[Federal Register Volume 65, Number 83 (Friday, April 28, 2000)]
[Rules and Regulations]
[Pages 24855-24870]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-10390]


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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.

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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, [email protected] 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.

[[Page 24856]]

    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 site-specific 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 Sec. 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, 
Sec. 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

[[Page 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 low-level 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 Sec. 73.51 or Sec. 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 deg.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

[[Page 24858]]

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

[[Page 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.

[[Page 24860]]

    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 Ci/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

[[Page 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, off-normal, 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 fuel-cladding interface and 
the fuel pellet-fuel 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

[[Page 24862]]

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

[[Page 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 deg. 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/100cm\2\ instead of 20 dpm/cm\2\.
    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.

[[Page 24864]]

    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

[[Page 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-to-center.'' 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 off-normal 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 off-
normal 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,

[[Page 24866]]

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, one-layer 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

[[Page 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 site-specific 
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 half-
way 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

[[Page 24868]]

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 off-site 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 site-specific 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

[[Page 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 [email protected].

Paperwork Reduction Act Statement

    This final rule does not contain a new or amended information 
collection requirement subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the 
Office of Management and Budget, Approval Number 3150-0132.

Public Protection Notification

    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 Sec. 72.214. This action does not constitute the 
establishment of a standard that establishes generally-applicable 
requirements.

Regulatory Analysis

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

[[Page 24870]]

    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 (Sec. 50.109 or 
Sec. 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 Sec. 72.214, Certificate of Compliance (CoC) 1027 is added to 
read as follows:


Sec. 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