherical weather cover

with elastomeric seals and why all dry

cask designs should or should not have

them.

Response: The torispherical weather

cover is a protective cover that provides

weather protection for the closure lid,

top neutron shield, and overpressure

system. The use of such a cover on other

cask designs is beyond the scope of this

rule.

Comment B–2: One commenter asked

why TN is allowed to use alternative

neutron shield materials as discussed in

the CoC. The commenter also asked why

the current materials of borated wrought

aluminum alloy or BorALYNTM have

not been approved with no alternative

and why the best material is not chosen

for the design at this point. The

commenter stated a concern about the

number and complexity of criteria for

BorALYNTM fabrication in that it results

in a complicated fabrication process.

The commenter recommended that

more research be conducted to find a

better neutron shield material with less

problems. The commenter stated that

TN appears not to be satisfied with the

current neutron shield materials

because they ‘‘envision an alternative

candidate’’ for which they need to

develop appropriate qualification test

data, and asked why TN and NRC are

not waiting until the improved neutron

shielding material is available before

certification of the CoC.

Response: The applicant’s proposed

materials have been found acceptable by

the NRC staff. After careful review of

this material and its properties under

various conditions, the staff is not aware

of any problems with this material in its

intended service.

Comment B–3: One commenter asked

if the casks can be moved with the

temperature above freezing.

Response: TS 3.1.6 requires that the

loaded cask not be lifted if the outer

surface of the cask is below ¥20°F.

There is no other temperature restriction

for moving the TN–68 cask.

Comment B–4: One commenter asked

if the cask meets ASME code standards,

asked if the applicant has adequately

justified an exemption from the code

requirements, and if the NRC staff has

verified this action.

Response: The cask is designed,

fabricated, and inspected under the

appropriate subsections of the ASME

Code. Exceptions to the ASME Code are

listed in Table 4.1–1 of the TSs and

Section 7 of the SAR. These exceptions

and associated justifications and

compensatory measures were reviewed

by the NRC staff and found to have no

adverse effects on the cask integrity. The

basis for cask approval is documented

in the SER.

Comment B–5: One commenter stated

that the CoC specifications about

fabricator verification of the quality of

the welding of the inner plate to the

confinement shell are somewhat vague

and do not specify firm requirements.

Examples cited were statements in the

CoC that ultrasonic testing (UT) of the

24858 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

weld will be performed on a best effort

basis, the joint examination can be

performed by a number of methods, the

joint may be welded after shrink fitting

of the shells, and that the geometry may

not allow for UT examination. The

commenter also asked if there had been

problems with the shield weld in

previous designs.

Response: The NRC disagrees that the

specifications are vague. ASME Code,

Section III, Division 1, Subsection NB–

5231(b) requires either ultrasonic or

radiographic examinations and either

liquid penetrant or magnetic particle

examinations be performed on the full

penetration corner welded joints.

Therefore, the applicant can choose

either ultrasonic or radiographic

examinations to inspect the corner

weld. The bottom inner plate weld is

inspected using ultrasonic examination

methods if the weld is applied before

the outer and inner shells are

assembled. If the weld is applied after

assembly, this inspection is done

radiographically. Both methods will be

supplemented by either liquid penetrant

or magnetic particle examinations. The

NRC staff is not aware of any problems

with the shield weld designs.

Comment B–6: One commenter asked

what the shrink fit process is and if it

has been used and time tested before,

questioned using shrink fit and

frictional forces to keep the shells from

separating, and asked if the shrink fit

will be performed before the welding of

the bottom confinement shell.

Response: The shrink fit is established

as follows: The gamma shield shell and

the confinement shell are fabricated

separately. To obtain a close fit between

these two shells, the outside diameter of

the confinement shell is slightly larger

than the inside diameter of the gamma

shield shell. The gamma shield shell is

preheated which causes it to expand

before slipping on the confinement

shell. As the gamma shield shell cools,

it shrinks and tightly clamps onto the

confinement shell. Shrink fit is a

common industrial practice that has

been used to fabricate various nuclear

components including those used

successfully in other NRC-approved

casks. Fire, tipover, or seismic events

would not cause the two shells to

separate as demonstrated in Sections 3,

4, and 11 of the SAR. The SAR specifies

welding either before or after shell

assembly. As long as the confinement

barrier is welded to meet ASME Code

Section III, Subsection NB requirements,

test standards, and acceptance

standards, the barrier will conform with

a standard that will satisfy all of the

safety requirements for this application.

Comment B–7: One commenter stated

that 30 days in the pool for a cask is a

long time, and asked what happens to

neutron absorber material, aluminum

paint, etc., during this extended period

of time. The commenter stated that

assemblies left in a cask cavity in the

pool are very different from just being

in the pool out of the cask, and asked

how fast the hot water is going to be

exchanged with cooler pool water when

fuel is left in the cask with the cover

removed. The commenter also asked if

the water in the pool is constantly

cooled, how cask walls will affect that

bit of pool water in the cask with the 68

assemblies compared to the rest of the

pool, how cask materials will affect pool

water and pool filters if left in the pool

for 30 days, if crud will come off the

assemblies that were dried and put back

in the pool, if iron oxide will come off

the paint, and what chemicals in the

pool could be affected by the cask being

in the pool for seven versus 30 days.

Response: The effect of the water on

these materials is negligible. The

reactions with pool water occur very

slowly and give rise to only small

amounts of hydrogen, ions, and/or

precipitates in the pool water that are

trapped by filters designed to capture

small items from the water. This is true

for aluminum, aluminum ‘‘paint,’’ and

the stainless and coated ferrous

materials used in this system. The

aluminum is not a paint but it is

aluminum and aluminum oxide that

when applied as a liquid spray of

aluminum to the cask surfaces, becomes

tightly adherent to the substrate onto

which it is applied. Some of the

aluminum becomes an oxide of

aluminum during this process. Neither

the aluminum oxide nor the iron oxide

is expected to come off the paint when

exposed to pool water. The system is

designed to allow the free movement of

pool water into the cask with the lid

removed, and systems are in place to

constantly cool the pool water. The

water in a BWR pool is typically pure

water which has no chemical addition

unless that chemical is evaluated on a

site-specific basis. The questions

specific to operation of the spent fuel

pool are beyond the scope of this rule.

Comment B–8: One commenter stated

that 48 hours without helium seems to

be the maximum time for the basket and

that even if fuel temperature limits are

not reached, there could be basket

damage. The commenter stated this

should be made clearer in the CoC.

Response: The NRC agrees that there

is a potential for exceeding the basket

temperature limits after 48 hours. To

protect the basket, the TSs require that

the licensee initiate and complete a

helium backfill procedure at the 42 and

48 hour marks, respectively. This is

stated in TS bases B3.1.1 and B3.1.2,

SER Section 4.5.2.4, and SAR Section

4.6.2.

Comment B–9: One commenter

questioned a CoC statement that flaws

in the gamma shield are not examined

no matter what is typically observed in

the material. The commenter suggested

that a large crack could let water in and

cause rusting of materials.

Response: The NRC disagrees with the

comment. The gamma shield is a forged

component. Flaws in forgings are very

small. There is no safety related risk or

materials problem related to the use of

a forging in this application. The

allowable flaws for various orientations

and locations are stated in Appendix 3E

of the SAR. Flaws of these sizes will not

propagate under service conditions.

Comment B–10: One commenter

asked why there are lower trunnions for

rotating the cask from horizontal to

vertical.

Response: The unloaded cask may be

shipped from the manufacturer to the

site in a horizontal orientation. The

lower trunnions provide capability to

rotate the cask to the vertical orientation

before loading of spent fuel. The upper

trunnions are the only components used

for lifting the loaded cask.

Comment B–11: One commenter had

a number of concerns related to the

neutron source and neutron shielding.

The commenter stated that enrichment,

burn up, and fuel cooling time seem to

be crucial to avoid having a neutron

source too high. The commenter also

stated that the neutron shield material

choice and structure is flimsy and a

better choice of material is needed, and

that because in the SER the NRC stated

‘‘all of the fire accident temperatures

were below short-term design-basis

temperatures with the exception of the

neutron shield material,’’ the design

should use another material. The

commenter asked what would be the

expected result of a long term fire for

the neutron material, why the design

includes a neutron shield material that

can off-gas during a fire, what gas would

be given off by the combustion of the

neutron shield, how the gas would

react, if the gas is explosive, or if it

would react with anything from a plane

crash or truck bomb to make the

problem worse. The commenter stated

that the fire accident should be

evaluated to consider the effects of

neutron shield resin burnup. The

commenter also stated that the KX–277

material in the VSC–24 design and the

proposed resin shielding in TN casks

can contain voids, is not strong, and is

flammable, while alloys being discussed

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24859

for Yucca Mountain seem much better

and more expensive. The commenter

further stated that having a

multipurpose cask with better shielding

would be better in the long run instead

of vendors using the cheapest materials.

Response: The NRC concurs with the

comment on the parameters important

to a neutron source term. These

parameters are controlled in Section 2 of

the TSs. The NRC staff disagrees that a

different neutron shield material is

needed. The proposed material was

evaluated and found to satisfy the safety

requirements for the application. The

top neutron shield and the radial

neutron shield have not been designed

to withstand all of the hypothetical

accident conditions. Cask structural

analyses have been performed assuming

that the neutron shield is completely

removed during accident conditions.

The results indicate that the cask

without the neutron shield is adequately

designed to withstand various load

combinations of the accident condition

as presented in Sections 2, 3, 4, and 11

of the SAR. The cask has been analyzed

for the post-fire condition and has been

found to meet the dose requirements of

10 CFR 72.106 even without the neutron

shielding being present. The question

on a long-term fire is beyond the testing/

analysis required by Part 72. The radial

neutron shield is a polymeric material

that includes about 50 weight percent

fire-retardant mineral fill, which makes

it self-extinguishing. The polymeric

neutron shield materials may char or

off-gas if directly exposed to fire or high

temperatures. The applicant has

modified the SAR to address the

combustibility of the neutron shield.

The off-gas products are formed from a

very small fraction of the total neutron

shield mass and are not explosive but

may burn during the fire. The heat input

from this reaction would be

insignificant relative to that of the

design basis fire. Comments on the

VSC–24 material, Yucca Mountain, and

multipurpose casks are beyond the

scope of this rule.

Comment B–12: One commenter

asked about information included on

Section 4.5.2.4 of the SER. Specifically,

the commenter asked if partial pressure

injection of helium had ever been

performed for a similar cask, where, and

what were the results. The commenter

also asked if the air-helium mixture will

really work. Further, the commenter

stated that the NRC referred to a

‘‘different cask system’’ and asked what

data is applicable to the different cask

system and if it can apply to the TN–

68 design.

Response: The purpose of the helium

injection is to improve the thermal

conductivity of the fill gas as a

temporary measure to provide an

opportunity to troubleshoot and repair

any problems during the drying or

helium fill process. ISG 7, ‘‘Potential

Generic Issue Concerning Heat Transfer

in a Transportation Accident’’ dated

October 2, 1998, provides NRC staff

guidance for mixtures of gases within a

spent fuel storage cask. In support of

ISG–7, a sensitivity study was

performed to evaluate the relative

change in cladding temperatures as a

result of significant reductions in the

thermal conductivity of the fill gas (e.g.,

30% that of helium). This evaluation

found that the cladding temperature was

relatively insensitive to gas thermal

conductivity as evidenced by an

increase in the fuel cladding and bulk

gas temperatures of about 3%. The NRC

staff did not review or require any

testing of the helium injection process

based on the analysis performed for

ISG–7 and the restrictions, imposed by

the TN–68 TSs, on operations without a

full helium environment to maintain the

desired protection for the cladding.

Comment B–13: One commenter

stated that the SER states the NRC staff

projected a peak cladding temperature

lower than the long term storage

cladding temperature limit if the

fabrication results in gaps of 0.05 in. or

less between component layers. The

commenter asked if the NRC would

accept up to a 0.05 inch gap and why

the applicant’s assumed gap of 0.01 in.

should not be the fixed limit.

Response: Gaps between the various

cask components were assumed in the

analysis to account for fabrication and

assembly tolerances and uncertainties.

The implemented QA program at the

fabricator’s facility provides reasonable

assurance that the as-built casks will

have gaps that are less than or equal to

those assumed in the analysis. In the

context of the statements referenced by

the commenter, the NRC performed a

sensitivity analysis to evaluate the

response of the cask thermal

performance to increased gap sizes. The

results of that evaluation found that

gaps could be five times that assumed

in the analysis and the fuel cladding

would remain within temperature

limits.

Comment B–14: One commenter

expressed concern over the continued

efficacy of the neutron absorber plates

over 20 years of storage. In addition, the

commenter stated that the NRC needs to

look more carefully at issues such as

unexpected erosion or corrosion,

potential explosions, and cracks in

welds for the life of the cask. The

commenter also stated dislike of

materials used in this design including

poured resin, borated aluminum, and

metal matrix.

Response: The neutron absorber is

designed to remain effective in the TN–

68 system for a storage period greater

than 20 years. Section 6.3.2 of the TN–

68 SAR describes the neutron absorber

and its environment, and evaluates

boron depletion due to neutron

absorption. Section 9.1.7 of the SAR

describes the testing procedures for the

neutron absorber material, which will

be manufactured and tested under the

control and surveillance of a quality

assurance and quality control program

that conforms to the requirements of 10

CFR part 72, subpart G. The

compositions and densities for the

materials in the computer models were

reviewed by the NRC staff and

determined to be acceptable. The NRC

staff notes that these materials are not

unique and are commonly used in other

spent fuel storage and transportation

applications.

The NRC staff disagrees that the stated

issues need to be looked at more

carefully. The NRC is already looking

carefully at the materials that may

impact the safe performance of storage

systems. As part of this effort, the NRC

has participated, over the past several

years, in the work of a Task Group of

Subcommittee C26.13 of the American

Society of Testing and Materials on life

extension questions. This Task Group

has been developing guidance for

components of storage cask systems for

periods up to a 100-year service life.

This work is taken into account in the

reviews that are ongoing for storage

systems. Erosion and corrosion are not

expected to occur at any level

significant enough to affect safe

performance of components of the cask.

The TN–68 is designed to withstand an

external pressure of 25 psi. This would

include a nearby explosion, debris

falling on the cask, etc. If a credible

explosion is identified that would apply

more than 25 psi to the outer surface of

the cask at a site, the site will have to

address this issue in its 10 CFR 72.212

evaluation. Any cracks in welds or other

flaws in components are small in

relation to what is needed to extend

these cracks in service. Fracture

mechanics calculations can be used to

show them to be stable (will not

propagate) for the levels of stress to be

sustained in service.

Regarding the commenter’s dislike for

particular materials, material selection

is the applicant’s responsibility. The

applicant must demonstrate that the

materials and the materials’ properties

satisfy the requirements for a given

application.

24860 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

Comment B–15: One commenter

recommended that the installation of a

blind flange on the overpressure

monitoring system (OMS) to mitigate a

latent seal failure event should be tested

to verify that it will work.

Response: The NRC disagrees with

this comment. The possibility of the

occurrence of the events needed to

occur concurrently for a latent seal

failure event is judged to be very

remote. If this unlikely event were to

occur, the mitigative action to install a

blind flange at the OMS port is

straightforward and well within the

capability of a nuclear power plant

licensee. Therefore, the NRC has

reasonable assurance that the action can

be taken without additional testing.

Comment B–16: One commenter

asked the NRC to explain ‘‘bubble leak

tests’’ in relation to resin enclosures and

leak passages on weld enclosures. The

commenter also asked how test failures

are rectified and rechecked.

Response: This test is described in

ANSI 14.5–97, ‘‘American National

Standard for Radioactive Materials—

Leakage Tests on Packages for

Shipment’’ February 1998. Deficiencies

are evaluated, repaired, and retested

under the cask vendor’s QA program, as

described in SAR Section 13.

Comment B–17: One commenter

stated that the following editorial

corrections should be made in the TS:

On the bottom of page 1.2–1,

‘‘continued’’ should be moved above the

line; on page 1.3–5, ‘‘Time the’’ should

be moved from the first column to the

second column of information; on the

bottom of page 3.0–1, ‘‘continued’’

should be added below the line; at the

top of page 3.0–2, ‘‘3.0 LCO

APPLICABILITY (continued)’’ should be

added; at the bottom of the page 3.0–2,

‘‘continued’’ should be moved above the

line; at the top of page 3.0–4, the

‘‘continued’’ above the line should be

deleted and the ‘‘continued’’ below the

line should begin with a lower case

letter; and on page 3.1.1–1, the double

line separating conditions B and C

should be changed to a single line.

Response: The NRC agrees with these

changes. The TSs have been reformatted

accordingly.

Comment B–18: One commenter

stated that on drawing 972–70–2 of the

SAR, the materials for the protective

cover should be changed to SA–516 GR.

70 or SA–105 to allow the cover flange

to be made from a forging.

Response: The NRC accepts this

change to the protective cover materials

because the material properties are the

same. This change will not affect the

structural analyses and the conclusions

reached in the SER. Drawing No. 972–

70–2 has been changed accordingly.

Comment B–19: One commenter

stated that on drawing 972–70–3 of the

SAR, a note should be added to allow

the protective cover flange to be made

from a one-piece forging.

Response: The NRC accepts this

change because it will not affect the

structural analyses and the conclusions

reached in the SER. Drawing No. 972–

70–3 has been changed accordingly.

Comment B–20: One commenter

stated that the material of the metallic

seals described in Chapters 2 and 7

should be changed to allow a stainless

steel or nickel alloy liner.

Response: The NRC agrees with this

comment. The use of either stainless

steel or nickel alloy is acceptable to the

NRC staff. The SAR has been changed

to reflect this change.

Comment B–21: One commenter

stated that on page 3–5 of the SER, the

basis for the allowable stress for the

6061–T6 alloy is in error.

Response: The NRC disagrees with

this comment. The basis for the

allowable stress for the 6061–T6 alloy is

Section III of the ASME Code, as stated

in Section 3.1.4 on page 3–5 of the SER.

C. Crud

Comment C–1: One commenter asked

what would be done if cask vent flow

of saturated steam could not be

discharged into the spent fuel pool

during reflooding of the cask before

unloading. The commenter also asked

what conditions could preclude

discharge to the spent fuel pool,

specifically asking about too much

radioactivity, failed fuel, crud, fuel

fines, and iron oxide debris.

Response: As shown in SAR Figure

8.2.1, the cask may be vented to the

spent fuel pool or to the radwaste

system. The reasons suggested by the

commenter that may impact the cask

vent location are interpreted to be

primarily radiological concerns. The

procedure descriptions for cask

unloading include appropriate reference

to development of site-specific

procedures and actions that will

maintain exposures to workers and

radiological releases to the environment

as low as reasonably achievable

(ALARA). The details of where the cask

will be vented are a site-specific matter

and beyond the scope of this rule.

Comment C–2: One commenter has a

number of concerns about crud on

boiling water reactor (BWR) fuel: What

material composes the crud and should

it be allowed in a cask; how crud is

analyzed in all aspects of cask loading,

transfer, storage, and unloading, and

when fuel is put back in the pool and

then loaded in a transport cask or

placed in different reactor pools; what

happens to the dried crud when it is put

back into the pool, and how it affects

pool water quality; whether crud covers

defects in cladding that may be revealed

when it dries and falls off; and if BWR

crud is different than pressurized water

reactor (PWR) crud.

Response: Crud generally consists of

oxides of metals (e.g., Co, Mn, Cr, Fe, Zr,

Zn) that are not chemically reactive in

the storage cask environment. The crud

collects on the exterior of the fuel

cladding during reactor operation. The

crud particles for BWR fuel are very

small with diameters ranging from 0.1 to

10 micrometers as reported in SAND88–

1358, ‘‘Estimate of Crud Contribution to

Shipping Cask Containment

Requirements’’ January 1991. SAND88–

1358 found that the crud on BWR fuel

was less adherent than that found on

PWR fuel. Some crud may be dislodged

or spall from the fuel cladding during

spent fuel dry storage or handling;

however, there were no differences

reported in the spallation behavior of

crud between the two fuel types.

The safety concern associated with

crud is its radiological impact. The

analysis provided by the applicant uses

a bounding assumption for crud activity

of 1254 μCi/cm2 of Cobalt-60 (this was

the maximum activity level found by

actual inspection of BWR fuel)

distributed over the entire fuel cladding

surface. The analysis demonstrates with

reasonable assurance that fuel loading,

storage, and unloading can be

performed safely. The NRC agrees with

the commenter that some crud may be

flushed from the cask to the spent fuel

pool as a part of the unloading process.

The operating procedure descriptions

address this possibility and the

precautions for handling this situation.

Regarding the impact of crud in the

spent fuel pool, there is crud from wet

fuel storage already present in a spent

fuel pool and the amount of crud from

the spent fuel cask is expected to be

very small. If any crud is discharged to

the spent fuel pool, it would be

captured in the spent fuel pool filtration

system.

Regarding the concern with crud

covering defects in cladding and later

being revealed when the crud dries and

falls away from the defect, the effects of

the dislodged crud were addressed

earlier in this comment response. The

comment also raises the possibility that

a cladding defect may be covered by

crud, thus allowing the defect to go

undetected during visual inspection of

the fuel before loading. Cask users must

ensure that the fuel loaded into the cask

meets the requirements of TS 2.1.1. This

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24861

TS precludes loading fuel that has

known cladding defects greater than

pinhole leaks or hairline cracks. Cask

users may use a variety of screening

methods to ensure that the fuel meets

the TS requirements. These screening

methods include review of operational

records, visual inspections, fuel

assembly sipping, and ultrasonic

examination. Because multiple

screening methods are used, the NRC

has reasonable assurance that the fuel

can be adequately screened for

compliance with the TS requirements.

Further, if a postulated assembly with a

cladding defect not meeting the TS

requirements was loaded, the NRC does

not expect a significant adverse impact

in the radiological consequences

because the confinement system

remains intact during normal, offnormal, and accident conditions.

The impacts of crud on transportation

activities are beyond the scope of this

rule.

Comment C–3: One commenter stated

in reference to page 9–4 of the SER that

during unloading a problem could arise

due to precipitates, or second-phase

particles, even if titanium decreases

their size, and noted that any particle or

precipitate in unloading, along with

crud, etc., is going to be a big concern.

Response: The NRC interprets the

comment as a concern for potential

loose particles in the cask cavity and

disagrees that the particles and

precipitates, discussed on page 9–4 of

the SER, are a cause for concern in

unloading. The discussion on page 9–4

of the SER refers to boride precipitates

that are components of the metal matrix

in the borated aluminum plate and will

not separate from the plate material

during unloading. In response to the

commenter’s question about other

particulates, including crud, Comment

C–2 responds to that concern.

D. Miscellaneous Items

Comment D–1: One commenter stated

that reference 4 on Page 5–7 of the SER

should be revised or updated.

Specifically, the commenter stated that

more current references than those from

the 1970’s should be used or the NRC

should do new research in the area to

develop more recent guidance for design

review.

Response: As stated on Page 5–2 of

the SER, references 4 and 5 were

consulted by the NRC staff to determine

the appropriate values for the assumed

cobalt impurity levels in the fuel

assembly hardware. Reference 5 is more

recent and was published in 1993.

Comment D–2: One commenter asked

what is the ‘‘potentially oxidizing

material’’ that must be removed from

the cask to protect the fuel cladding

during storage.

Response: Potentially oxidizing

impurities include oxygen, carbon

dioxide, carbon monoxide, and water.

Oxidizing impurities, their removal, and

their effects are discussed in detail in

PNL–6365, ‘‘Evaluation of Cover Gas

Impurities and Their Effects on the Dry

Storage of LWR Spent Fuel’’ November

1987.

Comment D–3: One commenter

requested that ‘‘fuel fines’’ be defined.

Response: From NUREG/CR–6487,

‘‘Containment Analysis for Type B

Packages Used to Transport Various

Contents’’ November 1996, fuel fines are

particulate material composed of fuel

compounds and are produced as a result

of mechanical stresses at both the fuelcladding interface and the fuel pelletfuel pellet interface. This definition is

applicable to both transport and storage

of light water reactor spent fuel.

Comment D–4: One commenter

recommended that reference 9, in NRC

Regulatory Guide 1.25, U.S. Nuclear

Regulatory Commission, ‘‘Assumptions

Used for Evaluating Accidents in the

Fuel Handling and Storage Facilities for

Boiling and Pressurized Water Reactors’

(March 1972), should be revised by the

NRC and updated.

Response: Updating this Regulatory

Guide is beyond the scope of this rule.

Comment D–5: One commenter

suggested that a berm be used in the

design.

Response: Under 10 CFR 72.212(b)(2),

each general licensee who uses the TN–

68 cask must perform an evaluation to

show that the regulatory off-site dose

limits are met at the licensee’s site. The

evaluations are made available for NRC

inspection and review. Depending on a

number of site specific factors including

cask array size and distance to the

nearest member of the public, a berm

may or may not be needed.

Comment D–6: One commenter

suggested that reference 1 listed on Page

10–4 of the SER, dated 1978, be

updated.

Response: Updating reference 1

(Regulatory Guide 8.8) is beyond the

scope of this rule.

Comment D–7: One commenter stated

that on page 3–5 of the SER, the third

paragraph ends in an extraneous ‘‘0.’’

Response: The NRC agrees with this

comment and the SER has been changed

accordingly.

Comment D–8: One commenter stated

that on page 7–6 of the SER, reference

5 should be updated to reflect issuance

of ISG–5, Revision 1.

Response: The NRC agrees with this

comment. ISG–5 Revision 1 and the

draft of the TN–68 SER were issued at

nearly the same time. Because the

principles and methods described in the

revised ISG were reflected in the SER,

it is appropriate to revise the SER to

update this reference.

E. Technical Specifications

Comment E–1: One commenter stated

that the use of logical connectors makes

technical specifications difficult to read.

The commenter asked if industry

workers have commented on the

technical specifications and find them

easy to understand.

Response: The NRC disagrees with the

comment. The TSs are modeled on the

Improved Standard Technical

Specifications (ISTS) for power reactors.

The ISTS were developed as a result of

extensive technical meetings and

discussions between the NRC staff and

the nuclear power industry in the early

1990’s, in an effort to improve clarity

and consistency of the power TSs and

to make them easier for the operators to

use. The most likely users of the TN–68

TSs are power reactor licensees familiar

with the format of the ISTS.

Comment E–2: One commenter

questioned why there are extensions of

time intervals in the surveillance

requirements and stated that the

surveillance should be done according

to schedule. The commenter stated that

the 25-percent extension of the specified

interval for performance of surveillance

in the TS will be confusing and used

when not applicable. The commenter

also stated the same goes for the delay

period of up to 24 hours or up to the

limit of the specified frequency when it

is discovered a surveillance has not

been performed. The commenter

suggested that extensions and extra

leeway should be the explained

exceptions rather than the regular

allowance, and that the writeups were

too complicated with too many options.

Response: The NRC disagrees that

extensions of time should not be

allowed. The basis for surveillance

requirement (SR) 3.0.2 is discussed in

the TN–68 Technical Specification

Bases Section B 3.0 ‘‘Surveillance

Requirement Applicability.’’ This

section explains the NRC staff’s

rationale for allowing a 25-percent

extension in the completion of periodic

surveillances. The NRC staff believes

that the 25-percent extension does not

significantly degrade the reliability that

results from performing the surveillance

at its specified frequency. For those

cases where it is necessary to adhere to

a strict time frame for completing a

surveillance, the specific SR will state

that the 25-percent extension of SR 3.0.2

is not applicable. The 25-percent

extension is also not applicable in cases

24862 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

when a surveillance frequency is

specified by a regulation, because

regulatory requirements take precedence

over TSs. The NRC staff believes that

the provisions of SR 3.0.2 are clear to

users of the TSs, and that they will

ensure that all required surveillances

will be performed within an acceptable

time period, consistent with the NRC

staff’s safety analyses.

Comment E–3: Two commenters

requested changes to the maximum rod

pitch and minimum rod outside

diameter in TS 2.1. One commenter

requested removal of these parameters

because they cannot be verified by

direct means. The other commenter

requested that the values be specified as

nominal [in the TS].

Response: The NRC disagrees with

removing the parameters and changing

them to nominal values. This design

information is crucial to the conclusions

reached by the NRC staff in its SER. The

rod pitch and diameter, along with other

design parameters, already include any

design tolerances considered in the

SAR. As stated in the TS bases for

TS2.1, that have been modified for

clarification, these parameters may be

checked by administrative review.

Comment E–4: Two commenters

requested changes to the maximum

uranium content in TS 2.1. One

commenter requested removal of this

parameter because it may be overly

restrictive. The other commenter

requested that the values be specified as

nominal.

Response: The NRC staff disagrees

that the maximum uranium content

parameters should be changed. This

design information is crucial to the

conclusions reached by the NRC staff in

its SER. The TS limits on uranium

content are based on the most limiting

values used in the criticality and

shielding analyses and include any

design tolerances considered in the

SAR. SAR table 5.2–1 shows that the

calculated maximum uranium content

used in the shielding analysis is higher

than actual values. Although TS Basis

2.1.1 states that the shielding evaluation

is based on nominal uranium content,

the values used in the SAR evaluation

are either greater than or equal to the TS

values. TS Basis 2.1.1 has been changed

to clarify those values.

Comment E–5: Two commenters

stated that the channel thickness in TS

2.1 should be identified as a nominal

value instead of a maximum [in the TS].

Response: The NRC staff agrees with

this comment. However, the applicant

provided the maximum rod channel

thickness and the supporting analysis in

its submittal, and did not provide

analysis to support nominals. Therefore,

the TS has not been changed, although

the basis has been modified for

clarification.

Comment E–6: One commenter asked

what are boiling water reactor (BWR)

fuel assembly channels.

Response: A fuel channel is the part

of the BWR fuel assembly that

surrounds the fuel bundle. The channel

is located between the upper and lower

tie plate and is made of Zircaloy.

Channels perform functions that form a

flow path for bundle coolant flow,

provide surfaces for control rod

guidance, provide structural stiffness to

the bundle, and provide for in-core fuel

sipping.

Comment E–7: Two commenters

stated that the parameter labeling of

Table 2.1.1–1 of the TS should be

revised as Minimum Initial Enrichment

and Maximum Burnup.

Response: The NRC agrees with this

comment for clarification of values. TS

Table 2.1.1–1 has been revised to use

the terms Minimum Initial Enrichment

and Maximum Burnup. Footnotes

clarifying that the actual minimum

enrichment is to be rounded down and

burnup is to be rounded up were also

added to the Table. Additionally, a

discussion related to the footnotes was

added to the bases for the TSs (B2.1.1)

located in Chapter 12 of the SAR.

Comment E–8: One commenter asked

for clarification on whether the cask

could be put in the pool for 30 days or

only 7 days when cask cavity drying

pressure could not be established within

limits, and if so, why.

Response: TS 3.1.1 provides the

requirements for cask cavity vacuum

drying. The action statements are to be

implemented when a condition

requiring entry into the ACTIONS

exists. The action statements for this TS

provide for interim cooling of the fuel

and basket by establishment of a

nominal helium environment if vacuum

drying was not completed within the

specified time. A 7-day limit to unload

fuel is applicable if a nominal helium

environment is not achieved. A longer,

30-day limit to unload fuel is applicable

when a nominal helium environment

has been achieved. These time limits

provide time to take reasonable

measures to complete fuel unloading

while minimizing the time duration that

the fuel is in a condition other than that

required for long term storage. A

complete discussion is provided in the

bases for this TS.

The time limits do not imply how

much time the cask must spend in the

pool. The actual amount of time the

cask is in the pool is a site-specific issue

and beyond the scope of this rule.

However, when the cask is returned to

the pool and the lid is removed, the

water surrounding the fuel will provide

adequate cooling.

Comment E–9: One commenter stated

that an example 1.4–3 of an ‘‘otherwise

stated’’ exception to the applicability to

the surveillance required by Limiting

Condition for Operation (LCO) 3.1.6

should be added to the TS.

Response: NRC disagrees with this

comment. The existing examples of

Section 1.4 provide sufficient

clarification for the correct

interpretation of the TSs. These

examples were developed as part of the

Improved Standard Technical

Specifications initiative through

extensive interactions between the NRC

staff and industry representatives. TS

3.1.6 clearly indicates when the

surveillance requirement applies, and

no additional explanation is considered

necessary.

Comment E–10: One commenter

stated that on page 3.1.1–1 of the TS,

LCO 3.1.1 requires, ‘‘* * * from

pumping station.’’ For consistency in

terminology, ‘‘pumping’’ should be

changed to ‘‘vacuum drying’’.

Response: The NRC agrees with the

comment and the TS has been changed

to ‘‘vacuum drying’’.

Comment E–11: One commenter

stated that on page 3.1.1–2 of the TS, SR

3.1.1.1 should be changed from ‘‘ * * *

at least 30 minutes’’ to read, ‘‘Verify that

the equilibrium cask cavity vacuum

drying pressure is brought to ≤ 4 mbar

absolute for ≥ 30 minutes.’’

Response: The NRC agrees that the

comment adds clarity and has changed

the TS to ‘‘≥ 30 minutes.’’

Comment E–12: One commenter

stated that on page 3.1.2–1 of the TS, the

Required Action and Completion times

for LCO 3.1.2 are provided without

technical basis and should be revised.

The commenter further stated that on

page 3.1.2–2 of the TS, the Frequency

for SR 3.1.2.1 should be changed from

42 to 48 hours.

Response: The NRC disagrees with

this comment. The heatup analysis

provided by the cask applicant only

supports a 48-hour elapsed time from

the completion of cavity draining to

completion of helium backfill. The

completion time of the SR in 42 hours

allows time (6 hours) to implement

action A.1 if the SR is unsatisfactory.

Action A.2 allows 48 additional hours

to troubleshoot/repair and reperform the

SR provided A.1 is also completed. The

SAR, Section 4.6.2, TS Bases B 3.1.2,

and the SER provide the technical basis,

which shows that the vacuum drying

and helium backfill must be completed

within 48 hours to maintain cask

component temperatures below their

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24863

allowable temperature limits. The

commenter provided no technical basis

supporting additional time for

completion of the helium backfill and

allowance of time to implement

appropriate corrective actions as

outlined in the action.

Comment E–13: One commenter

stated that on page 3.1.5–1, all

conditions and required actions have

not been identified.

Response: The NRC disagrees with the

comment. It is the intent of the TSs to

specify the minimum requirements for

safe operations and the required actions

if the minimum requirements are not

met. A complete discussion on TS use

and application is provided in TS 1.0.

The bases of TS 3.1.5 addresses

investigation of the cause of the low

pressure condition. If the investigation

finds that the cause of the low pressure

condition is leakage above the allowable

limit, then the appropriate TS action for

this condition would also be

implemented. Comment E–14: One commenter

stated that on page 3.1.5–2 of the TS, the

Frequency of SR 3.1.5.2 should be

changed from ‘‘Once, within 7 days of

commencing STORAGE OPERATIONS

and every 36 months thereafter’’ to read,

‘‘Once, within 7 days of commencing

STORAGE OPERATIONS AND 36

months thereafter.’’

Response: NRC agrees with the

comment. To make the format of the

surveillance requirements consistent,

the Frequency statement has been

revised to read, ‘‘Once, within 7 days of

commencing STORAGE OPERATIONS

AND 36 months thereafter.’’

Comment E–15: One commenter

asked if the cask can weep and has this

been verified on a real cask.

Response: No TN–68 casks have been

loaded and none have been tested for

weepage. However, the TN–32 casks are

of very similar design, and these casks

have been loaded at two reactor sites.

Slight weepage has occurred, but has

not caused a problem with cask

handling and storage. The TN–68 casks

must be below the surface

contamination levels in TS 3.2.1 before

they can be moved to the storage pad. Comment E–16: One commenter

stated that the frequency for

Surveillance Requirement 3.1.3.1

should read, ‘‘Once prior to

TRANSPORT OPERATIONS.’’ Two

commenters stated that the frequency

for Surveillance Requirement 3.1.4.1

should read, ‘‘Once prior to

TRANSPORT OPERATIONS OR Once

within 48 hours of commencing

STORAGE OPERATIONS.’’

Response: TS surveillance

requirement 3.1.3.1 currently states

‘‘Once, prior to TRANSPORT

OPERATIONS,’’ therefore no change is

required. For TS surveillance

requirement 3.1.4.1, the NRC agrees

with the comment to revise the

frequency requirement for clarification

as follows: ‘‘* * * OR Once within 48

hours of commencing STORAGE

OPERATIONS.’’ The affected TSs have

been revised as indicated.

Comment E–17: Two commenters

stated that the frequency of Surveillance

Requirement 3.1.6.1 of the TS should be

revised from ‘‘Once, after lifting cask’’

and prior to cask transfer to or from

ISFSI’’ to ‘‘prior to lifting the cask’’. Response: The NRC agrees with this

comment. It is acceptable to perform the

surveillance requirement before lifting

the cask. The TS frequency requirement

of SR 3.1.6.1 has been changed to state

‘‘Once, immediately prior to lifting the

cask and prior to cask transfer to or from

ISFSI.’’.

Comment E–18: One commenter

asked why 200 gallons of fuel in the

transporter is the limiting factor for fire

and explosions in the site-specific

parameters. The commenter states a

plane crash into a full cask array with

a full fuel load should be evaluated.

Response: The NRC disagrees with

this comment. The 200 gallons of fuel

for the fire accident is based on the

amount assumed to be carried by the

transporter. The fire duration for 200

gallons of fuel is 15 minutes. The

analyzed fire is assumed to burn at

1550° F and is assumed to produce the

worse case scenario of fire/heated air for

the TN–68. The fire is assumed to fully

engulf the cask, thus maximizing the

heat input into the cask. Fire of this

duration exposed to the outside of the

cask would have little effect on the cask

or its contents due to the thermal inertia

of the cask.

Before using the TN–68 casks, the

general licensee must evaluate the site

to determine whether or not the chosen

site parameters are enveloped by the

design bases of the approved cask as

required by 10 CFR 72.212(b)(3).

Included in this evaluation is the

verification that the credible sources of

an external explosion do not produce an

external pressure above 25 psi and that

any cask handling equipment used to

move the TN–68 cask to the pad is

limited to 200 gallons of fuel (refer to TS

4.3.5—Site Specific Parameters and

Analyses). Also, when a general licensee

uses the cask design, it will review its

emergency plan for effectiveness under

10 CFR 72.212. This review will

consider interdiction and remedial

actions to address accidents of all types

and coordination with local emergency

response teams. Comment E–19: One commenter

stated that within LCO 3.2.1b, the

values should read 20 dpm/100cm2

instead of 20 dpm/cm2. Response: The NRC agrees with this

comment. This was a typographical

error and LCO 3.2.1b has been

corrected.

Comment E–20: Two commenters

stated that LCO 3.2.1 would require

entry in the action as soon as loading

operations commenced, and that the

applicability for LCO 3.2.1 should be

changed to ‘‘During TRANSPORT

OPERATIONS.’’ One commenter stated

that if the applicability is not changed,

a note should be added to CONDITION

A to clarify the intent of the

specification. The other commenter

stated that the applicability of LCO

3.2.1, the required action, and the

completion time do not adequately

address the retrieval of a cask from an

ISFSI to the spent fuel pool to unload

the cask, and that SR 3.2.1.1 should be

performed before moving a cask from

any restricted area.

Response: Action under LCO 3.2.1 is

not necessary until the contamination

surveillance has been completed.

Transport of the cask to the ISFSI

storage pad cannot begin until the cask

surface is below the decontamination

limit. The surveillance requirement is

part of the loading phase. A note has

been added to LCO 3.2.1 and to the

basis for the TS (B3.2.1) located in

Chapter 12 of the SAR which states that

CONDITION A is not applicable until

after the surveillance for surface

contamination has been completed.

Regarding cask retrieval and

unloading, the primary focus of LCO

3.2.1 is to maintain radioactive

contamination and associated personnel

exposures As Low As Reasonably

Achievable (ALARA). The timing and

nature of specific corrective actions are

determined by the cask user under the

user’s radiation protection programs,

other relevant programs, and applicable

regulations, including 10 CFR part 20,

subpart C, Occupational Dose Limits.

Decisions on unloading a cask will be

made on a case-by-case basis if

appropriate decontamination can not be

achieved.

Comment E–21: One commenter

stated that on page 4.0–3 of the TS : the

title and first paragraph should be

changed from site specific to ISFSI

specific for clarity; item 3 should be

changed to state, ‘‘Seismic loads on the

ISFSI pad * * *’’; and engineered

features to reduce radiation exposure

should be classified as ‘‘not important to

safety.’’

Response: The NRC agrees with

comments 1 and 2. The terminology in

TS 4.0.3 has been revised to indicate

‘‘ISFSI * * *’’ in the title and the first

paragraph since this is a general license

that is not site-specific. Item 3 has been

revised to state ‘‘Seismic loads on the

ISFSI pad * * *’’ The third comment

on engineered features is addressed in

the response to comment E–30.

24864 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

Comment E–22: One commenter

stated that the TS indicates that the cask

cavity vacuum drying process

evaporates any water that has not

drained from fuel or basket surfaces.

The commenter expressed concern

about water not on the specified

surfaces and asked what in the cask,

including the cask materials, has or

could also contain water.

Response: In preparation for dry

storage, the loading process ensures the

removal of virtually all moisture and

oxidizing gases (less than 1 gram-mole

per cask) from the fuel cladding, any

fuel that may have pinholes or hairline

cracks, and from the cask internals. The

cask internals do not provide any

locations for significant moisture

entrapment. The cask is thoroughly

vacuum dried, as prescribed in the TSs

and the SAR. The vacuum drying

process, which involves two, complete,

evacuate-fill cycles, coupled with the

heat generation of the fuel, very

effectively removes residual moisture

that may be present in the fuel pellets

and interior components of the cask

system and oxygen that is inside the

cask. The helium fill gas is very pure

and dry and the cask is sealed to

prevent entry of water and air during

storage. The effectiveness of the vacuum

drying process, the sources of residual

impurities, and the potential effects of

impurities, are reported in PNL–6365,

‘‘Evaluation of Cover Gas Impurities and

Their Effects on the Dry Storage of LWR

Spent Fuel’’ November 1987.

Comment E–23: The commenter asked

what is BorALYNTM, borated wrought

aluminum, and other envisioned

alternate neutron absorber materials,

and if NRC has read the manufacturers’

descriptions as to what is in these

materials, their limitations for use, and

their reactions with other materials.

Response: BorALYNTM is a trademark

for a ceramic of boron carbide particles,

which are produced using natural

boron, e.g., boron containing the

isotopic mix found in nature. In

BorALYNTM, these particles are in a

matrix (formed mechanically with heat

and pressure) of a common and widely

used aluminum alloy. NRC has visited

the plant where this product is

produced to review details on the

process used to produce BorALYNTM.

NRC has required the applicant to do

extensive durability testing of the

material. NRC has reviewed the results

of these tests and found this material to

be acceptable for this application.

Borated aluminum is a wrought

aluminum alloy (made from the liquid

state) that uses an enriched boron as an

alloy addition to the alloy. Natural

boron contains a high-cross section

isotope called 10B, that is many times

more effective at capturing thermal

neutrons than 11B, the other isotope of

boron. The neutron absorber must

capture thermal neutrons during loading

and unloading operations. Enrichment

refers to the concentration of 10B.

Other alternative neutron absorber

materials are like the BorALYNTM and

the borated aluminum, except that they

are made with slight variations, e.g., the

base material is stainless steel in one

case, the boron carbide particles are a

different size in another case, etc. All

materials approved for use are materials

sufficiently nonreactive as to be suitable

for the environments that the materials

must tolerate well in service conditions

for normal, off-normal, and accident

conditions. None of these absorber

materials have special limitations in

relation to the function that they must

perform in the cask systems for which

they have been approved.

Comment E–24: The commenter

stated that any material encased or

welded inside another may either

expand or contract with the heat in the

cask, or react chemically if residual

water remains.

Response: Encased material may

expand and contract relative to

temperature changes. Thermal

expansion/contraction of cask

components was evaluated in the TN–

68 SAR Section 3.4.4.2. This evaluation

was acceptable to the NRC. See the

response to comment E–22 regarding

moisture in the cask cavity.

Comment E–25: The commenter

expressed concern about water leaking

into encased areas if a cask is allowed

to remain in a pool for seven or more

days, and asked if the casks are really

leak tight, citing the port vent and drain

hole areas specifically. The commenter

also asked if leak tightness has been

checked and how the cask is checked

for water retention after soaking for the

seven days.

Response: See the response to

comment E–22 regarding moisture in

the cask cavity. The remainder of the

cask is designed to preclude water

intrusion and retention for the purposes

of decontamination. For example, the

shell that encases the radial neutron

shield is sealed and leak tested after

fabrication as described in SAR Section

9.1.2. If water contacts the polymeric

resins, they are not expected to react

with the water, nor are the metals

expected to react to any extent that

could affect safety of the system. The

vent and drain port areas as well as the

seal areas are thoroughly dried during

preparation for storage.

Comment E–26: One commenter

asked why seven days is allowed to

reflood the cask and unload the fuel

when a nominal helium environment

cannot be achieved. The commenter

noted that the cask can go into the pool

for 30 days when the drying pressure

limits cannot be achieved, and also

asked why one limit is for seven days

and one is for 30 days.

Response: TS 3.1.1 provides the

requirements for cask cavity vacuum

drying. The action statements are to be

implemented when a condition

requiring entry into the ACTIONS

exists. The action statements for this TS

provide for interim cooling of the fuel

and basket via establishment of a

nominal helium environment if vacuum

drying was not completed within the

specified time. A 7-day limit to unload

fuel is applicable if a nominal helium

environment is not achieved. A longer

30-day limit to unload fuel is applicable

when a nominal helium environment

has been achieved. These time limits

provide for reasonable measures to

complete fuel unloading while

minimizing the time duration that the

fuel is not in a suitable long-term

storage condition. A complete

discussion is provided in the bases for

this TS.

The time limits do not imply how

much time the cask must spend in the

pool. The actual amount of time the

cask is in the pool is a site-specific issue

and beyond the scope of this rule.

However, when the cask is returned to

the pool and the lid is removed, the

water surrounding the fuel will provide

adequate cooling.

Comment E–27: One commenter

stated that the cell opening and boron

loading should be removed from Section

4.1.1 of the TS.

Response: The NRC disagrees with the

comment. Design features that may

affect safety if altered or modified are

included in the TS. As stated in SAR

Section 6.1, the TN–68 cask design

parameters relied upon for criticality

safety control are the fuel assembly

spacing and the use of the neutron

absorbing plates. This design

information is crucial to the conclusions

reached by the NRC staff in its SER.

Design tolerances considered in the SAR

for the boron loading and the cell

opening for the basket are included in

the TS limits.

Comment E–28: One commenter

stated that Section 4.1.3, Codes and

Standards, should be removed from the

TSs.

Response: The NRC disagrees with the

comment. This information is required

under 10 CFR 72.24(c)(4).

Comment E–29: One commenter

stated that in the Storage Location for

Casks, 4.2.1 of the TS, the 16-foot

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24865

dimension should be listed as a

minimum value or a tolerance should be

added.

Response: The NRC disagrees with

this comment. As written, the TS states

that ‘‘the casks shall be spaced a

minimum of 16 feet apart, center-tocenter.’’ This specification assures that

the minimum cask spacing assumed in

the analysis is achieved to allow proper

dissipation of radiant heat energy.

Comment E–30: One commenter

stated that references to consideration as

important to safety, be removed from

Section 4.3.6 of the TS.

Response: The NRC disagrees with

this comment. As defined in 10 CFR

72.3, structures, systems, and

components important to safety are

those features of the ISFSI or MRS

whose function is to maintain the

conditions required to store spent fuel

safely. Thus, when a berm or other

system, structure, or component is

installed to meet the normal condition

dose limits of 10 CFR 72.104 (i.e., to

provide safe storage), it is considered

important to safety. However, under 10

CFR 72.122, the quality standards for

the feature’s design, fabrication,

erection, and testing may be at a level

commensurate with the safety

importance of the function to be

performed. In general, features that are

not needed to meet the accident

conditions will not have to meet as high

a standard as those that need to function

in an accident.

Comment E–31: One commenter

stated that on pages 5.0–3 through 5.0–

5 of the TS, describing the cask surface

dose rate evaluation program,

inconsistent terminology is used

regarding the neutron shielding. A

single term ‘‘radial neutron shield’’

should be used consistently.

Response: The NRC agrees with this

comment. In the interest of clarity, TS

5.2.3 has been revised to consistently

use the term ‘‘radial neutron shield’’

where appropriate.

Comment E–32: One commenter

stated that on page 5.0–5 of the TS, the

reference to Figure 5.2.3–1 should be

deleted.

Response: The NRC disagrees with

this comment. Figure 5.2.3–1 is

provided as a quick reference for the

user and the public to help interpret the

measurement locations given in TS

5.2.3.7. The figure is an illustration, not

to scale, and the specification wording

more exactly defines the location of

each measurement.

Comment E–33: One commenter

stated that the NRC did not clearly state

why the interior cannot be preferentially

or unevenly flooded and asked why the

NRC did not analyze the scenario of a

cask partially filled with unborated

water and steam.

Response: As stated in SAR Section

6.1, nonuniform flooding of the basket

is not credible because all spaces in the

basket are interconnected. The applicant

evaluated the failure of the four center

basket cavities to drain and showed that

this was significantly less reactive than

a fully flooded cask. As stated in SER

Section 6.3.1, the applicant varied the

water density in the cask to bound any

possible density changes during loading

and unloading operations. The full

density water resulted in the highest

reactivity in all cases.

Comment E–34: One commenter

asked which fuel assembly has the

highest reactivity; 7x7 GE2, GE2b, or

10x10. Further, the commenter asked

why the NRC does not have a third

party verify both the NRC’s and

applicant’s calculations.

Response: As shown in SAR Table

6.4–3, the applicant evaluated both the

7x7 and 10x10 assemblies for all

normal, off-normal, and accident

conditions. The results in this table

show that the 10x10 assembly is the

most reactive under the most bounding

conditions. Because the NRC staff has

reasonable assurance that the cask meets

the design criterion for criticality safety,

further verification by a third party is

not required. s

Comment E–35: One commenter

stated that on page 3–17 of the SER,

reference 4 should be changed to,

‘‘ANSI N14.6, Special Lifting Devices

for Shipping Containers Weighing

10,000 Pounds or More for Nuclear

Materials, 1986.’’ The commenter also

stated that on page 9–8 of the SER,

reference 5 should be changed from

ANSI N14.6–1993 to ANSI N14.6–1986.

Response: The NRC disagrees with

this comment. ANSI N14.6–1993 was

used by the NRC staff in this evaluation.

Comment E–36: One commenter

stated that on page 4–9 of the SER, the

second sentence in the first paragraph

under Section 4.5.2.4 should be changed

to, ‘‘Assuming design basis heat load

fuel and completion of cask cavity

drying, helium backfill should be

completed within 48 hours.’’ This

change is needed to conform to TS 3.1.2.

Response: The NRC disagrees with

this comment. The heatup analysis

provided by the cask applicant only

supports a 48-hour elapsed time from

the completion of cavity draining to

completion of helium backfill. The

commenter did not provide a technical

basis supporting an additional 48 hours.

Comment E–37: One commenter

stated that on page 5–3 of the SER, the

use of spectral shift void history on

early design fuel (7x7) by TN provides

considerable conservatism and should

be reconsidered.

Response: The NRC disagrees with

this comment. The analysis provided to

support a general license design, which

applies to all licensees, needs to bound

all variations of cask contents unless

compensating factors are present. The

operational parameters assumed to

determine the source term in the design

basis fuel need to cover the range of

both current and past operating

practices of all authorized users.

Comment E–38: One commenter

stated that in Table 7–1 of the SER, the

percentage of rods that failed in offnormal and accident conditions are not

consistent with industry experience and

research. More reasonable values are on

the order of 0.0001% and 0.01% for offnormal and accident conditions

respectively.

Response: The rod breakage fractions

presented in Table 7–1 of the SER were

based on those already contained in

NUREG–1536, ‘‘Standard Review Plan

for Dry Cask Storage Systems’’ as

discussed on page 2–7. This NUREG

was previously subject to public

comment. Currently, the NRC is

confident that the rod breakage fractions

are bounding and provide reasonable

assurance of public safety with regard to

the confinement analyses of spent fuel

storage casks. Further, NRC and

industry initiatives to modify

assumptions for rod breakage fractions

are beyond the scope of this rule.

Comment E–39: One commenter

stated that in Table 7–1 of the SER, the

meteorological conditions to be used to

analyze the offsite dose consequences

should be consistent with those used for

the power plant.

Response: The NRC disagrees with

this comment. Since the meteorological

conditions for a specific site are not

known, the NRC has made bounding

assumptions for meteorological

conditions to establish a basis for cask

approval. General licensees who use a

cask approved under 10 CFR 72, subpart

L, must calculate dose equivalents for

their ISFSIs, considering site-specific

meteorology, other exposure pathways

such as ingestion and ground

deposition, and actual distances to the

site boundary.

Comment E–40: One commenter

stated there should only be a TEDE limit

in Table 7–2 of the SER and that the

calculation of other doses is redundant.

Response: The NRC does not agree

with this comment. Whole body (TEDE)

and organ dose limits are required in 10

CFR 72.104 and 10 CFR 72.106. Also, 10

CFR 72.106 provides dose limits on skin

and the lens of the eye. Therefore,

24866 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

evaluation of these doses is needed for

cask approval.

Comment E–41: One commenter

stated that on page 8–4 of the SER, the

last paragraph in Section 8.3.2 refers to

a check valve to restrict cooling water

flow if cask pressure exceeds 90 psia. A

pressure control valve would provide

the desired capability.

Response: The NRC agrees with the

comment that either valve will satisfy

the requirement to restrict flow. The

SER Section 8.3.2 has been changed to

reflect that a valve designed to restrict

flow will act to restrict cooling water

flow if cask pressure exceeds 90 psia,

which will allow flexibility by the cask

user. The SAR has also been revised by

the applicant to reflect this change.

Comment E–42: One commenter

stated that on page 10–3 of the SER, the

last paragraph under Section 10.3.1

should be deleted.

Response: The NRC disagrees with

this comment. As defined in 10 CFR

72.3, structures, systems, and

components important to safety are

those features of the ISFSI or monitored

retrievable storage installation (MRS)

whose function is to maintain the

conditions required to store spent fuel

safely. Thus, when a berm or other

system, structure, or component is

installed to meet the normal condition

dose limits of 10 CFR 72.104 (i.e., to

provide safe storage), it is considered

important to safety. However, under 10

CFR 72.122, the quality standards for

the feature’s design, fabrication,

erection, and testing may be at a level

commensurate with the safety

importance of the function to be

performed. Therefore, the last paragraph

is necessary.

Comment E–43: One commenter

stated that on page 11–1 of the SER, the

last sentence under Section 11.0 should

be changed from SAR Revision 4 to SAR

Revision 5.

Response: The NRC agrees with this

comment and has updated page 11–1.

F. Comments on Applicant’s Topical

SAR

Comment F–1: One commenter stated

that on page 8.1–3 of the SAR, the first

sentence of the description for the cask

transporter should be changed to read,

‘‘The cask transporter is generally set to

limit the lift height of the cask to ensure

that the maximum gravitational loading

force limit in the event of a cask drop

is met.’’

Response: The NRC agrees with the

comment with additional clarification.

The SAR has been revised to state: ‘‘The

cask transporter is set to limit the lift

height of the cask to ensure that the

loads from a postulated drop accident

will be bounded by the maximum

analyzed loads given in Technical

Specifications 4.1.2 and 5.2.2.’’

Comment F–2: One commenter stated

that drawing 972–70–1 of the SAR

should be revised to add a tolerance of

+0/¥.25 to 13.25-inch dimension to

accommodate variations due to welding.

Response: The NRC accepts this

change to the tolerance specified on

Drawing No. 972–70–1 because it will

not affect the structural analyses and the

conclusions reached in the SER.

Drawing No. 972–70–1 has been

changed accordingly.

Comment F–3: One commenter stated

that drawing 972–70–4 of the SAR

should be revised to add note 6 to allow

the clearance hole in the rail at the end

to be optional. The size of the clearance

hole should be increased from a 2.00-

inch diameter to a 3.56-inch diameter to

allow sufficient clearance for a socket

wrench.

Response: The NRC accepts these

changes to the clearance hole in the rail

because they will not affect the

structural analyses and the conclusions

reached in the SER. Drawing No. 972–

70–4 has been changed accordingly.

Comment F–4: One commenter stated

that Note 2 on drawing 972–70–5 of the

SAR should be revised from ‘‘PT

examination per ASME Section III,

Subsection NG–5231’’ to ‘‘PT

examination per ASME Section III,

Subsection NG–5233.’’

Response: The NRC disagrees with the

comment that Note 2 on Drawing 972–

70–5 needs to be changed. For thin, onelayer welds without filler material,

ASME Section III, Subsection NG–5231

is still applicable. For clarification of

the nondestructive examination

requirement in NG–5231, Table 4.1–1 of

the TSs has been revised.

Comment F–5: One commenter stated

that drawing 972–70–6 of the SAR

should be revised to add a note to allow

alternate plumbing configurations. Also,

an additional connection may be

required through the protective cover

for helium leak testing of the over

pressure (OP) system.

Response: The NRC agrees with this

comment. Alternate plumbing

configurations will add flexibility to the

design of the OP system without

adversely affecting the structural

analyses and the conclusions reached in

the SER. The note should also state that

the parts and equipment used are

equivalent to those specified in the

drawing. An adequate level of safety is

obtained by the quality assurance

process, plus the leak testing and

monitoring of the system as required by

the TSs. The addition of a test fitting in

the protective cover does not affect

safety because its purpose is to facilitate

leak testing of the overpressure

monitoring system. Drawing No. 972–

70–6 has been revised to reflect these

changes.

Comment F–6: One commenter stated

that it is not possible to perform PT on

the Plasma-Arc Welding (PAW) part of

the weld since the Gas Tungsten-Arc

Welding (GTAW) is part of the

automatic welding equipment.

Transnuclear has proposed a code case

to Section III, Subsection NG, on this

issue for guidance.

Response: The NRC agrees with the

applicant’s view that inspection after

PAW is not practical and that inspection

after GTAW is adequate. The proposed

code case is beyond the scope of NRC

review.

G. Accidents

Comment G–1: One commenter asked

if a cask will slide on the pad and could

slide into other casks or other structures

in the independent spent fuel storage

installation (ISFSI), stated that the pad

was described in a site-specific manner

instead of generically, and asked what

structures or vehicles are permitted to

be within the ISFSI fence.

Response: The SAR indicates that the

cask may slide 7.3 inches due to a 4,000

lb. missile (in this case, an automobile)

impacting below the center of gravity of

the cask at 126 mph. This is much

smaller than the approximately 94-inch

distance between casks. Therefore,

impacts between TN–68 casks on the

pad would not occur. In the unlikely

event that two 4,000 lb missiles were to

impact below the center of gravity of

two adjacent casks from opposite

directions at the same time, the two

casks still would not collide with each

other. Furthermore, the automobile is

conservatively assumed to be rigid and

absorbs no energy in the analysis. In an

actual impact, the majority of the energy

will be absorbed by the crushing of the

automobile rather than moving of the

cask. The pad is a site-specific issue that

needs to be addressed in the cask user’s

10 CFR 72.212 evaluation. TS 5.2.1,

referenced by the commenter, simply

requires the cask user to verify that the

coefficient of friction for the concrete

pad matches the coefficient of friction

used in the SAR’s cask sliding analysis.

The structures and vehicles permitted

within the ISFSI fence is a site-specific

issue and is beyond the scope of this

rule.

Comment G–2: One commenter stated

that all things in loading and unloading

areas should be evaluated for a cask

drop or tip over accident.

Response: This comment is beyond

the scope of this rule. The use of a

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24867

generally licensed cask by a utility

requires that the user ensure that the

site is not subject to any potential

accident that has not been analyzed for

the general license.

Comment G–3: One commenter noted

that explosive overpressure is not

addressed, stated this should be done

now and should have been done before

the SER was completed, and asked why

it was not addressed. They stated that

this evaluation is not suitable for a sitespecific evaluation and should be

addressed as part of the generic review.

The commenter suggested that a

sabotage explosion such as a truck bomb

ramming the fence or a plane explosion

needs evaluation for current cask

approval.

Response: NRC disagrees with this

comment. The TN–68 is designed to

withstand an external pressure of 25 psi.

This would include a nearby explosion,

debris falling on the cask, etc. If a

credible explosion is identified that

would apply more than 25 psi to the

outer surface of the cask at a site, the

site will have to address this issue in its

10 CFR 72.212 evaluation.

Comment G–4: One commenter stated

that earthquake analysis should not rely

on site analysis for the nuclear power

plant because the analysis for the plant

does not apply to the pad, and the plant

and pad are not on the same soil

location.

Response: The NRC disagrees with the

recommendation that each ISFSI pad be

required to have a specific seismic

analysis. This is beyond the scope of

this rule. The licensee using a particular

cask design has the responsibility under

the general license to evaluate the match

between reactor site parameters and the

range of site conditions (i.e., the

envelope) reviewed by the NRC for an

approved cask. The licensee should also

consider if there are any site conditions

associated with the actual pad and cask

locations that could affect cask design

and that were not evaluated in the NRC

SER for the cask.

Comment G–5: One commenter stated

that the effects of lightning need to be

evaluated.

Response: The effects of lightning are

addressed in Section 2.2.5.2.8 of the

SAR. Section 3.1.2.1.8 of the SER has

been revised to clearly indicate this fact.

Comment G–6: One commenter asked

if there is a more recent reference

document than the 1974 document

referenced in the CoC that addresses

tornadoes.

Response: The document referenced

in the CoC that addresses tornadoes is

a Regulatory Guide entitled ‘‘Design

Basis Tornado for Nuclear Power

Plants.’’ There has been no revision on

this Regulatory Guide after the 1974

publishing date.

Comment G–7: One commenter asked

why the lid is not modeled for

maximum temperature in storage

conditions and the cask bottom is not

modeled for peak temperature in a fire

accident.

Response: The cask lid will perform

its intended safety function

(confinement) for the normal conditions

of storage. The cask bottom will perform

its intended safety function

(confinement) for the fire accident.

Based on the applicant’s modeling and

analysis which demonstrated that there

was no challenge to the safety functions

of these components, explicit modeling

of these components in the conditions

specified by the commenter was not

required.

Comment G–8: One commenter asked

if an emergency plan had been

developed to retrieve a buried cask, how

a TN–68 cask would be excavated in the

most efficient and rapid way, and has

this been evaluated. The commenter

asked if emergency staff at the site and

in the nearby communities are trained

to deal with cask fires, how training is

administered, and if anyone oversees

the training to ensure that it is effective.

Response: Cask general licensees are

required by 10 CFR 72.212 (b)(6) to

evaluate their emergency plans and

revise them accordingly before using a

cask certified under 10 CFR 72 subpart

L. The details of site specific emergency

response are beyond the scope of this

rulemaking.

H. Radiation Protection

Comment H–1: One commenter had

questions about radiation in a full cask

array, particularly how the radiation or

skyshine from casks of the same design

and casks of different designs affect

each other and if research has been done

to evaluate the effects. The commenter

also asked if surface dose rates should

be taken again at the pad after the casks

have been moved to the pad. The

commenter also asked where most

loaded casks are presently located.

Response: The shielding analysis

addresses the interaction of radiation

between the casks of the same design in

a storage array. The interaction between

casks of different designs is not a part

of this rule, but is not expected to be

significantly different than that

considered in the original analysis. As

a final check, each user of a storage cask

must perform a site-specific analysis to

show that the regulatory dose limits will

be met at the user’s site including the

effects of other cask designs if present.

For the purposes of TS 5.2.3, a second

dose rate measurement is not needed

after the cask has been moved to the

storage pad. The normal and accident

condition analyses of the cask show that

the dose rates are not expected to

change during transport to the storage

pad. However, the licensee’s radiation

protection program will include general

area measurements at the pad.

The Oconee reactor site has the largest

number of loaded dry storage casks.

Comment H–2: One commenter stated

that Figure 5.2.3–1, which shows

contact dose rate measurement

locations, should be changed to show

the cask trunnions.

Response: The NRC disagrees with

this comment. Figure 5.2.3–1 is

provided as a quick reference for the

user and the public to help interpret the

measurement locations in TS 5.2.3.7.

Measurement locations with respect to

the trunnions are contained in the

specification. The exact location of the

trunnions is shown in the SAR

drawings.

Comment H–3: One commenter asked

where Hansen couplings, basket key,

basket rail shims, security wire and

seals, and alignment pins are located on

Figure 1–1 of the SER. The commenter

also asked why Figure 1–1 of the SER

does not show the gamma shield. The

commenter stated that the figure also

should better depict where the outer

neutron shield is installed, and asked if

the outer neutron shield stops above the

bottom trunnion and below the top

trunnion or goes around them. The

commenter stated that the outer shell

design is very unclear and that a better

drawing is required.

Response: The NRC disagrees that a

more detailed drawing is required in the

SER. Figure 1–1 is only intended to

depict the general configuration of the

cask. The applicant’s SAR includes

drawings and design detail that enable

the NRC to make a safety finding. That

same level of detail does not need to be

repeated in the SER, because the SAR

drawings are available on the docket

and are retrievable by the NRC staff and

the public. The neutron shield runs the

full length of the active fuel region of

the fuel assemblies which is the location

of the neutron source term, extending

from below the bottom trunnion to halfway around the top trunnion.

Comment H–4: One commenter stated

that a date should be provided for

reference 5 on page 4–12 and for

reference 3 on page 6–8 of the SER, and

that the NRC should add dates to all

references as regular practice.

Response: Typically, computer codes

are listed by version and not by date

(e.g., version 4.3, 4.4, etc.). ANSYS

Version 5.4 was released in September,

1997. SCALE Version 4.4 was released

24868 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

in September, 1998. These dates were

added to the SER.

Comment H–5: One commenter

requested that the NRC clarify why the

1-inch thick steel shell above the radial

neutron shield is optional.

Response: As stated in TS 5.2.3, the

1-inch thick shield does not need to be

installed if it is not needed to meet the

surface dose rate limits in the

specification. The surface dose rate

limits were taken from the shielding

analysis.

Comment H–6: One commenter stated

that the discussion on Page 5.2 of the

SER concerning cobalt impurities in

stainless steel is vague and is based on

unrelated documents. Further, the

commenter asked how much cobalt

impurity can vary based on supplier and

date of manufacture and how a

fabricator knows what is being

provided.

Response: The NRC disagrees that the

documents are unrelated. The references

are widely used reports produced by

national laboratories and are considered

to be appropriate sources of information

for establishing the assumed cobalt

impurity levels. Early on, cobalt

impurities in fuel assembly hardware

were not as well controlled as today and

could vary; therefore, appropriate

bounding values were established using

the data in the references. After the

effect of tramp amounts of cobalt

became apparent, fabricators and

designers began to specify limits on the

cobalt content in materials procurement

documents. In the last 10 to 15 years,

fabricators typically specify the

acceptable impurity limits as part of

their procurement process subject to the

applicable quality assurance

procedures.

Comment H–7: One commenter had a

number of concerns related to the cobalt

content of stainless steel used in cask

fabrication: What are the tolerance

specifications for the components in the

stainless steel and how varying the

tolerance would affect their

performance; how cobalt affects cask

handling and unloading in any way;

what cobalt data on a specific batch of

stainless steel is reported by the

supplier; and if this should be factored

into analysis each time a new batch is

used.

Response: Thermal (slow) neutrons

are required to activate the cobalt in the

components that make up the storage

cask system. There are essentially no

thermal neutrons that collide with these

components in storage systems.

Therefore, questions concerning the

cobalt in this material are not relevant

in relation to activation. As for

mechanical properties, many if not all

are likely to be enhanced by the

addition of cobalt to the alloy, but this

is not done for economic reasons. The

cobalt might be reported by the supplier

if it was at a high enough concentration

to be detected by the analytical

procedures that are normally used for

chemical analyses of these alloys.

Tramp elements are not always

reported, except by special request.

Therefore, the NRC staff is not

concerned about cobalt in materials

used for these components. See also

comment H–6.

Comment H–8: One commenter stated

concerns relating to how the neutron

source is evaluated taking into account

the natural uranium blankets used in

the BWR fuel that has changed over the

years. The commenter stated that a

utility needs to carefully evaluate

neutron sources to precisely reflect the

fuel age and type that is to be loaded in

casks, that TN erred in computing the

neutron sources in the SAR table, and

asked how an applicant could make

such a mistake and how the NRC could

accept such a mistake. The SAR neutron

source table and its calculations need to

be done correctly and the SAR needs to

be revised to reflect the correct values

before the NRC accepts the document.

Response: Less than 10% of the offsite dose comes from neutrons. Thus,

uncertainties in the neutron source

strength are not significant. A general

license analysis does not need to be

bounding in every term as long as the

overall result is bounding. The NRC

staff’s review determined that the small

underestimation of the neutron source

term was more than compensated for by

the applicant’s overestimation for the

gamma-ray source term. Therefore, the

applicant’s estimated dose from the cask

is bounding. The general license

analysis is based on generalized

operating assumptions. However, each

licensee user must perform a sitespecific analysis to show compliance

with the regulations. The site-specific

analysis is the appropriate place to

address the type, age, and operating

conditions for the actual fuel to be

loaded at the site.

Comment H–9: One commenter asked

how the fuel reacts at the top and

bottom of the cask when exposed to

steam during quenching.

Response: Thermal stress associated

with reflooding and quenching is

discussed in SAR Sections 3.5.2 and

4.6.1. SER Section 4.1 contains the

analysis and NRC acceptance of

quenching effects described in the SAR.

Comment H–10: One commenter

stated a concern with streaming at the

trunnions and asked why detailed

confirming calculations were not

modeled, asked what is the trunnion

material, asked whether the trunnions

should be lowered, and stated that

workers will have to be around the

trunnions adjusting the lifting devices

and that the vendor should work to

reduce unnecessary doses.

Response: The modeling detail of the

trunnions in the shielding analysis is at

a level that equals the capability of the

analytical code. Further detail in the

trunnion calculations is not necessary

because radiation streaming around the

trunnions is very localized and will

have negligible effect on meeting the

regulatory limit for the off-site dose.

Worker doses are subject to ALARA as

discussed in item 4 below. The

trunnions are made of steel with a

central plug of borated polyester resin.

Placement of the trunnions was a design

decision made by the applicant and is

beyond the scope of this rule. The

shielding performance of the trunnion

design has been reviewed and found to

be adequate. The radiation protection

program of the licensee user will have

the responsibility to implement

measures to keep the dose of workers

around the trunnions as low as

reasonably achievable. Any streaming

points will be monitored and avoided

during cask handling operations.

Comment H–11: One commenter

asked why the neutron shield does not

cover the entire cask and if the design

is based on the location of the

trunnions.

Response: Radially, except at the

trunnions, the neutron shield runs the

full length of the active region of the

spent fuel assemblies, that are the

source of neutron radiation. The design

of the neutron shield is based on

meeting the regulatory requirements and

is acceptable.

Comment H–12: One commenter

asked about the ‘‘radiation return from

radial neutron shield’’ reduction of

photon dose from 860 mrem/hr to 749

mrem/hr and why the NRC did not

conduct confirmatory calculations to be

sure that this reduction is correct. The

commenter also recommended that the

NRC should not accept expected values

and should not leave it up to the

licensee to determine how to maintain

doses ALARA, but should instead

provide guidelines as part of the

approval process for this design.

Response: In lieu of performing a

separate accident calculation, the NRC

staff used the results from the normal

conditions calculation to bound the

dose rate at the cask surface. The NRC

staff’s analysis shows good agreement

with the applicant’s calculations. In

addition, the maximum off-site dose

from a cask under accident conditions is

Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations 24869

about one tenth of the regulatory limit.

Even with a higher value of 860 mrem/

hr, the performance of the cask in the

hypothetical accident would be well

within regulatory limits. Guidelines for

a licensee’s ALARA are contained in

Regulatory Guide 8.8, ‘‘Information

Relevant to Ensuring that Occupational

Radiation Exposures at Nuclear Power

Stations will be As Low as Reasonably

Achievable.’’

Comment H–13: One commenter

recommended that an eye lens

calculation be added to Table 7–2 of the

SER so that the effects of radiation dose

to the eye can be known.

Response: The NRC has chosen not to

add an eye lens calculation to Table 7–

2. As discussed in the TN–68 SER,

compliance with the dose-equivalent

limit for the lens is achieved by

demonstrating compliance with the

dose-equivalent limit for the skin and

the effective dose-equivalent limit. This

approach is consistent with guidance in

ICRP–26, International Commission on

Radiation Protection, ‘‘Statement from

the 1980 Meeting of the ICRP,’’ ICRP

Publication 26, Pergammon Press, New

York, New York, 1980.

Summary of Final Revisions

The NRC staff modified the rule

language, the CoC, the TSs, and its SER.

Rule Language Change

The rule language has been modified

to clarify that it is the Certificate that

expires.

CoC Changes

The CoC has been changed for

consistency with other issued

certificates.

TN–68 TS Changes and Associated

Comments

TSs were reformatted into Corel 8

WordPerfectTM software that addressed

the editorial changes in comment B–17.

TS1.1 Definition of Intact fuel was

revised based on the NRC staff’s

initiative.

Table 2.1.1–1 revised labels to add in

minimum and maximum, and added

three footnotes based on comment E–7

and the NRC staff’s initiative.

LCO 3.1.1 was revised to state, ‘‘from

the vacuum drying system’’ based on

comment E–10.

SR 3.1.1.1 was revised to state, ‘‘≤ 4

mbar absolute for ≥ 30 minutes’’ based

on comment E–11.

SR 3.1.4.1 was revised to state, ‘‘Once

within 48 hours of commencing

STORAGE OPERATIONS’’ based on

comment E–16.

SR 3.1.5.1 Frequency has been revised

to state, ‘‘OPERATIONS AND 36 months

thereafter’’ based on comment E–14.

SR 3.1.6.1 Frequency has been revised

to state, ‘‘Once, immediately prior to

lifting cask’’ based on comment E–17.

LCO 3.2.1 b. was revised to state,

‘‘20dpm/100 cm2 ’’ based on comment

E–19, and a note added ‘‘Not applicable

until SR 3.2.1.1 is performed’’ based on

comment E–20.

Table 4.1–1 has been clarified to

address PT examination under ASME

Section III, Subsection NG–5231, based

on comment F–4.

TS 4.3 has been revised to state,

‘‘ISFSI Specific’’ and ‘‘load on the ISFSI

pad’’ based on comment E–21.

TS 5.2.3 has been revised to use the

terminology ‘‘radial neutron shield’’

throughout the section based on

comment E–31.

Agreement State Compatibility

Under the ‘‘Policy Statement on

Adequacy and Compatibility of

Agreement State Programs’’ approved by

the Commission on June 30, 1997, and

published in the Federal Register on

September 3, 1997 (62 FR 46517), this

rule is classified as compatibility

Category ‘‘NRC.’’ Compatibility is not

required for Category ‘‘NRC’’

regulations. The NRC program elements

in this category are those that relate

directly to areas of regulation reserved

to the NRC by the Atomic Energy Act of

1954, as amended (AEA), or the

provisions of the Title 10 of the Code of

Federal Regulations. Although an

Agreement State may not adopt program

elements reserved to NRC, it may wish

to inform its licensees of certain

requirements via a mechanism that is

consistent with the particular State’s

administrative procedure laws, but does

not confer regulatory authority on the

State.

Finding of No Significant

Environmental Impact: Availability

Under the National Environmental

Policy Act of 1969, as amended, and the

NRC regulations in Subpart A of 10 CFR

part 51, the NRC has determined that

this rule, if adopted, would not be a

major Federal action significantly

affecting the quality of the human

environment and, therefore, an

environmental impact statement is not

required. This final rule adds an

additional cask to the list of approved

spent fuel storage casks that power

reactor licensees can use to store spent

fuel at reactor sites without additional

site-specific approvals from the

Commission. 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 Gordon

Gundersen, Office of Nuclear Material

Safety and Safeguards, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, telephone (301) 415–6195,

email GEG1@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 OMB 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 will add the Transnuclear TN–

68 cask system to the listing within the

list of NRC approved casks for spent

fuel storage in § 72.214. This action does

not constitute the establishment of a

standard that establishes generallyapplicable requirements.

Regulatory Analysis

On July 18, 1990 (55 FR 29181), the

Commission issued an amendment to 10

CFR part 72. The amendment provided

for the storage of spent nuclear fuel in

cask systems with 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. In that rule, four spent fuel storage

casks were approved for use at reactor

sites and were listed in 10 CFR 72.214.

That rule envisioned that storage casks

certified in the future could be routinely

added to the listing in 10 CFR 72.214

through the rulemaking process.

Procedures and criteria for obtaining

NRC approval of new spent fuel storage

cask designs were provided in 10 CFR

part 72, subpart L.

24870 Federal Register / Vol. 65, No. 83 / Friday, April 28, 2000 / Rules and Regulations

The alternative to this action is to

withhold approval of this new design

and issue a site-specific license to each

utility that proposes to use the casks.

This alternative would cost both the

NRC and utilities more time and money

for each site-specific license.

Conducting site-specific reviews would

ignore the procedures and criteria

currently in place for the addition of

new cask designs that can be used under

a general license, and would be in

conflict with NWPA direction to the

Commission to approve technologies for

the use of spent fuel storage at the sites

of civilian nuclear power reactors

without, to the maximum extent

practicable, the need for additional site

reviews. This alternative also would

tend to exclude new vendors from the

business market without cause and

would arbitrarily limit the choice of

cask designs available to power reactor

licensees. This final rule will eliminate

the problems above and is consistent

with previous NRC actions. Further, the

rule will have no adverse effect on

public health and safety.

The benefit of this rule to nuclear

power reactor licensees is to make

available a greater choice of spent fuel

storage cask designs that can be used

under a general license. The new cask

vendors with casks to be listed in 10

CFR 72.214 benefit by having to obtain

NRC certificates only once for a design

that can then be used by more than one

power reactor licensee. The NRC also

benefits because it will need to certify

a cask design only once for use by

multiple licensees. Casks approved

through rulemaking are to be suitable

for use under a range of environmental

conditions sufficiently broad to

encompass multiple nuclear power

plants in the United States without the

need for further site-specific approval

by NRC. Vendors with cask designs

already listed may be adversely

impacted because power reactor

licensees may choose a newly listed

design over an existing one. However,

the NRC is required by its regulations

and NWPA direction to certify and list

approved casks. This rule has no

significant identifiable impact or benefit

on other Government agencies.

Based on the discussion above 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 NRC

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

Transnuclear. 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 (§ 50.109 or § 72.62) does

not apply to this direct final rule

because this amendment does not

involve any provisions that would

impose backfits as defined. Therefore, a

backfit analysis is not required.

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, 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. 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. In § 72.214, Certificate of

Compliance (CoC) 1027 is added to read

as follows:

§ 72.214 List of approved spent fuel

storage casks.

* * * * *

Certificate Number: 1027.

SAR Submitted by: Transnuclear, Inc.

SAR Title: Final Safety Analysis

Report for the TN–68 Dry Storage Cask.

Docket Number: 72–1027.

Certificate Expiration Date: May 28,

2020.

Model Number: TN–68.

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. 00–10390 Filed 4–27–00; 8:45 am]

BILLING CODE 7590–01–P

DEPARTMENT OF TRANSPORTATION

Federal Aviation Administration

14 CFR Part 39

[Docket No. 2000–NM–85–AD; Amendment

39–11699; AD 2000–08–13]

RIN 2120–AA64

Airworthiness Directives; Learjet

Model 45 Airplanes

AGENCY: Federal Aviation

Administration, DOT.

ACTION: Final rule; request for

comments.

SUMMARY: This amendment adopts a

new airworthiness directive (AD) that is