[Federal Register Volume 65, Number 54 (Monday, March 20, 2000)]
[Rules and Regulations]
[Pages 14790-14810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-6630]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG 18


List of Approved Spent Fuel Storage Casks: TN-32 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-32 cask system to the list of 
approved spent fuel storage casks. This amendment allows the holders of 
power reactor operating licenses to store spent fuel in this approved 
cask system under a general license.

EFFECTIVE DATE: This final rule is effective on April 19, 2000.

FOR FURTHER INFORMATION CONTACT: Merri Horn, telephone (301) 415-8126, 
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, in cooperation with the private sector, for the 
dry storage of spent nuclear fuel at civilian nuclear reactor power 
sites, with the objective of establishing one or more technologies that 
the [Nuclear Regulatory] Commission may, by rule, approve for use at 
the sites of civilian nuclear power reactors without, to the maximum 
extent practicable, the need for additional site-specific approvals by 
the Commission.'' Section 133 of the NWPA states, in part, ``[t]he 
Commission shall, by rule, establish procedures for the licensing of 
any technology approved by the Commission under Section 218(a) for use 
at the site of any civilian nuclear power reactor.''
    To implement this mandate, the NRC approved dry storage of spent 
nuclear fuel in NRC-approved casks under a general license, publishing 
a final rule in 10 CFR Part 72 entitled ``General License for Storage 
of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). 
This rule also established a new Subpart L within 10 CFR Part 72 
entitled, ``Approval of Spent Fuel Storage Casks'' containing 
procedures and criteria for obtaining NRC approval of dry storage cask 
designs.

Discussion

    This rule will add the Transnuclear TN-32 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, ``TN-32 Dry Storage Cask Topical Safety 
Analysis Report (TSAR).'' The NRC evaluated the Transnuclear submittal 
and issued a preliminary Safety Evaluation Report (SER) and a proposed 
Certificate of Compliance (CoC) for the Transnuclear TN-32 cask system. 
The NRC published a proposed rule in the Federal Register (64 FR 45923; 
August 23, 1999) to add the TN-32 cask system to the listing in 10 CFR 
72.214. The comment period ended on November 8, 1999. Four 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 and the Technical 
Specifications (TSs) for the TN-32 cask system. The NRC staff has also 
removed the bases section from the TSs. The NRC staff has modified its 
preliminary SER. The NRC staff has also modified the rule language by 
changing the word ``Certification'' to ``Certificate'' to clarify that 
it is the Certificate that expires.
    The proposed CoC has been revised to clarify the requirements for 
making changes to the CoC by specifying that the CoC holder must submit 
an application for an amendment to the certificate if a change to the 
CoC, including its appendices, is desired. The CoC has also been 
revised to delete the proposed exemption from the requirements of 10 
CFR 72.124(b) because a recent amendment of this regulation makes the 
exemption unnecessary (64 FR 33178; June 22, 1999). The staff has also 
updated the CoC, including the addition of explicit conditions 
governing acceptance tests and maintenance program, approved contents, 
design features, and authorization, and has removed the bases section 
from the TSs attached to the CoC to ensure consistency with NRC's 
format and content. In addition, other minor, nontechnical changes have 
been made to CoC 1021 to ensure consistency with NRC's new standard 
format and content for CoCs.
    The NRC finds that the TN-32 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 TN-32 
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 TN-32 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 April 19, 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 four comment letters on the proposed rule. The 
commenters included the applicant, a user's group, and two letters from 
members of the public. Copies of the public comments are available for 
review in the NRC Public Document Room, 2120 L Street, NW (Lower 
Level), Washington, DC 20003-1527.

Comments on the TN-32 Cask System

    The comments and responses have been grouped into nine subject 
areas: general, radiation protection, accident analysis, criticality 
analysis, thermal, materials, design, technical specifications, and 
miscellaneous issues. Several of the commenters provided specific 
comments on the draft CoC, the NRC staff's preliminary SER, and the 
TSs. To the extent possible, all of the comments on a particular 
subject are grouped together. The listing of the TN-32 cask system 
within 10 CFR 72.214, ``List of approved spent fuel storage casks'' has 
not been changed as a result of the public comments. A review of the 
comments and the NRC staff's responses follow:

A. General

    Comment A.1: One commenter stated that the NRC is certifying more 
casks

[[Page 14791]]

generically rather than on a site-specific basis. This is not 
consistent with the Nuclear Waste Policy Act (NWPA) guidance and 
results in more site specific changes or amendments, confuses workers 
in the industry, complicates the approval process, requires significant 
NRC resources to address problems as casks get loaded, and requires 
special NRC inspection teams to address new cask problems. The 
commenter further suggested that a standard design should be developed 
by having DOE, NRC, Congress, and other organizations work together to 
choose the best design proposed by vendors. The commenter asked how 
many designs the NRC would ultimately certify, their compatibility with 
the total transport and disposal system and the time and money that 
will be spent approving so many designs.
    Response: The NRC disagrees with the comment. The NWPA directs the 
NRC to establish one or more technologies and does not include specific 
guidance on the number and types of cask designs that should be 
considered, approved, or used. The NRC does not require that a cask be 
universal or be useable at every reactor site. This comment is beyond 
the scope of this rule that is focused solely on whether to place a 
particular cask design, the TN-32 cask system, on the list of approved 
casks.
    Comment A.2: One commenter stated that having different designs at 
one site confuses workers because of the need for different procedures 
and the need to be aware of all changes made to the CoC, the SAR and 
amendment changes. Further, the commenter stated that multiple designs 
will add the potential for human error and could have an adverse affect 
on public health and safety and that the NRC should evaluate how 
multiple cask systems used at one plant can affect safe operations at 
the plant.
    Response: This comment is beyond the scope of this rule that is 
focused solely on whether to place the TN-32 cask system on the list of 
approved casks.
    Comment A.3: One commenter stated that regulations should be 
written more simply to enhance successful implementation.
    Response: The NRC agrees with the commenter that the regulations 
should be easy to understand; however, the commenter did not offer any 
specifics as to what in the regulation was confusing. The actual rule 
change is the addition of the TN-32 cask system to the listing of 
approved casks. The NRC staff is committed to issuing its regulations 
in plain English including this rule.
    Comment A.4: One commenter stated that the NRC should form a 
committee to consider the nuclear waste ``picture'' based on current 
NRC practices and how it will change in the future.
    Response: This comment is beyond the scope of this rule.
    Comment A.5: One commenter stated that NRC approving a large number 
of casks generically results in more site specific changes and 
amendments being needed, confuses workers in the industry, complicates 
the approval process, requires significant NRC resources to address 
problems as casks get loaded, and requires special NRC inspection teams 
to address new cask problems.
    Response: This comment is beyond the scope of this rule.
    Comment A.6: One commenter stated that allowing TN-32 casks to be 
fabricated by exemption adds risk to the public because they will be 
used with as little change as possible. The commenter further stated 
that no TN-32 casks should have been built until a generic 
certification is issued and the documents are finalized and accurate.
    Response: The NRC exemption that allows the casks to be fabricated 
before the design being approved included a technical evaluation of the 
impacts of this action. This evaluation reflected that fabrication of 
the casks with no fuel loading does not add any measurable risk to the 
public. Casks are not used (loaded) until they conform to the final NRC 
approval in the form of an issued CoC or site-specific license.
    Comment A.7: One commenter discussed the Wisconsin Public Service 
Commission (WPSC) lack of concern about TN not having to use positive 
means to verify continued efficacy of the neutron absorbing material in 
the casks.
    Response: Issues related to WPSC are beyond the scope of this rule.
    Comment A.8: One commenter asked if a generic TN-32 had ever been 
built and tested, if there are similar designs being used at the Surry 
nuclear plant and if the Surry casks are site-specific designs, if 
Wisconsin Electric Power Company (WEPCO) has built similar cask 
designs, if any similar designs have been loaded, what the track record 
has been for similar designs, how long the casks have been used at 
other sites, whether closure seals have been replaced and where, 
whether exemptions were required elsewhere, and whether these other 
casks will need changes to meet the current proposed design.
    Response: As noted in the SAR in Chapter 1, the standard TN-32 cask 
was approved by the NRC as a Topical Report in 1996 for reference in 
site-specific applications. Currently there are nine TN-32 casks 
located at the Surry site that were first loaded in 1996 and five at 
the North Anna site first loaded in 1998. A successful dry run was 
performed before the first loading at each site. WEPCO has the VSC-24 
design casks at its Point Beach site that is a different design than 
the TN-32. O-rings have been replaced on casks on the Surry site. 
Exemptions have been granted to other cask designs and are publically 
available. There will be no requirements to change already approved 
cask designs to meet the specifications of the design being approved by 
this rule, because there has been no NRC finding as part of the current 
review that calls into question any NRC safety findings on previous TN-
32 designs.
    Comment A.9: One commenter stated that the environmental assessment 
(EA) using the tiered approach on past environmental analysis is not 
valid and an environmental impact analysis should be performed for the 
TN-32 and every other new cask design.
    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-32 in a generic setting. The NRC has given specific consideration of 
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 believes that meaningful new environmental insights would not 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 rulemaking, fully comply with 
NRC's environmental regulations in 10 CFR Part 51. The Commission's 
environmental regulations in Part 51 implement the National 
Environmental Policy Act (NEPA) and give proper consideration to the 
guidelines of the Council of Environmental Quality (CEQ). The EA and 
FONSI prepared for the TN-32, as required by 10 CFR Part 51, conform to 
NEPA procedural requirements. Tiering on past environmental Impact 
Statements (EISs) and EAs is a standard process under NEPA. As stated 
in CEQ's ``Forty 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.

[[Page 14792]]

    Comment A.10: One commenter provided a number of comments and 
questions on the use of TN-32 casks by WEPCO. The commenter asked about 
why casks may be made in Japan and who would regulate this process. 
This commenter also expressed concern about the increased costs for the 
TN-32 over that of the VSC-24.
    Response: These comments are beyond the scope of this rule that is 
focused solely on whether to place the TN-32 cask system on the list of 
approved casks. Decisions made by specific utilities on why a specific 
cask is chosen over another design are beyond the scope of this rule. 
If WEPCO chooses to use the TN-32 cask design at the Point Beach site, 
the licensee will be required to perform an evaluation in accordance 
with 10 CFR 72.212 to determine whether activities related to storage 
of spent fuel under the general license would involve any unreviewed 
safety question as provided under 10 CFR 50.59. In accordance with this 
regulation, the licensee would make changes to existing lifting systems 
and any physical changes to the facility as necessary to accommodate 
new cask designs. Each of these changes would need to be evaluated per 
10 CFR 50.59 to determine their impact on other systems and on existing 
safety analyses. The NRC does not have a role in selecting particular 
manufacturers for a cask. Each CoC holder and licensee is responsible 
for ensuring that the quality assurance requirements for a cask are met 
by the fabricator. The cost of cask fabrication is beyond the scope of 
this rule.
    Comment A.11: One commenter stated an opinion that burnable poison 
rod assemblies (BPRAs) and thimble plug assemblies (TPAs) should not be 
placed in casks but should be shipped in low level waste containers to 
low level waste storage facilities. This would make the shipping 
process less costly and would result in simpler procedures and 
analyses.
    Response: The NRC disagrees with this comment. The inclusion of 
BPRAs and TPAs in the spent fuel casks provides better protection by 
limiting potential radiation exposure for the plant workers and the 
public than handling these items separately. Even though the radiation 
source term in the casks due to BPRAs and TPAs is higher, the user at 
each site must take steps and measurements to ensure that the 
regulatory limits on dose rates are met. The cask users will have 
procedures to address the differences in handling casks with and 
without BPRAs and TPAs. Storage of spent fuel assemblies and their 
associated hardware that includes BPRAs and TPAs in a cask is not 
prohibited by NRC regulations. The comment about storage of BPRAs and 
TPAs at a low level waste facility is beyond the scope of this rule.
    Comment A12: One commenter asked who verifies that fabricators are 
qualified to build casks and suggested that the NRC set up evaluation 
criteria and enforcement programs to bar unqualified companies. The 
commenter also voiced a concern that vendors and subcontractors are new 
to the nuclear industry and require strong and effective quality 
assurance.
    Response: The CoC holder and licensee are responsible for verifying 
that fabricators are qualified. The choice of who fabricates a 
container is a business decision made by the licensee or certificate 
holder seeking to build containers. The CoC holder and licensee must 
have an NRC-approved Quality Assurance (QA) Program that is approved as 
part of the licensing or CoC issue process. This QA program must meet 
the requirements of 10 CFR 72.148 and 10 CFR 72.154 for the selection 
of fabricators. Also, the procurement documents issued to the 
fabricator must comply with 10 CFR 21.31. These requirements are passed 
onto fabricators as part of a contract or through other procurement 
documents. The licensee/CoC holder is required to verify that all 
regulations applicable to the container are met. The NRC inspects the 
licensee/CoC holders and fabricators to verify compliance as well. The 
NRC has a defined process for taking enforcement actions against those 
that do not comply with NRC regulations.
    Comment A.13: One commenter recommended that the NRC certify only 
dual purpose casks in the future.
    Response: This comment is beyond the scope of this rule. The 
current regulatory framework does not preclude an applicant from 
requesting certification of either a transport, storage, or dual 
purpose cask. The NRC may approve any one of these designs.
    Comment A.14: One commenter disagreed with the NRC position that 
the independent spent fuel storage installation (ISFSI) must be 
designed to withstand the same safe shut down earthquake as for the 
adjacent nuclear power plant. Instead, this commenter recommended that 
an ISFSI pad should be required to have its own specific seismic 
analysis because the reactor and the ISFSI may be located on different 
types of soil or land forms.
    Response: The NRC disagrees with the recommendation that each ISFSI 
pad be required to have a specific seismic analysis. Before using the 
TN-32 cask, 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.
    Comment A.15: One commenter addressed the references included in 
the NRC SER. This commenter suggested that all references should be 
dated and that more current versions of references should be listed.
    Response: The NRC agrees with this comment. Reference dates have 
been added and more current versions of references have been added to 
the SER where appropriate.
    Comment A.16: One commenter stated that the utility should not 
decide the amount of dose to the public that will be generated by the 
casks and that there should be a public hearing for each design that is 
proposed for use by a utility. The commenter further stated that the 
public knows nothing about how utilities choose cask designs at most 
locations, does not read the Federal Register, and feels incapable of 
reading NRC documents. The commenter added that the public should be 
given a choice as to what they want to be placed on a pad in the 
vicinity of their homes.
    Response: This comment is beyond the scope of this rule. The NRC is 
not involved in the decision process used by utilities to select a cask 
design. A utility may choose any certified cask design for spent fuel 
storage. However, the potential dose to the public from the cask use 
may not exceed NRC regulatory dose limits. The rulemaking process used 
by the NRC for generic approval of casks is the regulatory vehicle used 
to obtain public input and ensure protection of public health and 
safety and the environment. This final rule adds the TN-32 cask design 
to the list of approved casks available for use by a power plant 
licensee under the conditions of the general license in 10 CFR Part 72. 
Those conditions require each licensee to determine if the reactor site 
parameters are encompassed by the cask design bases considered in the 
cask SAR and SER.
    Comment A.17: One commenter stated that the computer based safety 
analysis that is discussed in SER Chapter 6 is not a realistic way of 
dealing with the design and accidents and requested that actual 
conditions be evaluated.
    Response: The NRC disagrees with this comment. As stated in SER 
Section 6.3.1, the most limiting conditions are combined and bound all 
credible conditions. The NRC staff accepts analytic conclusions based 
on sound engineering methods and practices. NRC accepts the use of 
computer modeling

[[Page 14793]]

codes to analyze cask performance. The NRC found the computer codes and 
models used by TN to be appropriate as discussed in the SER.
    Comment A.18: One commenter asked who would be responsible for 
conducting a heat load test following a cask design change. The 
commenter suggested that a cask user would probably evaluate the design 
change under 10 CFR 72.48 rather than conducting a heat load test. The 
commenter also stated that the use of 10 CFR 72.48 results in ``goofing 
up'' design documents.
    Response: The comment on the 10 CFR 72.48 process affecting design 
document quality is beyond the scope of this rule. As required by the 
TN-32 Certificate of Compliance (CoC), prior to loading a cask with a 
heat load equal to or greater than 23.7 kilowatts, the heat transfer 
performance of a cask shall be verified by a thermal test. The CoC also 
requires that any changes to the fabrication process be evaluated for 
thermal impact. If the change is found to be significant, the heat 
transfer performance of a modified cask shall be verified by an 
additional thermal test prior to loading a modified cask with a heat 
load equal or greater than 23.7 kilowatts. If the heat load exceeds the 
CoC specified value, there is no option to use 10 CFR 72.48 to avoid 
performing a repeat test.
    Comment A.19: One commenter asked how the NRC will ensure that TN 
will independently verify the adequacy of the cask design and that 
changes to design documents will be reviewed and approved by the same 
organizations who performed the original design.
    Response: Independent design verification reviews and reviewing 
design changes are governed by an NRC-approved QA program. The NRC 
performs inspections to verify that a CoC holder meets its approved QA 
program requirements. 10 CFR 72.232, ``Inspection and Tests'' provides 
the NRC permission to perform inspections and tests at any time. The 
NRC will be able to determine the adequacy of the independence of TN's 
cask design verification through the inspection program.
    Comment A.20: One commenter stated that design records should be 
legible because their use is important during emergency situations, 
that the requirement for independent inspections should be emphasized 
in documentation and enforced, that calibration records are of grave 
importance and need constant verification that each action was 
completed, and that verifications should be done in a timely manner.
    Response: The NRC agrees that design records must be legible and 
should be complete and accurate. Several regulations address the 
quality of records that are maintained by applicants and licensees. 
Each CoC holder must have an NRC-approved QA program. An NRC-approved 
program includes specific requirements for quality assurance records, 
independent inspection and testing, and control of measuring and test 
equipment. Ultimately, according to their approved QA program, the 
licensee/CoC holder must maintain necessary and sufficient records as 
evidence of activities affecting quality under routine and emergency 
conditions.
    Comment A.21: One commenter asked what the standard industry 
decommissioning practices are for decommissioning (referred to on Page 
14-1 of the SER), asked if implementation of decommissioning will be a 
big problem for the TN-32 design, how wet transfer will be dealt with 
for casks in the future, and when a dry transfer method will be used 
along with dual purpose casks. Also, the commenter asked if there is a 
proposal for a dry storage method with an associated dry transfer 
process and what the results were from the Transnucleaire of France 
report to DOE and EPRI on dry transfer.
    Response: The phrase ``standard industry decommissioning 
practices'' in the SER refers to general practices of decontamination, 
cask disassembly with adequate radiological and occupational safety 
controls, fuel handling procedures, and safe component transportation 
and disposal. Decommissioning implementation will be addressed as a 
site-specific issue. The remaining portions of this comment are beyond 
the scope of this rule.
    Comment A.22: One commenter stated that the review should consider 
the ultimate disposal of the spent fuel.
    Response: This comment is beyond the scope of this rule. The CoC 
for the TN-32 is intended for the interim storage of spent fuel. Use of 
the TN-32 cask design for disposal at a high-level waste repository is 
beyond the scope of this rule. DOE has not yet made final decisions 
regarding cask design or deployment for the cask design to be used in 
the high-level waste repository.
    Comment A.23: One commenter asked if the TN-24 design had ever been 
used and why it is not in production currently. Also, the commenter 
asked why casks holding 24 and 32 assemblies are being approved and 
used while the Yucca Mountain facility description discusses casks with 
a 21 assembly capacity.
    Response: The comments on the TN-24 design are beyond the scope of 
this rule. A final decision on the design of storage casks for disposal 
at Yucca Mountain has not been made.
    Comment A.24: One commenter stated that unloading procedures should 
be placed in the NRC public document room.
    Response: The NRC disagrees with the comment. Detailed loading and 
unloading procedures are developed and evaluated on a site-specific 
basis by the licensee using the cask. There is no requirement to have 
detailed procedures placed in the public document room.
    Comment A.25: One commenter stated that the NRC should always 
remember that the priority is public and worker safety, not keeping the 
plants operating; and that the NRC should do the certifications of new 
cask designs very carefully and not as fast as the utility schedule 
demands.
    Response: The NRC's highest priority is to protect health and 
safety of the public including those working at a nuclear plant. Each 
cask certification requires a thorough and careful review of the design 
details and how each design complies with existing regulations. The NRC 
is aware of utility schedules but the NRC completion of certifications 
is based on available resources, the adequacy and completeness of 
applicant submittals, and the complexity of identified technical 
issues.

B. Radiation Protection

    Comment B.1: One commenter stated that the assumptions of ranges of 
cobalt impurities included in the SAR are too great and that more 
current and accurate information should be used. The commenter also 
asked why the value for grid spacers is 4700 ppm and if anyone really 
knows what an accurate measurement is for all of the cobalt in the 
cask. The commenter stated that if the NRC confirmatory calculations 
resulted in 15% lower values for cobalt source terms, then there was a 
mistake.
    Response: The NRC disagrees with the comment. The measurement data 
cited in the SER on cobalt impurity levels in fuel assembly hardware 
was collected before the effects of cobalt impurity were fully 
appreciated. More recently, cobalt impurity levels have been controlled 
during the fabrication process and typically do not exceed 1200 ppm. 
The assumed impurity value of 4700 ppm is accepted as a bounding value 
that will cover past, present, and anticipated future fabrication 
practices for Inconel hardware, and is conservative.

[[Page 14794]]

    The difference between the applicant's and NRC staff's calculations 
for the cobalt source term is not a mistake. The methods available for 
estimating the Co-60 source term are not exact and the results depend 
on the assumed reactor operating conditions that change over time and 
vary from plant to plant. Some variation in results is expected. The 
fact that the NRC staff's values were lower for the Co-60 source term 
show that the applicant's calculations for this term were bounding.
    Comment B.2: One commenter noted that in the SER the NRC stated 
that the integration of the neutron source as a function of axial 
position resulted in a 28% larger total neutron source than that given 
in Table 5.2-3 of the SAR . The commenter asked if the applicant's 
calculations were wrong. Further the commenter suggested that a cask 
design should not just meet requirements but should be ``ALARA-not up 
to the limit.''
    Response: The NRC disagrees with the comment. The calculations of 
the neutron source term made by the applicant are correct. The methods 
available for calculating neutron source terms as well as gamma-ray 
source terms are not exact and some variation between code results is 
expected. The neutron dose rate on the surface of the neutron shield is 
only 10 percent to 15 percent of the total dose from the cask. The 
difference in neutron source term was offset by the higher gamma-ray 
source term estimate by the applicant. Overall, the applicant provided 
a bounding shielding analysis.
    Provided the applicant's design meets the regulatory limits for 
off-site dose, the NRC finds this acceptable. Doses to individuals will 
be determined when the cask is used at an actual site. Each general 
license user of the cask will have a radiation protection program that 
seeks to identify operational alternatives to keep the dose to workers 
as low as reasonably achievable (ALARA).
    Comment B.3: One commenter noted that the NRC concludes in the SER 
that because the aluminum tubes containing the neutron shield material 
have a wall thickness of only 1/8 inch and actual measurements have not 
detected streaming, that streaming through the aluminum wall is not 
significant. The commenter asked who did the measurements and if the 
streaming evaluation was carefully performed. Also, the commenter asked 
who developed the information in Appendix 5A, whether this was the 
source of the measurements, and if the measurements were accurate.
    Response: Appendix 5A does not address the streaming issue and was 
initially provided to support an analysis in a second appendix that was 
later deleted. However, the applicant left Appendix 5A in the SAR for 
informational purposes only. In response to a request for additional 
information on the potential for streaming, the applicant cited other 
measured dose rates around the TN-24P (EPRI NP-5128), the TN-40 and TN-
32 casks, that, ``have shown no streaming effects in moving 
circumferentially around the neutron shield.'' This information was 
considered during the NRC staff's review. Measurements by licensees are 
subject to NRC inspection and no further investigation of their 
accuracy was deemed necessary.
    Comment B.4: One commenter asked why the radial neutron shield is 
not the full length of the cask because dose rates can be higher above 
and below the shield and BPRAs and TPAs have higher doses. The 
commenter also asked what the real dose a person inspecting the cask 
can be expected to receive near the trunnion area above and below the 
neutron shield especially to the head and feet (not just the average 
dose to a person working near the side of the cask).
    Response: The radial neutron shield runs the full length of the 
active fuel region of the fuel assemblies that is the location of the 
neutron source term. The peak surface dose rates at the top and bottom 
edges of the neutron shield are very localized and drop off rapidly as 
one moves away from the shield edge. The applicant's estimate of worker 
exposure did account for the higher doses at the edges of the neutron 
shield coupled with the number and duration of tasks necessary in those 
regions. The estimated dose for loading operations around the upper 
corner of the cask is 2.9 person-rem. The user's ALARA program is 
established to identify local hot spots such as the trunnion area and 
take measures to avoid worker proximity to those areas as much as 
possible. The ALARA program will control the actual doses when the 
casks are loaded at the plant.
    Comment B.5: One commenter stated that the accident dose would be 
much less if BPRAs were not loaded in the casks. The commenter asked 
how the BPRAs affect the total dose to the public in a full cask array, 
how close the calculated doses are to regulatory limits, and how the 
doses compare to those of other approved casks.
    Response: The data provided in the SAR show a less than 20 percent 
increase in the normal and accident doses due to the presence of BPRAs. 
For normal conditions, each general licensee who uses the TN-32 cask 
must perform an evaluation to show that the regulatory off-site dose 
limits will be met at the licensee's site. Thus, a direct comparison to 
the regulatory limits will depend on site-specific conditions and 
usage. The analysis of a typical cask array shows that the dose limit 
to a public resident is met at a distance of approximately 450 meters 
from the storage pad. The accident dose at 100 meters from a cask is 
estimated to be approximately 15 percent of the regulatory limit. 
Because the NRC evaluates the cask design versus the regulatory limits, 
comparison of the TN-32 design to other approved cask designs is beyond 
the scope of this rule.
    Comment B.6: One commenter stated that the casks are really site-
specific from a dose perspective because the dose from everything at a 
site needs to be considered including effluents, low level waste, old 
steam generators, etc. The commenter suggested that a berm would be 
needed, especially to minimize the dose to the public. The commenter 
also asked who evaluates this (the licensee or the utility) and if NRC 
checks the dose calculations.
    Response: The NRC agrees that the actual doses are a site-specific 
issue that will be addressed by the cask users ALARA program. Under 10 
CFR 72.212(b)(2), each general licensee who uses the TN-32 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.
    Comment B.7: One commenter asked if the total dose of 4.25 person-
rem per cask is acceptable to the NRC, if other cask designs have much 
lower total doses, and if the total dose may exceed this value in the 
future. The commenter suggested that an acceptable dose is one that is 
closest to the minimum.
    Response: Although acceptable, the operational dose estimates in 
the application are considered to be bounding values (conservative 
overestimates) and actual doses are expected to be lower. Occupational 
dose limits are set in 10 CFR Part 20. The total dose received during 
cask loading will be shared by a number of workers and is monitored by 
the user's radiation protection program. That program must ensure that 
occupational doses do not exceed regulatory limits. One component of an 
approved radiation protection program is an ALARA program and is 
subject to NRC inspection. Because the NRC evaluates the cask design 
versus the regulatory limits, comparison of the TN-32 design to other 
approved cask designs is beyond the scope of this rule.

[[Page 14795]]

    Comment B.8: One commenter asked if Regulatory Guide 8.8, 
``Information Relevant to Ensuring that Occupational Radiation 
Exposures at Nuclear Power Stations will be As Low as Reasonably 
Achievable'' applies to doses to the public.
    Response: The Regulatory Guide does not directly address dose to 
the general public. It specifically addresses occupational doses to 
reactor station personnel.
    Comment B.9: One commenter asked for the dose rate under the bottom 
plates and how radioactive the pad would become by the time the pad is 
decommissioned.
    Response: The applicant estimated a dose rate of 498 mrem/hour on 
the bottom surface of the cask for a full load of design basis fuel 
assemblies. The amount of activation in the pad is expected to be small 
and will depend on the actual fuel loaded and time of storage. At the 
time of final decommissioning, the cask user will be required to 
measure any induced radiation in the pad and activated material will be 
handled according to the regulatory requirements.
    Comment B.10: One commenter asked why the shielding analysis is 
based on nominal uranium content that is slightly less than the 
specified values.
    Response: For a given fuel design, there will be slight variations 
in the uranium content due to occasional minor modifications made to 
meet the special needs of the buyer for a particular batch of fuel. The 
range of variations is much less than the accuracy of the methods 
currently available for the analysis and will not change the finding of 
reasonable assurance for approval of the design. The maximum limits on 
uranium content specified in TS 2.1.c are set to bound all potential 
variations for the particular design. The values used in the analysis 
are more representative of the fuel most likely to be stored in the 
cask.
    Comment B.11: One commenter asked how hard it is to decontaminate 
the outside of the TN-32 after being in the pool. The commenter further 
inquired as to the extra dose received by the worker in decontaminating 
the cask.
    Response: Decontamination is not a particularly difficult task but 
does take some time and care. Steps are performed to aid the process of 
decontamination as the cask is placed in the pool. Tests are performed 
to determine that effective decontamination is achieved and additional 
decontamination will be performed when needed. Decontamination is 
estimated in the SAR to take 1.5 hours with a maximum worker dose of 
0.27 person-rem.
    Comment B12: This commenter asked if the expected dose rates for 
the TN-32 would be three times that of the VSC-24.
    Response: The projected annual dose from one TN-32 loaded cask is 
described in Table 10.2-1 in the SER and shows what the dose would be 
at different distances. This dose for this design is within regulatory 
limits. The NRC does not conduct its dose review on a comparative basis 
considering other cask designs. The expected doses from other approved 
designs are reflected in SARs from those designs and are publically 
available.

C. Accident Analysis

    Comment C.1: One commenter stated that the 15 minute transporter 
fuel fire should not be the bounding fire accident and recommended that 
the NRC evaluate a large airplane crashing into a full array of casks, 
a lightening strike induced fire, a fuel fire fed by aircraft fuel, or 
a missile that causes a fire at the pad breaking up casks, and burning 
the plastic, seals, and resins. The commenter also asked what the total 
amount of material that could be off-gassed, melted, and burned up if 
several casks were hit by an airplane; what an emergency crew would be 
expected to do given a catastrophic crash into a cask array; what a 
fire crew should spray or dump on a fire to mitigate its severity; how 
one would move and unload a cask with a destroyed neutron shield and 
burned out seals; and whether local emergency crews have action plans 
for such severe fires at a storage pad.
    Response: The NRC disagrees with this comment. The basis for the 
15-minute fire is associated with the time it would take to burn 
approximately 200 gallons of fuel, presumably carried by the 
transporter. 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-32. 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 contents due to the 
thermal inertia of the cask. The weather cover o-ring and neutron 
shield may burn or char if exposed to the design basis fire. Complete 
combustion of the weather cover o-ring would contribute an 
insignificant amount of heat to the TN-32 and would not affect any 
components that are important to safety. The radial neutron shield is a 
polyester material which includes about 50 percent fire retardant fill, 
which makes it self-extinguishing when exposed to fire. The top neutron 
shield is polypropylene which is slow burning or may not burn at all in 
a fire environment. The applicant has added information to the SAR to 
address the combustibility of the neutron shield.
    Other external sources of heat associated with the TN-32 are solar 
insolation and ambient temperatures. These sources are included in the 
thermal analysis in section 4 of the TN-32 SAR. External pressure 
sources include normal atmospheric conditions, flood submersion, and 
explosions. These sources are included in the safety analysis in 
Sections 2, 3, and 11 of the TN-32 SAR.
    The applicant's evaluation of a lightning strike is provided in 
section 2.2.5.2.8 of the TN-32 SAR. No significant thermal effect was 
identified since the electricity would be conducted through the metal 
components to ground. Other vehicles causing the fire (such as 
airplanes, trains, delivery trucks or missiles) are not plausible and 
are beyond the scope of this rule. However, the applicant did evaluate 
the design capability to withstand an explosion with a force up to 25 
psi of external pressure. (See also discussion for Comment C.7.)
    Before using the TN-32 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 
no credible source of an external explosion that would produce an 
external pressure above 25 psi and that any cask handling equipment 
used to move the TN-32 cask to the pad is limited to 200 gallons of 
fuel (refer to Technical Specification 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 in accordance with 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 C.2: One commenter questioned what tornados, lightning, 
fire, and puncture damage would do to the effectiveness of the neutron 
shield. The commenter also questioned whether the plastic seals burn 
easily.
    Response: The top neutron shield and the radial neutron shield have 
not been designed to withstand all of the hypothetical accident loads. 
There may be local damage due to accidents such as tornado missiles, 
fire, etc. Therefore, cask structural analyses have been performed 
assuming that the neutron

[[Page 14796]]

shield is completely removed during the 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 lid 
seals are metal. The design has been found capable of maintaining the 
confinement of radioactive material under the identified credible 
accident conditions even with the loss of the neutron shield. Thus, any 
dose to the public is controlled and would be within regulatory limits.
    Comment C.3: One commenter stated that casks should not be 
permitted to slide at all or much less than the 7.88 in. discussed in 
the SER. Further, the commenter suggested that the analysis should 
assume that the casks could slide in more than one direction. The 
commenter also asked if sliding affects other casks already certified.
    Response: The NRC disagrees. The TN-32 cask will not tipover or 
slide due to tornado and wind loading as analyzed in Section 2 of the 
SAR. The SAR indicates that the cask may slide 7.88 in. 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. 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, 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 reality, upon impact the majority of the energy will be 
absorbed by the crushing of the automobile rather than moving of the 
cask. The NRC has not identified any design issues in the TN-32 review 
which affect any other casks previously approved.
    Comment C.4: One commenter asked that during a tornado, what 
structures are near the casks that could hit one of them and whether a 
meteorological analysis had been done to evaluate the effects of 
tornados on the casks.
    Response: This is a site-specific issue. The cask user will have to 
address this issue in its 10 CFR 72.212 evaluation.
    Comment C.5: One commenter asked about how a cask could become 
buried and what assumptions were used for causes for the burial 
accident.
    Response: TN-32 SAR Section 11.2.10 provides possible causes for 
accidental cask burial such as an earthquake.
    Comment C.6: One commenter stated that the unloading function is 
not given much attention in the full safety analysis of the cask for 
accidents.
    Response: General procedure descriptions for these operations are 
summarized in Section 8.2 of the SAR. These procedure descriptions were 
reviewed by the NRC. As discussed in Section 8 of the SER, the NRC 
concluded that these procedure descriptions were acceptable for use in 
developing detailed site-specific procedures. Detailed loading and 
unloading procedures will be developed on a site-specific basis by the 
cask user.
    Comment C.7: One commenter asked a number of questions relating to 
the accident analysis assumptions for explosions involving combustible 
materials shipped to reactor sites and on transportation links near 
nuclear power plants. Specifically, the commenter asked about controls 
over what is shipped near the Point Beach plant and a number of other 
potential sources of explosion.
    Response: This comment about Point Beach is beyond the scope of 
this rule. The applicant, Transnuclear, did evaluate the TN-32 cask 
design for its capability to withstand an explosion with a force up to 
25 psi of external pressure. Further, the NRC has evaluated the effects 
of a truck bomb located adjacent to storage casks. The use of a 
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. This would include any potential active or passive source of 
explosion at or near the pad.
    Comment C.8: One commenter stated that consideration of a sabotage 
threat is not up to date for ISFSI designs.
    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 (55 FR 29181; July 18, 1990). The NRC still finds 
the results of the 1990 rule current and acceptable. In addition, each 
Part 72 licensee is required by 10 CFR 73.51 or t 73.55 to develop a 
physical protection plan for the ISFSI. The licensee is also required 
to install systems that provide high assurance against unauthorized 
activities that could constitute an unreasonable risk to public health 
and safety.

D. Criticality

    Comment D.1: One commenter stated that in the KENO input file of 
page 6.6-7, the last zero in the unit cell resonance correction input 
should be changed to a 3.
    Response: The NRC agrees with the typographical correction 
suggested by the commenter. The correct unit cell data was used in the 
NRC staff's confirmatory calculations and demonstrated that this error 
had a negligible effect on the criticality safety analysis results. The 
SAR has been revised as appropriate.
    Comment D.2: One commenter asked a question about what confirming 
demonstration and analysis the NRC used to show that ``significant'' 
degradation of the neutron absorbing material used in each cask can not 
occur over the life of the facility. The commenter also disagreed with 
the NRC statement in the SER that neutron absorber plates would have a 
continued efficacy over the 20 year cask life because there is no 
knowledge basis for this and fabricators do not meet perfection in 
their products.
    Response: The NRC staff does not consider the loss or degradation 
of fixed neutron poisons credible after installation into the cask 
because the poisons are fixed in place and contained. The neutron 
absorber is designed to remain effective in the TN-32 system for a 
storage period greater than 20 years and there are no credible means to 
lose the neutron absorber. Section 6.3.2 of the TN-32 SAR describes the 
neutron absorber and its environment, and evaluated boron depletion due 
to neutron absorption. Section 9.1.7 of the SAR describes the testing 
procedures for the neutron absorber material that 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.
    Comment D.3: One commenter asked what the comprehensive fabrication 
test was that is capable of verifying the presence and uniformity of 
the neutron absorber and if any of these tests really exist.
    Response: As stated in SER Sections 6.1 and 6.3.2, the fabrication 
requirements and neutron and visual acceptance tests that must be 
performed are described in SAR Section 9.1.7. In SER Section 9.1.5, the 
NRC staff found the tests are adequate to validate the specified boron 
content and fabrication quality.

[[Page 14797]]

    Comment D.4: One commenter asked why the applicant did not perform 
a calculation to verify that criticality safety is maintained for each 
type of fuel with TPAs that will be stored in the cask versus relying 
on a bounding analysis for fuel containing BPRAs.
    Response: In SAR Section 6.4.2, the applicant explicitly evaluated 
all of the proposed fuel types to determine the most reactive fuel 
configuration. The most reactive fuel type was then used in the 
remainder of the criticality safety evaluation. The SAR shows that 
displacement of highly borated water within the active fuel region 
causes a slight increase in reactivity for this cask under the 
conditions evaluated. The BPRAs bound the TPAs. A fuel assembly can 
only contain either a BPRA or a TPA. The BPRAs extend down into the 
active fuel region and, as stated in SAR Section 2.3.4.1, they displace 
more borated water than the TPAs.
    Comment D.5: One commenter asked about the fuel pin pitch parameter 
role in the calculation of keff, if the NRC understands what 
happens as it varies, and if the NRC expects different effects on 
keff than the applicant does. The commenter also asked if 
the fuel pins ``straighten up'' and become ``more centered'' as water 
comes in around them, and stated that there are a lot of unknowns about 
fuel behavior in dry casks.
    Response: The pin pitch is the distance between fuel pins and can 
decrease if the fuel assembly grid spacers fail as evaluated in SAR 
Section 6A. The NRC staff compared the effects of varying the amount of 
borated water between an array of fuel pins and varying the amount of 
borated water between fuel assemblies in a TN-32 cask. As pin pitch is 
reduced for assemblies in a TN-32 cask, the amount of borated water 
between assemblies increases, resulting in a decrease in reactivity.
    Comment D.6: One commenter asked what operating experience in cask 
unloading is used to establish the frequency for checking the boron 
concentration.
    Response: The frequency for checking the changes to the water boron 
concentration is based on spent fuel pool operating experience that 
does not require experience in cask unloading. There is significant 
spent fuel pool operating experience that supports the TS frequency for 
checking the boron concentration of the water.

E. Thermal

    Comment E.1: One commenter asked why the maximum fuel cladding 
temperature had been reduced from 348  deg.C [as approved in another 
cask design] to 328  deg.C [for the TN-32 design].
    Response: The fuel cladding temperature is established to protect 
the cladding from failure during the storage lifetime. This temperature 
limit is based on several factors including the cladding hoop stress 
and the length of time the fuel has been cooled. Cladding hoop stress 
is related to the rod internal pressure. The rods are pressurized by 
gas present in the plenum and gap. Because casks certified under 10 CFR 
72 Subpart L have a broad range of applicability and in response to the 
NRC staff comments, the applicant selected an upper-bound rod internal 
pressure to develop the clad temperature limits. The resulting maximum 
cladding temperature limit was 328  deg.C. The temperature limit is 
based on the methods given in PNL-6189, Levy, I.S., et al., Pacific 
Northwest Laboratories, ``Recommended Temperature Limits for Dry 
Storage of Spent Light-Water Zircaloy Clad Fuel Rods in Inert Gas'' May 
1987. The limit was found acceptable by the NRC staff in the TN-32 
Preliminary SER. The methods in PNL-6189 are also referenced in NUREG-
1536, ``Standard Review Plan for Dry Cask Storage Systems'' January 
1997.
    Comment E.2: One commenter asked why SAR Revision 11A on page 4-1 
was referenced for this design while Revision 9A was used in a previous 
SAR for the TN-32, and asked if there was a problem with the previous 
analysis and what substantial thermal changes had been included in the 
new analysis.
    Response: The TN-32 SAR Revision 11A, was the version of the SAR 
reviewed and approved by the NRC staff as part of the process of cask 
certification under 10 CFR 72, Subpart L. SAR Revision 9A was the 
version reviewed by the NRC staff for site-specific licensing for casks 
used at Surry and North Anna. The information included in SAR Revision 
9A was not applicable to the TN-32 thermal design and was not reviewed 
by the NRC staff for this current design approval. Therefore, the 
differences in thermal design between the two designs are beyond the 
scope of this rule. The NRC staff did not identify any safety issues in 
Revision 11A that applied to any other cask designs.
    Comment E.3: One commenter asked that the NRC define clearly what 
is meant by ``short term'' for the temperature limit of 1058  deg.F on 
page 4-1 of the SER.
    Response: The short term temperature limit is applicable to 
temporary spikes in cladding temperature such as those that may occur 
in some accidents or during operations like vacuum drying. The NRC 
agrees that this term is unclear and has adopted the concept of a 
transient temperature limit. Guidance on the application of the 
transient temperature limit (referred to as ``short term'') is 
discussed in NUREG-1536, Section 4,V.1. The basis of the 1058  deg.F 
temperature limit for zirconium alloy clad fuel, given in NUREG-1536, 
is from A.B Johnson and E.R. Gilbert, Pacific Northwest Laboratories, 
``Technical Basis for Storage of Zircaloy-Clad Spent Fuel in Inert 
Gases'' PNL-4835, September 1983. Experimental data in that report 
demonstrated no damage to zirconium alloy cladding when subjected to 
1058 deg. F for 30 days. The basis for the temperature limit is to 
avoid conditions that could cause a rod to burst due to excessive 
internal pressure and to limit the amount of creep that may occur at 
the elevated temperatures.
    For spent fuel storage, the NRC staff generally expects the length 
of time the cladding would be at elevated temperatures above the long 
term limit to be much less than 30 days. This expectation is consistent 
with technical specification actions to implement temporary cooling of 
the fuel or to establish acceptable conditions for normal storage 
within periods that are much less than 30 days. These actions typically 
limit the time fuel is allowed to approach and remain at the transient 
temperature limit. In addition, if suitable long term storage 
conditions do not exist or cannot be established, the technical 
specifications may also require further actions such as removal of the 
fuel from the cask within 30 days. This expectation is also consistent 
with the assumptions for accident durations of 30 days or less.
    Comment E.4: One commenter asked a number of questions concerning 
the cask heat up model discussed on pages 4-4 through 4-8 of the SER. 
Specific comments addressed: whether BPRAs and thimble plugs were 
included in the model for weight inputs and for radiation hot spots; 
why the model assumes gaps between the basket and cask bottom, between 
the basket and rails, and between the rails and cavity wall; and 
whether the gaps really exist or are added for conservatism.
    Response: The heat contribution from the BPRAs and thimble plugs 
was considered in the cask analysis. Rather than explicitly modeling 
the fuel assemblies, BPRAs, and thimble plugs, they were modeled as 
homogenized units that had equivalent heat transfer characteristics. 
The weights of various

[[Page 14798]]

fuel assemblies including the heaviest BPRAs for storage in the TN-32 
cask are presented in Table 2.1-1 of the SAR. In SAR Table 5.1-2, the 
applicant provided the incremental dose rate resulting from the BPRAs 
and TPAs at the same locations around the cask as for the fuel 
assemblies including the hot spots above and below the neutron shield. 
Gaps are assumed in the modeling as discussed in the response to 
Comment E.5.
    Comment E.5: One commenter stated opposition to the thermal 
performance of the cask design being based on the gap size in the cask 
body layers. The commenter stated that fabricators will not be able to 
control the gap size to 0.04 inch that errors will occur, and that the 
limit is not conservative enough, adding risk to public safety. 
Further, the commenter noted that requiring only one demonstration test 
of conformance to the gap limit by the applicant will not guarantee 
that the other casks used will meet the same limit.
    Response: Gaps between the various cask components were assumed in 
the analysis to account for fabrication and assembly tolerances and 
uncertainties. The NRC staff expects that the as-built casks will have 
gaps that are less than or equal to those assumed in the analysis. The 
implemented QA program at the fabricator's facility provides reasonable 
assurance that this will occur. However, to demonstrate the adequacy of 
the fabrication process and to provide defense-in-depth, the NRC will 
require thermal testing of a single cask by each agent or subcontractor 
authorized by the certificate holder to complete final assembly of the 
TN-32 cask body. This test shall be performed before the first loading 
of any cask assembled by that agent and/or subcontractor with a heat 
load equal to or greater than 23.7 kilowatts. The test will evaluate 
thermal performance for a range of heat loads up to and including the 
maximum authorized heat load of 32.7 kilowatts. Further, any changes to 
the fabrication process are required to be evaluated for thermal 
impact. If the change is found to be significant, the heat transfer 
performance of the modified cask must be verified by an additional 
thermal test.
    Comment E.6: One commenter suggested that the vendor or applicant 
should conduct tests of the unloading process and full scale testing of 
the cask with a complete load and a representative fuel basket before 
the design is certified by rulemaking. Further, the commenter suggested 
that the test results should be presented to a public service 
commission hearing before a utility decides which cask to purchase and 
use, that the NRC should specify criteria on how the approval process 
is to be conducted and what specifications should be included in the 
SAR and other design documents, and that the NRC should specify that 
vendors and applicants will be fined or contracts will be terminated if 
fabricators do not meet the design criteria. Also, the commenter asked 
why the applicant does not know the thermal responses for the design 
and if the thermal test will be conducted with both sets of trunnions 
for thermal results.
    Response: The NRC disagrees with the comment. The TN-32 storage 
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 required when the analytic methods have not been validated 
or assured to be appropriate and/or conservative. In lieu of testing, 
the NRC finds analytic conclusions that are based on sound engineering 
methods and practices to be acceptable. The NRC staff has reviewed the 
analyses performed by the applicant and found them acceptable. The 
authority of a public service commission to approve a design or the use 
of tests is beyond the scope of this rule. The NRC has issued a number 
of guidance documents including NUREG-1536, ``Standard Review Plan for 
Dry Cask Storage Systems'' that provide information about the criteria 
used by the NRC to approve spent fuel storage cask designs. The design 
approval process is outlined in 10 CFR Part 72. It is the vendor's or 
applicant's responsibility who contracts with a fabricator to ensure 
that the casks and components are built in accordance with the approved 
design specifications and criteria, and in accordance with the CoC 
holder's QA program. If the NRC determines through inspection or other 
means that a cask has been fabricated that does not meet design 
criteria, then the NRC will take necessary enforcement action against 
the CoC holder or utility that is using the cask. The SAR provides a 
thermal analysis acceptable to the NRC staff as discussed in the TN-32 
SER, Section 4. The purpose of the thermal test is discussed in the 
response to Comment E.5 above.

F. Materials

    Comment F.1: One commenter stated that the use of a coating on the 
carbon steel in the cask design will cause problems and stated that 
stainless steel should be used in fabrication.
    Response: The NRC disagrees with this comment. The materials used 
in the fabrication of the cask are described in Chapters 1 and 3 of the 
SAR and discussed in Section 3.1.4 of the NRC SER. Materials have been 
found to have properties that are acceptable as they meet the 
requirements for their respective applications in the cask system. The 
coating on the cask interior is flame sprayed aluminum that is a 
tightly adherent and stable coating in the spent fuel storage 
environment. These materials have been found to be suitable for the 
expected loading and storage in wet and dry environments, including 
corrosion and galvanic effects as discussed in Section 3.2.1 of the 
SER. There is no requirement for designers to select materials from a 
given class, e.g. stainless steels.
    Comment F.2: One commenter stated that freeze-thaw causes icicles 
to hang down from the top of cask and have covered outlets on a VSC-24 
cask at Point Beach and at Fort Saint Vrain. The commenter then asked 
if this can occur on a TN-32 cask and cause dripping along the neutron 
shield; if the resins in the shield can become water saturated; if the 
aluminum sleeves are water tight; if chemical reactions can occur; if 
snow, ice, and water can enter cracks or flaws in the gamma shield and 
reach the containment outer wall; if gaps exist in the trunnion area 
where water can enter; and if corrosion of carbon steel is a concern in 
this design. The commenter also asked if fog, rain, mist, and air 
pollution can affect these casks over time.
    Response: The TN-32 design does not include vents, and therefore, 
there is no concern about ice formation. The outer shell of the neutron 
shield consists of a cylindrical shell section with closure plates at 
each end. The closure plates are welded to the surface of the gamma 
shield. The resins are encased (on all sides) with aluminum or steel. 
Therefore, it is unlikely that water will come in contact with resins. 
However, if water contacted the resin, there is no concern because the 
neutron shielding materials are common plastics that are inert with 
respect to water. The carbon steel is painted to prevent corrosion and 
the integrity of that paint will be monitored by the cask user, and 
repairs will be made if needed.
    Comment F.3: One commenter asked if the quenching effect on BPRAs 
and thimble plugs has been evaluated; if the BPRAs and plugs absorb 
water, expand, and add weight when the cask is reflooded; if the BPRAs 
or plugs fall apart or depressurize, will that affect the removal of 
assemblies from the cask; if pinhole and hairline cracks in the fuel 
rods will absorb water and then later expand as the rods are dried out; 
and if the reflooding water is factored into the

[[Page 14799]]

lifting weight of the cask. Further, the commenter asked if fuel rods 
absorb water, will that prevent removal after long term storage. 
Lastly, the commenter recommended that tests include unloading of a 
real cask at Surry or elsewhere and that an inspection be conducted to 
determine what has happened to the fuel pellets, zircaloy, etc.
    Response: BPRAs rods are constructed in a manner similar to fuel in 
that the neutron absorbing material is placed in sealed tubes made of 
either stainless steel or zirconium alloy. The thimble plug devices are 
solid stainless steel rods. Both BPRAs rods and thimble plug rods are 
attached to a stainless steel baseplate. The NRC staff has not 
identified any conditions in spent fuel dry storage, including 
quenching, that would cause failures of BPRAs or thimble plugs that 
would allow them to absorb water or break apart and affect unloading. 
Further if they are assumed to break apart, the NRC staff has concluded 
there are no adverse safety consequences. Table 1.2-1 in the TN-32 SAR 
provides the cask weight when filled with water.
    Comment F.4: One commenter stated that the applicant should know 
the actual Charpy data rather than providing preliminary data; the flaw 
should not be parallel, radial, or in a line; the flaw depth and width 
should be known; and a special examination of the gamma shield is 
necessary even if the identified flaw size is less than the allowable. 
The commenter also asked how the Charpy V-notch testing will be 
verified before the tested materials are to be used in fabrication, and 
that the NRC clarify just what it is allowing and why.
    Response: The NRC disagrees with the comment. The ``preliminary 
data'' is data based upon other plates and heats made to the same 
specifications as the gamma shield material, SA 266 Grade 2. The 
materials used in actual TN-32 casks will be tested before use in the 
cask to ensure that their properties meet or exceed what is required, 
as indicated in SER Section 9.1.1. The Charpy and other properties 
enumerated in the SER ensure safe performance under service thermal 
conditions. Charpy tests are always conducted using a standard ASTM 
method, E23, ``Standard Test Methods for Notched Bar Impact Testing of 
Metallic Materials.''
    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. Appendix 3E of the SAR 
specifies the allowable flaws for various orientations and locations. 
Flaws of these sizes will not propagate under service conditions. Any 
flaw in the gamma shield will be smaller than these sizes.
    Comment F.5: One commenter noted that the NRC stated in a November 
1, 1996, letter that aluminum oxide flaking might occur in the cask 
during initial heating and cooling, and that the flakes would most 
likely fall to the bottom of the cask and not come in contact with the 
fuel basket. The commenter disagreed with this statement because cask 
transportation and unloading evolutions could cause the flaking to 
contact the basket. The commenter asked what the basis is for NRC's 
position discussed in the 1996 letter and recommended that further 
analysis be performed to determine what happens to the aluminum oxide 
at the end of cask life.
    Response: The comment about the November 1996 letter is beyond the 
scope of this rule. The letter is not related to this cask design. As 
discussed in the response to Comment I.13 the NRC staff expects no 
oxide flaking to occur in the cask.
    Comment F.6: One commenter asked if aluminum flame spray induced 
stains can generate hydrogen or cause other chemical reactions that 
could cause problems, whether there is sufficient time in procedures to 
address this problem, if the NRC understands what the stain is, and if 
it could cloud the pool water and hinder unloading of the cask.
    Response: There is no safety significant effect of the staining due 
to iron or other contaminants in the aluminum oxide. The concentration 
of impurities needed to lead to staining is believed to be so small 
that the NRC staff does not require analysis of chemical reactions that 
might result from the presence of these impurities. There is no 
expected effect on water quality or unloading operations.
    Comment F.7: One commenter asked if basket support rails or the 
basket itself will yield and if an evaluation of the effects of yield, 
temperature changes and drying, and a side or vertical drop or tipover 
for unloading at the end of cask life has been conducted.
    Response: The basket and the rails are evaluated in Appendix 3B of 
the SAR. The aluminum rail will not yield, even under vertical or tip-
over conditions. The internals are always hot. There is no freeze-thaw 
condition. At the end of life, these internal components are expected 
to be in exactly the same condition as they were at the beginning of 
the storage period.

G. Design

    Comment G.1: One commenter stated the assumption that the new top 
lifting trunnions are compatible with the Point Beach transporter and 
will be addressed by existing procedures. The commenter then asked if 
the TN-32 trunnions have been tested with the Point Beach transporter 
criteria, if there may be gaps and streaming at trunnion locations, 
what the dose effect may be, what heavy load criteria exist, and what 
testing will be done.
    Response: The comment on the trunnion testing and compatibility 
with the Point Beach transporter is beyond the scope of this rule. 
Point Beach will have to address the issue in its site-specific 
evaluation under 10 CFR 72.212. Under a cask user's ALARA program to 
minimize worker exposure, localized radiation hot spots such as gaps 
and streaming around the trunnions will be avoided, or have temporary 
additional shielding during cask handling and preparation for transport 
to the storage pad.
    Comment G.2: One commenter asked a number of questions about the 
fuel basket cavity that included: what the weight or total load that is 
transferred from the fuel basket cavity to the lip on the gamma shield 
shell is, and where the load is transferred; how the shrink fit works, 
how it is performed, why it is done, and if it has been tested; could 
water get between the containment shell and the gamma shield shell; 
what the potential is for corrosion between the two shells; whether an 
external event such as an airplane crash, tipover, or seismic event 
could cause the shells to separate; whether tests for freeze-thaw 
temperature changes for the life of the cask have been done; whether 
the two shells contract or expand together; if there is a way that 
pressure or stress can be transferred from one shell to the other and 
cause cracks in the welds or the containment wall; how much stress is 
created in the welds and on the bottom plate; and how the inner shell 
is lifted without removing the inner containment.
    Response: The area referred to by the commenter as the ``lip on the 
gamma shield shell'' is interpreted by the NRC staff to be the 
confinement shell top forging. The fuel basket and fuel assemblies that 
weigh about 66,000 pounds rest directly on the bottom confinement 
plate. Therefore, the fuel basket and fuel assemblies weights are not 
transferred to the confinement shell top forging.
    The shrink fit is established as follows: The gamma shield shell 
and the confinement shell are fabricated

[[Page 14800]]

separately. In order 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 is then 
preheated which causes it to expand before slipping on the confinement 
shell. After the gamma shield shell cools, it shrinks and tightly 
clamps onto the confinement shell. Therefore, the fit between these two 
shells is very tight and no water could migrate between the two shells 
over the life of the cask. Consequently, corrosion between the two 
shells is not a concern. An external event such as a fire, tipover, or 
seismic event would not cause the two shells to separate as 
demonstrated in Sections 3, 4, and 11 of the SAR.
    Likewise, temperature effects on the cask are evaluated in Sections 
3 and 4 of the SAR. Due to the similarity of materials, both shells 
will contract or expand together. The 1.5-inch thick confinement shell 
is supported by the 8-inch thick gamma shield shell. Under accident 
conditions, the gamma shield shell protects the confinement shell from 
damages. The amount of stresses that are created in the welds and on 
the bottom plate due to various service loading combinations are less 
than the ASME allowable values and are presented in Sections 3.4 and 
Appendix 3A of the SAR. The TN-32 cask has a confinement shell that can 
not be removed.
    Comment G.3: One commenter stated that the shape of the cask is of 
concern because the neutron outer shell does not cover the gamma shell 
at the top and bottom. The commenter then asked if this is due to the 
location of the trunnions and suggested that in a drop accident, the 
bottom trunnions might crack off and the edge of the neutron shell 
could be easily crushed or smashed.
    Response: The NRC does not agree with the comment. Radially, except 
at the trunnions, the neutron shield runs the full length of the active 
region of the spent fuel assemblies which is the source of neutron 
radiation. The accident analysis for the TN-32 cask assumes that the 
neutron shield and steel outer shell were removed completely. With this 
assumption, the accident analysis bounds any lesser damage to the 
neutron shield and shell, and the estimated dose is within regulatory 
limits.
    Comment G.4: One commenter asked if the load bearing aluminum rails 
can be jammed during unloading, whether crud or paint particles can 
fall into the rail slots and cause a movement problem, what other 
movement problems exist, or whether there would ever be a reason to 
remove the basket.
    Response: The aluminum rails are located outside the basket. They 
do not interfere with the unloading operation. The aluminum rails 
establish and maintain basket orientation, and enhance heat transfer. 
The rails that surround the basket are oriented parallel to the axis of 
the cask body and are attached to the inner cavity wall of the cask 
body. Consequently, lateral movement of the basket inside the cavity is 
restricted by the rails. Although the basket is not attached to the 
cask body, there is no need to remove the basket from the cask cavity 
during an unloading operation.
    Comment G.5: One commenter stated that the TN-32 is designed not to 
be susceptible to brittle fracture in temperatures as low as 20 deg.F 
and noted that this was a positive characteristic for storage in cold 
climates.
    Response: No response is necessary.
    Comment G.6: One commenter asked a number of questions on fuel rod 
gas including: why the assumption is made that fuel gas internal 
pressure is present when the NRC permits an unlimited number of pinhole 
leaks and hairline cracks that would apparently permit the gas to 
escape over the 20 year life of the cask; what happens to the gas and 
does it mix with the helium; what the gas is; and what chemical 
reactions it can cause inside the cask.
    Response: Based on operational experience, only a very small 
fraction of the fuel rods develop leaks (pin holes, hairline cracks, 
etc.) during reactor operation and pool storage. At the time of dry 
storage, the majority of fuel rods are intact and contain pressurized 
gas. The gas present in the spent fuel rod after removal from the 
reactor is from two sources helium fill gas placed in the rod during 
manufacture and a fraction of the fission gases (mostly krypton, xenon, 
and tritium) produced and released from the fuel pellets during reactor 
operation. Maintenance of intact cladding and retention of the gases 
within the rods, throughout dry storage, is part of the cask design 
consideration to protect operational personnel from unnecessary dose 
during unloading and to provide defense-in-depth. If the unlikely 
release of gas from a rod were to occur, the gases would mix with the 
cask fill gas and remain within the confinement boundary. The bulk of 
these gases are chemically inert and will not react with materials 
inside the cask. The trace amounts of gases that are chemically 
reactive include cesium (a volatile expected to exhibit gas-like 
behavior at cask conditions). There may be some chemical reactions 
between these reactive materials and the zirconium and steel in the 
cask. These reactions would be minimal and would not adversely affect 
the functions of any components that are important to safety.
    Comment G.7: One commenter asked a number of questions about fuel 
pellets including, what the basis is for determining the weight of the 
fuel pellets after reactor exposure and pool exposure; whether pellets 
crack or break up over time, whether the pellets can absorb or adsorb 
water coming in from pinhole leaks and hairline cracks, and how it is 
determined that the pellets are dry when put into storage.
    Response: The fuel weight is based upon data supplied originally 
for new fuel. Increases in fuel weight due to service exposure are 
minimal because they are due to oxidation. The fuel is UO2 
and does not oxidize unless the fuel cladding fails in service. After 
exposure to oxygen or water (in failed fuel) it becomes more rich in 
oxygen. This is represented as U4O9. Because most 
of the weight is in the uranium (mass about 238) and not in the oxygen 
(mass 16), this small increase represents an insignificant change. An 
average weight for the fuel type is taken into account in any 
calculations that require knowledge of mass of this system component.
    The pinholes and hairline cracks would not absorb water, although 
they may be involved in the sorption of moisture and uptake of oxygen 
within the fuel because they could permit pool water, cask moisture, or 
cask oxygen to enter the fuel rod and contact the fuel pellets. 
Pinholes and hairline cracks are not expected to form during dry 
storage because the storage environment for the fuel cladding is 
maintained under protective and durable conditions.
    The behavior of the fuel pellets is well studied and many 
literature references are available on this topic. Cracking in the fuel 
pellets generally occurs during reactor operation . The fuel pellets 
are fairly inert in the absence of oxygen. Therefore, the fuel is dried 
and then stored in a dry, helium gas (water and oxygen free) 
environment to preclude further oxidation.
    In preparation for dry storage, the loading process ensures that 
moisture is removed from the fuel cladding, any fuel that may have 
pinholes or hairline cracks, and from the cask internals. The cask is 
thoroughly vacuum dried as prescribed in the technical specifications 
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

[[Page 14801]]

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. Because the storage 
system provides an inert environment throughout the licensed period, 
very little further oxidation is expected to occur under normal storage 
conditions.
    Comment G.8: One commenter asked how the aluminum boxes filled with 
resin are arranged around the cask, how far apart they are, how they 
are held to the gamma shield wall while the outer shell is installed, 
and to what are they attached.
    Response: As shown in TN-32 SAR drawing 1049-70-2 and described in 
TN-32 SAR Sections 1.2.1, 3.1.1, and 4.1, about 60 aluminum boxes are 
tightly fitted around the exterior of the gamma shield. A steel outer 
shell completely encloses the aluminum boxes and holds them in place 
after construction. Details, such as temporary measures to hold the 
boxes in place during construction will be addressed by fabrication 
procedures and are beyond the scope of this rule.
    Comment G.9: One commenter asked if the fuel basket rails discussed 
on page 4-2 of the SER can come loose over time along with the basket 
and affect the unloading of the fuel, and why they are not welded 
instead of being bolted as designed.
    Response: Neither the applicant nor the NRC staff has identified 
any mechanisms that would cause the basket rail bolts to come loose 
over time. The basket rails are bolted to the cask wall because they 
are aluminum (for heat transfer) and the container wall is steel. The 
only function of the basket rail attachment bolts is to attach the 
basket rail to the inner cavity wall of the cask body; the bolts do not 
support any other loads. A bolted attachment functions as well as a 
welded attachment. Therefore, there is no need to weld the basket rail.
    Comment G.10: One commenter suggested that changes be made to the 
SER concerning the shield lid design for the TN-32. The commenter 
stated that the only drawing in the SER of the shield lid is not very 
clear and asked if it is accurate. Further, the commenter suggested 
that the drawing should add details about TN-32 designs A and B to show 
the differences in lid designs and why they exist. The commenter 
suggested that on page 5-1 of the SER that the NRC should provide a 
better explanation of the lid thickness calculations and that the SER 
should discuss the materials that are being used in the lid design and 
how the changes affect the analysis of the cask.
    Response: The NRC disagrees with this comment. The commenter 
requested changes in the level of detail included in the SER to better 
describe the cask design. The applicant's SAR includes a level of 
design detail that enables the NRC to make a safety finding. However, 
that same level of detail does not need to be repeated in the SER 
because it is already available on the docket and is retrievable by the 
NRC staff and the public. The NRC further disagrees that additional 
information on thickness calculations, a discussion of lid materials, 
and how changes in materials affect cask analysis should be added to 
the SER in Chapter 5. The applicant chooses design materials, 
dimensions, and methods of shielding, and includes details on this and 
supporting analysis in its SAR. The NRC followed its review guidance in 
NUREG-1536, ``Standard Review Plan For Dry Cask Storage Systems'' 
January 1997 and provided the appropriate level of detail and 
information specifically in Chapter 5 to reflect areas of review and 
findings.
    Comment G.11: One commenter noted a concern about the applicant's 
proposed compression of BRPA springs or using a modified lid design. 
The commenter suggested that this was another example of a generic 
design being changed to a site-specific one in effect and, therefore, 
this should have been requested as a site-specific cask design 
application for approval of storage of BPRAs. The commenter then asked 
why the applicant had not designed the casks to hold Westinghouse 14x14 
fuel in the beginning rather than changing the design later to 
accommodate longer assemblies due to the BPRAs collar. The commenter 
also stated that the design changes lead to confusion.
    Response: The NRC disagrees with this comment. A vendor can choose 
to include any design characteristics and must demonstrate that the 
design is safe and in compliance with existing regulations. Adding the 
capability to store BPRAs could also have been requested under a site-
specific license, but the regulations do not suggest one method over 
the other. The question about why the applicant did not originally 
design the cask to hold 14x14 fuel is beyond the scope of this rule.
    Comment G.12: One commenter asked for a description of the 
difference between the configuration of the 6 inch shield plate and the 
4.88 inch shield plate of the lid, why the 4.88 inch plate is 
acceptable, for a description of the 1.25 inch plate incorporated in 
the neutron shield, how the lids are put together, how having different 
lid designs will affect handling procedures, if the top neutron shield 
(4 inch thick polypropylene) is encased by 0.25 inches of steel on the 
top and bottom resulting in a total thickness of 4.5 inches, what being 
encased means, how the neutron shield encasing is welded together, if 
polypropylene is flammable or if it holds water, how it reacts under 
accident conditions of increased pressure and temperature caused by 
fire or explosions, how it could effect the resins on the outside of 
the cask during a fire event, and if it could be repaired after being 
melted. The commenter also suggested that it should be assured that the 
design can accommodate a correct fit of the drain pipe through the lid 
and that vent and drain closures are appropriate for the design.
    Response: ``Encased'' means that the neutron shield is enclosed in 
a steel shell on all sides. The casing seams are welded with full 
penetration or fillet welds depending on the joint configuration. The 
alternate lid design for the TN-32A removes 1.12 inches of steel from 
the under-side of the lid and adds a 1.25 inch plate on the top side of 
the lid. Thus, for the TN-32 and TN-32B, the neutron shield casing is 
0.25 inches thick on the top and bottom giving a thickness of 0.5 
inches in the casing material plus 10.5 inches (6 inches + 4.5 inches) 
in the lid for a total steel thickness of 11 inches at the top of the 
cask. For the TN-32A, the bottom of the casing is the 1.25-inch thick 
supplemental plate and the top of the casing is a 0.38-inch thick steel 
plate giving a thickness of 1.63 inches in the casing material plus 
9.38 inches (4.88 inches + 4.5 inches) in the lid for a total steel 
thickness of 11.01 inches at the top of the cask. Thus, the effective 
thickness of the lid was not changed and is acceptable. The two thick 
steel plates in the lid are welded together and the neutron shield in 
its casing is bolted to the top of the lid.
    The radial neutron shield is a polyester that includes about 50 
weight percent fire-retardant mineral fill, making it self-
extinguishing. The top neutron shield is polypropylene that is ``slow 
burning to nonburning'' according to Table 24, Section 1 of the 
``Handbook of Plastics and Elastomers.'' Furthermore, the weather 
protective cover isolates the top neutron shield

[[Page 14802]]

material from sources of ignition and the radial neutron shield is 
completely encased by the aluminum tubes and by the outer shell.
    Both neutron shielding materials are common commercial plastics 
that are inert with respect to water. Again, the weather cover and the 
outer shell protect the material from direct contact with water.
    Each user of the cask will have operating procedures to address the 
different lid designs if more than one design is used onsite. The two 
different lid designs are configured to accommodate the correct fitting 
of the drain and vent closures and associated hardware.
    Comment G.13: One commenter asked if the requirement of 10 CFR 
72.236(c) for redundant sealing is only for O-rings and about the 
applicability of this requirement for the welds for the VSC-24, and 
whether the shield lid weld is verified by ultrasonic testing.
    Response: The requirement of 10 CFR 72.236(c) for redundant sealing 
is applicable to all casks that are approved under 10 CFR 72 Subpart L. 
The question about VSC-24 is beyond the scope of this rule. The TN-32 
cask does not have a shield lid weld.
    Comment G.14: One commenter asked if hydrogen would build up above 
the water level if the evacuation line became iced up and blocked, how 
the cask remains vented, whether it uses a metal pipe that runs far 
from the cask rather than using a flammable plastic pipe or duct tape, 
whether there are any sources of ignition if hydrogen did escape from 
the venting, what the heat source is that the NRC discusses in the SER 
and how it would affect the vented hydrogen, and whether a clogged line 
could cause water to remain in the cask longer than expected like it 
did at Arkansas Nuclear One.
    Response: A discussion about hydrogen generation and control is 
discussed in the response to Comment I.5. The TN-32 SAR, Table 8.1-1, 
recommends the use of heat tape as the heat source to preclude icing of 
the evacuation line during vacuum drying.
    Clogging of the drain line due to a design or material condition in 
the TN-32 cask is judged by the NRC staff to be an unlikely occurrence. 
However, if a clogged line caused delays in a draining operation, there 
is not an immediate safety concern because the fuel will be adequately 
cooled, or a criticality or shielding concern, and any hydrogen that 
may form will be vented.
    Comment G.15: One commenter suggested that the marking on a dry 
cask should carry more information than model number, identification 
number, and empty weight, and that the marking should be on a plate 
that is covered and will not rust, and should state all important 
information about its contents because paper records can be lost or 
destroyed. This labeling would be useful in an accident, sabotage, or 
war to identify cask contents. The commenter also asked if the NRC has 
carefully reviewed the labeling of casks and the storage of supporting 
documents.
    Response: The NRC agrees in part with this comment. Each cask must 
be conspicuously and durably marked with model number, a unique 
identification number, and with its empty weight under 10 CFR 72.236. 
The NRC did evaluate the need for and types of labeling in the 
statements of consideration for 10 CFR Part 72. The applicant in 
Section 1.2.1 of the SAR states that each cask will be marked with the 
required information but did not address the durability and visibility 
aspect of the marker. The SAR has been modified to reflect this missing 
information.
    However, NRC regulations do not require the identification of cask 
contents on permanent markings affixed to the cask. The NRC notes that 
72.212(b)(8) requires that each general licensee accurately maintain a 
record for each cask that lists the spent fuel stored in the cask. This 
record must be maintained by the cask user until decommissioning of the 
cask is complete.
    Comment G.16: One commenter suggested that this cask design 
requires a berm to minimize doses to the public and that all dry cask 
storage installations should require berms to reduce line of sight for 
potential sabotage, vehicle access, and dose to the public.
    Response: The NRC disagrees with this comment. These are site-
specific issues that will be addressed by the cask user's ALARA program 
and physical protection program.
    Comment G.17: One commenter stated that replacement of O-rings in a 
cask causes unnecessary dose consequences, requires time and resources 
and creates schedule problems for pool use. The commenter asked if 
using O-rings in the design was a good idea because of the need for 
replacement over time, how complicated the replacement process is, and 
if it must be performed with the cask in the pool.
    Response: The materials used for the cask seals are durable and are 
expected to remain functional for the lifetime of the cask. In SAR 
Section 2.3.2.1, the applicant included test results for seals that 
have been in service since 1973. These tested seals are similar to 
those planned for use in the TN-32. Those results demonstrate a very 
good record of seal integrity, performance, and endurance.
    If a seal required replacement, the expected dose for the workers 
performing that task would be less than or equal to the dose expected 
for cask unloading operations. The actions to replace the seal will be 
similar to those required for unloading except that fuel manipulation 
is not required. Seal replacement for the TN-32 would require placing 
the cask in a suitably shielded environment such as the spent fuel 
pool.
    Comment G.18: One commenter had several questions/concerns on the 
design of the TN-32 cask as follows: whether the neutron shield on the 
top overlaps the gamma shield enough to cover the area of streaming up 
the gap on the sides of the lid, why there isn't a bolt on the left of 
the gamma lid, whether the rim with all of the bolt holes has been 
evaluated for stresses and cracking around the bolts and holes, why the 
neutron shields don't go up higher and down lower to cover the entire 
area, and why the trunnions don't fit into the neutron shield rather 
than above and below the shield. The commenter further questioned 
whether the doses would be lower with a different outside neutron 
shield.
    Response: The neutron shield on the cask lid overlaps the outer 
most edge of the fuel by about one inch and is sufficient to prevent 
vertical streaming of the neutrons. The effects of angular streaming 
were considered in the analysis and included in the estimated 
operational dose to the workers and off site. The drawing is simply 
showing different sections of the cask in the same view and is not a 
symmetrical cross section. There are 48 bolts for attaching the lid to 
the cask body. Closure bolts were simulated in the finite element model 
by the coupling of corresponding nodes at the location of the bolt. 
Stresses in the closure bolts and surrounding areas due to various 
serviced loading combinations are less than the ASME allowable values 
as demonstrated in Appendix 3A of the SAR. Consequently, cracking 
around the bolts and holes will not occur. The radial neutron shield 
runs the full length of the active fuel region that is the location of 
the neutron source. The design has been found to be acceptable after a 
review against the regulatory requirements. The neutron shield extends 
half way up the upper trunnion so the trunnion must penetrate through 
the shield to attach to the cask body.

[[Page 14803]]

The placement of the trunnions is influenced by operational and 
handling considerations as well as regulatory factors. As long as the 
cask design meets the regulatory requirements, the details of design 
are the applicant's prerogative.
    Comment G.19: One commenter expressed concern over the issue of ice 
clogging drain lines and asked if some company could develop a vacuum 
draining system that wouldn't have ice clogging concerns.
    Response: The potential for ice formation in the vacuum lines can 
occur from the cooling effect of water vaporization and system 
depressurization that occur during evacuation. Icing is not expected in 
the cask because of the heat generated by the fuel. Reasonable 
precautions such as heating the evacuation lines (using heat tape) or 
controlling the evacuation rates by performing the evacuation in a 
series of stages are adequate to preclude icing problems.
    Comment G.20: One commenter suggested changes to the tolerances in 
SAR drawings 1049-70-1, 1049-70-3, 1049-70-4, and 1049-70-5.
    Response: The NRC agrees with the comment. These changes to the 
tolerances specified on the SAR drawings will not affect the structural 
analyses and the conclusions reached in the SER. The drawings have been 
changed accordingly in the SAR.
    Comment G.21: One commenter stated that a note should be added to 
drawing 1049-70-4 specifying that a test fitting may be supplied on the 
access port cover plate.
    Response: The NRC agrees with this comment. The addition of this 
fitting does not affect safety. Its purpose is to facilitate leak 
testing of the overpressure monitoring system. The drawing has been 
revised to reflect this change.
    Comment G.22: One commenter stated that a note should be added to 
drawing 1049-70-7 allowing alternate configurations for the plumbing of 
the pressure monitoring system.
    Response: The NRC agrees with this comment. 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 and leak testing and monitoring of the 
system as required by the technical specifications. The drawing has 
been revised to reflect this change.
    Comment G.23: One commenter stated that the vent and drain port 
cover seal groove diameters on drawing 1049-70-3 should be changed as 
follows; 5.88 groove O.D. to 5.92, and 4.70 I.D. to 4.65.
    Response: The NRC agrees with this comment. The changes to the 
drawing do not affect the structural design or the confinement 
boundary. A note has been added to the drawing to follow the 
manufacturer's recommendations.
    Comment G.24: One commenter stated that in SAR Chapters 2 and 7 the 
metallic O-ring seal liners should be specified as stainless steel or 
nickel alloy.
    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 revised.
    Comment G.25: One commenter asked how the bottom plate is welded to 
the confinement shell, how the gamma shield bottom plate is welded to 
its shell, how the plates are arranged, how weld locations affect 
stresses, what the actual stresses are, and what mechanism could cause 
the plates to be detached.
    Response: The weld between the bottom confinement plate and the 
confinement shell is a complete penetration weld. The weld has the same 
thickness as the plate and shell. Therefore, it makes no difference 
whether the weld is outside or inside the shell. The gamma shield shell 
rests on the bottom gamma shield plate and is welded all around the 
outside perimeter of the joint. Weld locations are included in the 
finite element model. Stress intensities for different cask components 
and welds for each service condition and the load combination are 
presented in Appendix 3A of the SAR. The bottom plate could become 
detached from the gamma shield shell if the weld connecting the gamma 
shield shell to the bottom plate were to fail completely. The mechanism 
for possible failure of this weld is discussed in Appendix 3E of the 
SAR. Special examinations are required for this weld to ensure that 
defects are detected and repaired before use for fuel storage. These 
requirements are presented in Appendix 3E of the SAR and discussed in 
Section 3.1.4.4 of the SER.
    Comment G.26: One commenter stated that at this point in time TN 
should know if the bottom inner plate weld is going to be applied 
before or after the outer and inner shell assembly. The commenter asked 
if it was the shrink fit and why TN did not appear to know. The 
commenter stated an understanding that one shell has a seam and the 
carbon steel is wrapped into a cylinder and welded at the one meeting 
seam, while the other shell is in two halves requiring two seams and 
asked if that was correct. Then the commenter asked if there is a 
concern if one seam is located over or near the seam of the other, if 
the plate pushes out at the shell wall around its thickness, or if the 
shell of either the containment or gamma shield rests on their bottom 
plates, how this affects the weight distribution, how these two shells 
are put together, when welding is to be performed, and exactly how the 
welding will be inspected. The commenter noted that the use of the word 
``if'' in the acceptance test section of the SER is not acceptable 
because the level of detail in the design and fabrication should be 
decided before a design is certified.
    Response: The NRC disagrees with the comment. 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 be in conformance with a standard that will satisfy all of the 
safety requirements for this application. No adverse effects on the 
cask integrity is expected from either of the two fabrication 
alternatives; either alternative is acceptable. Therefore, the SAR can 
specify welding either before or after shell assembly. See the 
discussion for Comment G-2 about how the shells are assembled.
    The steel of the seam meets the requirements of the steel used for 
the vessel. Location of seams in relation to one another will not 
affect performance. In terms of any alterations in stress (or weight 
distribution), it is noted that the containment vessel (and its seams) 
is ground to tight tolerances so that it will be exactly the right size 
to make the shrink fit process work. Circumferential and longitudinal 
confinement boundary welds are examined volumetrically by radiography 
and liquid penetrant or magnetic particle methods accepted by ASME NB-
5000 standards. 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 fabricator 
can choose either ultrasonic or radiographic examinations to inspect 
the corner weld. In this case, 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. Non-confinement welds are inspected in accordance with 
the ASME Code, Subsection NF. Additional inspections will also be 
performed on the gamma shield shell to the bottom shield weld

[[Page 14804]]

and the lid to the shield lid weld as specified in SAR Section 9.
    Comment G.27: One commenter stated that using mirrors and auxiliary 
lighting to inspect welds that were not directly visible ``sounded 
tricky.'' The commenter noted that ensuring that the basket retains its 
form throughout its life is important and asked the NRC to clarify what 
a plug weld is and how they are inspected.
    Response: The NRC has accepted a number of methods to visually 
inspect hardware to verify materials quality, including the use of 
mirrors and auxiliary lighting as appropriate. The basket will retain 
its shape over the life of the containment system because it is 
fabricated using acceptable methods. Also, the cask is filled with 
helium that precludes environmentally induced alterations. Further, the 
basket is designed to accommodate the thermal cycles of the application 
without substantial distortions. The plug weld technique is used to 
connect the stainless steel tubes together as part of the fuel basket 
using solid stainless steel connecting bars. Each plug weld penetrates 
the full thickness of the stainless steel tube wall. These welds are 
not only 100 percent visually inspected, but sample coupons made by the 
same welding procedures, technique, and weld machine are tested to 
verify quality.

H. Technical Specifications

    Comment H.1: One commenter stated that the maximum uranium content 
should be deleted from Section 2.1 of the TSs because this information 
is already included in the SAR.
    Response: The NRC disagrees with this comment. This design 
information is crucial to the conclusions reached by the NRC about the 
TN-32 design in its SER. The maximum uranium masses, along with other 
fuel parameters, include the design tolerances considered in the SAR 
and, therefore, are not overly restrictive. The uranium content in the 
TSs are set to bound all potential variations for the design. Further, 
the NRC considers the maximum uranium content to be a fuel parameter 
that is a part of the design that can not be changed without NRC review 
and approval. Therefore, it should remain in the TSs.
    Comment H.2: One commenter stated that the parameter labeling of 
Table 2.1.1-1 of the TSs should be revised as Minimum Initial 
Enrichment and Maximum Burnup to avoid confusion.
    Response: The NRC agrees with this comment. 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/B2.2) located in Chapter 12 of the SAR.
    Comment H.3: One commenter stated that the frequency for 
Surveillance Requirement (SR) 3.1.3 and 3.1.4 should use the term 
TRANSPORT OPERATIONS for consistency.
    Response: The NRC agrees with this comment. The affected TSs have 
been changed to use the term TRANSPORT OPERATIONS.
    Comment H.4: One commenter stated that the frequency of SR 3.1.6.1 
should be revised to state ``immediately prior to lifting the cask . . 
.''.
    Response: The NRC agrees with this comment. The 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 H.5: One commenter stated that the applicability of SR 
3.2.1 should be revised to ``during TRANSPORT OPERATIONS'.
    Response: The NRC disagrees with this comment because it is not 
necessary to include this information in the body of the TS. However, 
it is appropriate for clarity to insert a comment in the basis for the 
TS (B3.2.1) located in Chapter 12 of the SAR. The SAR has been revised 
accordingly.
    Comment H.6: One commenter stated that the cell opening and boron 
loading should be removed from Section 4.1.1 of the TSs.
    Response: The NRC disagrees with this comment. This design 
information is crucial to the conclusions reached by the NRC in its 
SER. The minimum boron loading and the minimum cell opening for the 
basket include any design tolerances included in the SAR. Design 
features that may affect safety if altered or modified are included in 
the TSs.
    Comment H.7: One commenter stated that the Codes and Standards 
Section, 4.1.3, of the TSs should be removed.
    Response: The NRC disagrees with the comment. This information is 
required under 10 CFR 72.24(c)(4).
    Comment H.8: One commenter stated that in the storage location for 
Casks, 4.2.1 of the TSs, the 16-foot dimension should be listed as a 
minimum value or a tolerance should be added.
    Response: The NRC does not agree with this comment to add a 
tolerance. As written, the TSs state 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 H.9: One commenter stated that references to consideration 
as important to safety [for a berm] be removed from Section 4.3.6 of 
the TSs.
    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 (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. See NUREG/CR 6407, ``Classification of Transportation 
Packaging and Dry Spent Fuel Storage Components According to Importance 
to Safety'' February 1996. Generally, 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 H.10: One commenter stated that the proposed TN-32 TSs are 
confusing, more complicated than those of the VSC-24, and are not 
written in plain English. For example, the commenter noted that 1.3 
``completion times'' on page 1.3-2 is confusing with too many words.
    Response: The NRC disagrees with the comment. The TN-32 TSs are 
modeled on the improved Standard Technical Specifications (ISTS) for 
power reactors. The ISTS were developed as the result of extensive 
technical meetings and discussions between the NRC staff and the 
nuclear power industry in the early 1990s in an effort to improve 
clarity and consistency of the power reactor TSs and to make them 
easier for operators to use. The most likely users of the TN-32 spent 
fuel storage cask technical specifications are power reactor licensees 
familiar with the format of the ISTS. Although different in form than 
the VSC-24 TSs, the NRC staff believes that the format of the proposed 
TN-32 TSs will make them easier for operators to use and will help to 
achieve consistency between power reactor TSs and spent fuel dry cask 
storage TSs. The NRC staff also believes that the specific wording of 
Section 1.3, ``Completion Times'' helps to clarify the TSs by walking 
the user through each step in

[[Page 14805]]

detail and by explaining the conditions, the required action, and the 
allowable time to complete the required action.
    Comment H.11: One commenter requested an explanation of SR 3.0.2. 
The commenter stated that the use of 1.25 times the interval specified 
was confusing and that workers should have definite clear directions. 
One commenter questioned the 25 percent extension of time allowed by SR 
3.0.2. The commenter stated that the surveillance should not be missed 
and should be completed on time.
    Response: The basis for SR 3.0.2 is discussed in the TN-32 
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 finds 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 when a surveillance frequency is specified by a 
regulation because the requirements of the regulations take precedence 
over TSs. The NRC staff believes that the provisions of SR 3.0.2 are 
clear to users of the TSs and will ensure that all required 
surveillances will be performed within an acceptable time period, 
consistent with the NRC staff's safety analyses.
    Comment H.12: One commenter asked the frequency of alarm checks and 
calibration for accuracy. The commenter stated that automatic testing 
and alarms at the plant should be developed. The commenter also stated 
that the testing interval of every 36 months for the channel 
operational test (COT) in SR 3.1.5.2 was inadequate due to the 
importance of the pressure switch.
    Response: The NRC agrees that the instrumentation for monitoring 
the seals is important and that is why the NRC required TSs for 
surveillance of this instrumentation. The surveillance requirements for 
the cask interseal pressure monitoring (e.g. alarm checks and 
calibration frequency) are given in SR 3.1.5.1 and SR 3.1.5.2.
    SR 3.1.5.1 requires monitoring of the interseal pressure at a 
frequency of once per 7 days. This check ensures that the condition of 
the alarm is verified at an acceptable frequency. The surveillance 
frequency is acceptable to the NRC staff based on the measures to 
verify the integrity of the cask seals and the pressure monitoring 
system during cask preparations, the static and passive nature of the 
seals, and the low likelihood that the seal or the monitoring system 
will develop a leak after placing them into service.
    SR 3.1.5.2 requires a channel operational test (COT) of the 
pressure monitoring instrumentation at a frequency of once every 36 
months. According to the TS definition, the COT tests the pressure 
sensing instrumentation and low pressure indication feature by 
injecting an actual or simulated signal as close to the sensor as 
possible to verify the operability of the alarm functions. The COT also 
includes adjustments, as necessary, of the required alarm setpoint so 
that the setpoint is within the required range and accuracy. Section 
8.3 of the SAR that was reviewed and accepted by the NRC further 
describes the COT for the TN-32.
    By establishment of requirements in the TSs, the NRC imposed 
minimum performance requirements of the equipment used for the 
overpressure monitoring system. It is incumbent on the general licensee 
to procure and install pressure monitoring equipment on the TN-32 cask 
that has acceptable reliability. This would include having provisions 
for instrument drift to ensure that the requirements for continuous 
monitoring of the cask seal are met. Based on considerable industry 
experience with instrumentation, suitable instrumentation that meets 
the performance requirements for the TN-32 is available. Further, the 
monitoring frequency is also acceptable to the NRC staff based on the 
measures to verify the integrity of the cask seals, the pressure 
monitoring system during cask preparations, and the low likelihood that 
the seal or the system will develop a leak after placing them into 
service.
    Regarding the commenter's suggestion of creating an automated or 
computerized method for testing the instrumentation, details on site-
specific design of this equipment is beyond the scope of this rule.
    Comment H.13: One commenter stated that the measurement location 
for the cask temperature in SR 3.1.6.1 should be outside the auxiliary 
loading area in ambient conditions before the cask starts the transfer 
to the ISFSI. The commenter further stated that additional criteria is 
needed to specify the location because the cask temperature should not 
be taken next to a heater inside the building.
    Response: The NRC disagrees with this comment. The cask body 
temperatures are not going to undergo rapid change induced by ambient 
conditions. This is because the mass of the cask is so large. It is the 
cask user's responsibility to ensure that the temperature measurement 
represents the actual temperature of the outer surface of the cask 
rather than some other heat source that might be located in the 
vicinity of the cask. This level of detail is beyond the scope of this 
rule.
    Comment H.14: One commenter asked for clarification of the 
limitations on changes discussion in LCO 3.0.4. The commenter felt that 
the way the LCO was worded is ambiguous because it allows actions to be 
taken that are not endorsed.
    Response: The ultimate purpose of LCO 3.0.4 is to allow the cask to 
be placed in a safe condition in accordance with the Required Action(s) 
of the governing TS. LCO 3.0.4 precludes placing the cask in an 
unacceptable condition; specifically one in which the governing LCO 
would not be met in the Applicability desired to be entered, or if that 
Applicability would have to be exited at a certain time to comply with 
the Required Actions. LCO 3.0.4 does not allow actions to be taken that 
are not approved by the NRC staff.
    Comment H.15: One commenter stated that a clarification to the code 
should be made in Table 4.1-1 of the TSs that the weld of the lid 
shield plate to the lid is not impact tested.
    Response: The NRC agrees with the comment and the TSs have been 
changed accordingly.
    Comment H.16: One commenter stated that in Section 4.1.3, and Table 
4.1-1 of the TSs, all references to NB should be changed to NF for the 
basket.
    Response: The NRC agrees with the comment because the TN-32 is a 
storage only cask, and have changed the TSs accordingly.

I. Miscellaneous

    Comment I.1: One commenter pointed out that an NRC letter and 
technical report calculation numbers are not the same.
    Response: The NRC agrees with this comment. The SAR has been 
revised to correct the discrepancy.
    Comment I.2: One commenter stated that the SAR Rev. 11A references 
an old technical report revision date.
    Response: The NRC agrees in part with this comment and has 
determined that the technical report referenced by the applicant in the 
SAR was the one used in the supporting analysis and is not the most 
recent version. The NRC has determined that the information used in the 
supporting analysis is consistent with that included in the more recent 
revision. Therefore, using the more recent revision would not

[[Page 14806]]

impact the applicant's analysis and the NRC requires no update to the 
reference in the SAR.
    Comment I.3: One commenter asked if a vacuum pump fails while a 
cask is filled with air and some water, how long could workers take to 
fix the pump before heat up took place in the cask?
    Response: The time and rate of heatup of a cask partially filled 
with water would depend on the type of fuel, its burnup, and 
enrichment. According to SR 3.1.1.1, vacuum drying must be complete 
within 24 hours of the completion of cask draining. Therefore, if 
vacuum drying is not complete for whatever reason by the 24-hour 
period, specific actions are required by TS 3.1.1 to place the fuel in 
the desired safe condition.
    Comment I.4: One commenter asked for information about Interim 
Staff Guidance (ISG) 7 referred to in the SER. Further, the commenter 
asked what different cask is referred to, if partial helium injection 
is effective, and if it has been tested. Also, the commenter recommends 
that testing be conducted for the TN-32.
    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 breakdowns 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 the VSC-24, 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 percent 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 percent. The NRC staff did not review nor 
require any testing of the helium injection process based on the 
analysis and the restrictions imposed by the TSs on operations without 
a full helium environment.
    Comment I.5: One commenter suggested that an unloading test should 
be done to see what would happen. The commenter asked how the check 
valve is put into the documents [procedures], how workers can validate 
this, what water level is in the cask with how much space above, can 
hydrogen accumulate in that space, and if the draining and venting is 
performed through connected hoses. The commenter also suggested that 
the procedure is dangerous, could be confusing for a new worker, and 
that figure 8.2-1 of the SAR should be added to the SER for clarity. 
Further, the commenter asked a number of questions about the reflooding 
evolution: what happens to steam and if hydrogen can form and mix and 
could exit the cask; what other chemical reactions could occur, if 
paint, crud or BPRAs pieces, or bits of aluminum could fall and clog 
equipment; what would occur if cooling water were put in at the top 
hole instead of in the drain pipe at the bottom of the cask; and if the 
SER and SAR provide sufficient and correct guidance on the fill, vent, 
and drain opening for loading and unloading.
    Response: The NRC disagrees with this comment about testing. 
Testing is normally required when the analytic methods have not been 
validated or assured to be appropriate and/or conservative. In lieu of 
testing, the NRC finds analytic conclusions that are based on sound 
engineering methods and practices to be acceptable. The TN-32 Dry 
Storage Cask design including the unloading process has been reviewed 
by the NRC. The basis of the safety review and findings are clearly 
identified in the SER and CoC. In addition, as a condition of the CoC, 
each cask user must demonstrate the ability to unload a cask as a part 
of its pre-operational testing and training exercise. The demonstration 
of the ability to unload a cask, in combination with NRC staff review 
and acceptance of the analyses performed by TN, provides reasonable 
assurance that the TN-32 cask can be safely unloaded.
    The unloading process including the check valve is described in TN-
32 SAR Section 8.2. Detailed site-specific procedures for performing 
unloading operations are required to be developed and demonstrated at 
each facility that uses the TN-32. Cask users are required to provide 
adequate procedures, training, and quality oversight to ensure that the 
procedure actions are performed as required. The vent and drain ports 
have different size pipe threads in order to aid in precluding any 
confusion for the worker. A note has been added to the SAR drawing and 
Chapter 8 for clarification.
    For hydrogen generation to occur, there must be either a chemical 
interaction between the water in the cask and cask materials or 
radiolysis of the water. Hydrogen generation itself is not a safety 
problem because there must also be conditions that allow for 
accumulation of hydrogen and air (or oxygen) to an ignitable mixture 
and an ignition source. For the materials present in the TN-32 cask, 
the rate of hydrogen generation is low when compared to other materials 
such as zinc based coatings. The applicant provided an evaluation of 
hydrogen in the TN-32 SAR, Section 3.4.1.4, that addressed hydrogen 
generation, measures to preclude hydrogen accumulation, and that the 
TN-32 does not have any ignition sources because the cask closure is 
bolted.
    During loading, the cask is completely filled with water and 
continuously vented which precludes the accumulation of hydrogen. For 
the cask draining operation, the cask remains vented. The applicant 
concluded that the hydrogen buildup in a 2-hour period (the expected 
time for draining) would be well below the ignitable limit of 4 
percent. Vacuum drying is performed after draining. In this condition 
there is no longer a source of hydrogen generation.
    During cask reflood, the cask is continuously vented, precluding 
the accumulation of hydrogen. The cask fill gas and possibly steam will 
be forced out of the cask through the vent until the cask is full and 
the reflood is complete. After the reflood is complete, the cask 
remains vented as it is placed in the pool and the lid removed. The 
procedure descriptions in the TN-32 SAR, Section 8, include specific 
provisions for venting of the cask during times when the cask is filled 
with water such as during draining and reflood operations.
    The NRC staff reviewed and accepted the analysis of hydrogen 
generation and procedure descriptions to load and unload the TN-32 cask 
in Preliminary SER Sections 3.1.4.1 and 8. A discussion of crud 
development is included in the response to Comment I.13.
    Comment I.6: One commenter stated that based on the information in 
the documents that pressurized water reactor (PWR) fuel burned to 
45,000 MWD/MTU with a 6-year cooling time can not be loaded in the cask 
because it increases the neutron source by 12%.
    Response: The NRC agrees with the comment. As specified in Table 
2.1.1-1 of the TSs, fuel must be cooled at least 7 years before it can 
be stored in the TN-32 cask.
    Comment I.7: One commenter asked how pinhole leaks and hairline 
cracks can be detected or seen in rods in the middle of an assembly, 
how many of these defects are permitted in one rod, (as many as 100?), 
what is the acceptable defect size, if blisters and crud can be 
present, if a rod or BPRAs can be depressurized, and if utilities or 
the NRC are clear on what is acceptable.

[[Page 14807]]

    Response: An example of pinhole leaks and hairline cracks is given 
in SAR Section 6A.3. Only assemblies that are intact are allowed in the 
TN-32. The TN-32 meets the criticality safety requirements of 10 CFR 
Part 72 without any additional fuel condition requirements. The 
criteria for an intact assembly are defined in TS Section 1.1 as fuel 
assemblies without known or suspected cladding defects other than 
pinhole leaks or hairline cracks and can be handled by normal means. 
Partial fuel assemblies (fuel assemblies with missing fuel rods) must 
not be classified as intact fuel assemblies unless dummy fuel rods are 
used to displace an amount of water greater than or equal to that 
displaced by the original fuel rods. As proof that the fuel to be 
loaded is undamaged, the NRC will accept, as a minimum, a review of the 
records to verify that the fuel is undamaged, followed by an external 
visual examination of the fuel assembly before loading to identify any 
obvious damage. For fuel assemblies where reactor records are not 
available, the level of proof will be evaluated on a case-by-case 
basis. The purpose is to provide reasonable assurance that the fuel is 
undamaged. Depressurized rods and BPRAs do not impact the safe 
operation of the cask as discussed in the response to F.3 above.
    Comment I.8: One commenter asked how helium purity is tested, if it 
will be done and if it will be in the documents [procedures].
    Response: Testing or sampling for helium purity is performed by the 
helium supplier and certified to the cask user upon delivery. TS 4.1.4 
requires that the cask be filled with helium with a purity of at least 
99.99 per cent and documented accordingly. The purity of the helium 
will be controlled under the licensee's quality assurance program. Only 
pure helium will be used to backfill the cask; no other gasses will be 
added during backfill. Acceptable helium purity for dry spent fuel 
storage casks was defined by R. W. Knoll et al. At Pacific Northwest 
Laboratory in, ``Evaluation of Cover Gas Impurities and Their Effects 
on the Dry Storage of LWR Spent Fuel'' PNL-6365, November 1987.
    Comment I.9: One commenter asked if the 0.10 fraction for release 
of full fines is valid, if there has been any more testing after the 
1992 Sandia report, and if anything new has been conducted after the 
1980 rod burst tests.
    Response: The NRC staff has accepted the 0.10 fraction for 
releasability of fuel fines for the TN-32. The basis for this 
acceptance is provided in the TN-32 SER Section 7.3. The NRC staff does 
not have information on experiments or testing more recent than that 
referenced in the SER.
    Comment I.10: One commenter asked why there is a progression to 
backfill twice with helium.
    Response: This process ensures a high confidence that residual 
moisture and oxidizing impurities are removed from the cask cavity. It 
is a recommendation of PNL-6365, ``Evaluation of Cover Gas Impurities 
and Their Effects on the Dry Storage of LWR Spent Fuel'' November 1987.
    Comment I.11: One commenter asked what would happen if a gas sample 
found that helium had been lost , if a water sample reflected crud 
particulates, paint flakes, or parts of a deteriorated BPRA or TPA. The 
commenter noted a concern that the above materials in the water could 
clog equipment and require filter changes in the pool if mixed with 
spent fuel pool water. The commenter suggested that some equipment to 
filter the cask water before to mixing with pool water is needed along 
with a filtration system to control gas releases from the cask.
    Response: There is not a requirement to sample the cask cavity for 
the presence of helium. The cask is designed and analyzed to maintain a 
helium environment for the duration of the authorized storage period. 
However, if helium is hypothetically assumed to not be present in a 
cask during storage, there is a possibility that the fuel may be 
degraded. Any leakage from these postulated degraded fuel rods would be 
retained by the cask confinement system that acts as a barrier to 
releases of radioactive materials to the environment. Specific actions 
are outlined in the unloading procedure descriptions in SAR Table 8.2-1 
regarding analyzing a gas sample for radioactive material (to detect 
degraded fuel). If degraded fuel is detected, appropriate actions are 
required for the cask user to develop procedures to minimize exposures 
to workers and releases to the environment. A requirement to sample the 
water discharged from the cask during reflood operations is beyond the 
scope of this rule.
    The NRC disagrees with the recommendation to add a filtration 
system to cask water because the spent fuel pool already has a 
filtration and purification system in place. Further, the design of the 
cask precludes the need for a gas filtration system. A discussion of 
crud development is included in the response to Comment I.13.
    Comment I.12: One commenter discussed the use of poured resin 
material, the importance of procedures for mixing and pouring, the need 
for detailed procedures for workers who may never have worked on 
nuclear application material, the need for management supported work 
ethics, the need to report and correct mistakes, and the need for 
production workers making the boron aluminum sheets to be aware of the 
effects of flaw removal. The commenter asked if there are clear 
criteria for inspection and testing of resin material by the 
fabricator, and what the measures are to ensure the absence of voids in 
resin material and if they are clear.
    Response: The fabrication of the resin neutron shield will be 
performed under specific controls and procedures to provide a uniform 
and effective material. Radiation surveys performed around the cask 
after loading are designed to detect flaws or mistakes that will 
adversely affect the ability of the cask to meet the offsite dose 
limits. Fabrication, testing, and repair of the components in the cask 
important to safety are covered by an NRC approved quality assurance 
program either directly or as a supplier or subcontractor to a holder 
of a QA program. The applicable QA requirements are contained in 10 CFR 
Part 72 Subpart G.
    Comment I.13: One commenter asked a number of questions on the cask 
filling and venting process and how the procedures will preclude 
ignition of hydrogen. The commenter asked how the cask filling process 
works and if the fill and drain lines vent gases during cask filling 
and if hydrogen can form during the process; if steam, hydrogen, paint 
flakes, or crud (debris) will fall into the fuel basket and between 
rods and clog the drain lines, and what happens to these materials 
during the fill process; if effluent flows from the cask to the fuel 
pool and affects pool water quality or reacts chemically with materials 
; and if casks can be safely unloaded based on the above.
    Response: The filling and venting process are discussed in the 
response to Comment I.5. Except for crud, the NRC staff does not expect 
paint flakes, particles or debris in the TN-32 cask because the coating 
on the cask interior is flame sprayed aluminum that is a tightly 
adherent and stable coating in the spent fuel storage environment and 
the other cask materials do not create debris during any of the 
expected conditions in the cask. Some crud may be dislodged from the 
fuel cladding during spent fuel dry storage, but the crud particles for 
PWR fuel are very small with diameters ranging from 1 to 3 micro-meters 
as reported in SAND88-1358, ``Estimate of CRUD Contribution to Shipping 
Cask Containment

[[Page 14808]]

Requirements.'' Particles of this size do not pose a clogging concern 
in vent/drain lines for this cask. Apart from the crud, no materials 
other than water and steam are discharged to the pool, where crud from 
wet fuel storage is already present. The amount of crud from the spent 
fuel cask is expected to be very small and would be captured in the 
spent fuel pool filtration system. Crud is generally made up of metal 
oxides that are not chemically reactive.
    The unloading process is outlined in section 8.2 of the TN-32 SAR 
along with supporting analysis in sections 3 and 4. The NRC staff 
reviewed and accepted the operating descriptions and analysis, and 
concluded in the SER that there was reasonable assurance that the casks 
could be safely unloaded by qualified personnel using detailed 
procedures developed by the cask user at an ISFSI site.
    Comment I.14: One commenter asked the basis for the 24 hour 
timeframe for conducting the dryness test; the basis for stating that a 
high vacuum is an indication that the cavity is dry; what analysis 
provides the basis for the height of the vacuum and whether the 
analysis is for the specific materials in the cask; the definition of a 
dry cavity and whether it includes the aluminum paint, drain pipe, 
bottom plate, zircaloy, pellets, etc.; how do you really know that the 
contents of the cavity are dry.
    Response: The basis for the 24-hour time limit to achieve the 
required vacuum and cask dryness is discussed in detail in TS bases 
B.3.1.1. The purpose of the time limit is to prevent the temperature of 
the basket components from exceeding their analyzed temperature range. 
A high vacuum ensures that most of the moisture will be removed from 
all components in the cavity including coatings. The vacuum drying 
process is further discussed in the response to Comment G.7.
    Comment I.15: One commenter asked if an analysis has ever been 
completed to see what happens when the fuel and other cavity contents 
are dried out and then placed back in the pool. The commenter asked how 
the materials react with the pool water and whether it affects the 
pool.
    Response: The information provided in SAR Section 3.4.1 discusses 
material interactions and would apply to when the fuel and cavity were 
dried out and placed in the pool with the introduction of water to the 
materials. The NRC staff has determined that this information is 
complete and acceptable.
    Comment I.16: One commenter asked if it is appropriate to use the 
same procedures to dry the cask out after being in the pool for seven 
days and if the process is still accurate.
    Response: The procedures used for drying the cask and the expected 
materials and fuel interactions are discussed in the response to 
Comment G.7. The procedures are applicable to an exposure of the cask 
to pool water of any duration.
    Comment I.17: One commenter asked if there was a weld problem in 
Precision Components Fabrication and how the problem was resolved. The 
commenter believed there was a concern with the shims and asked where 
the shims are located and how they are removed at unloading.
    Response: Shims are not used in the TN-32 cask closure design; the 
lid is bolted on and not welded. Questions related to the fabrication 
activities at Precision Components Fabrication are beyond the scope of 
this rule.
    Comment I.18: One commenter asked if the TN-RAM shipping problems 
had been resolved and if this was a concern for the TN-32 design.
    Response: This comment is beyond the scope of this rule that deals 
with a storage cask design.
    Comment I.19: One commenter stated that the gaps (in welds) was one 
of the real concerns for the TN-32 and asked what are the gaps.
    Response: Gap welds are not a concern with the TN-32 cask design 
because the lid is bolted not welded in place. Therefore, this comment 
is not applicable to this rule.
    Comment I.20: One commenter asked if the documents at Transnuclear 
and the subcontractors were controlled according to their Quality 
Assurance (QA) program. The commenter stated that there had been some 
problems with the Transnuclear QA manual and asked if the problem was 
now resolved. The commenter further asked if workers understand what a 
defect is and if the QA program clearly defines a defect. The commenter 
stated that there should be a requirement to store documents in process 
in fireproof boxes at the end of each work day.
    Response: The NRC recognizes the relationship of the comment with 
the inspection findings noted in NRC Inspection reports 71-0250/97 and 
72-1021/97-206. The inspection findings were addressed and the 
resolution was reviewed by the NRC. TN was notified that their response 
was acceptable and that no further information was required in NRC 
letters to Transnuclear dated July 28, 1997, and August 8, 1997. There 
is no regulatory requirement or applicant procedure to store design 
documents in fireproof boxes.
    Comment I.21: One commenter asked if the SAR has been updated and 
if it will have another update with the CoC approval.
    Response: The applicant has revised the SAR in response to 
rulemaking comments and questions before CoC issuance. The final 
version issued with this rule is available in the NRC Public Document 
Room.
    Comment I.22: One commenter asked if soil liquefaction has been 
adequately addressed in the TN-32 pad design.
    Response: Soil liquefaction is a site-specific issue and is beyond 
the scope of this rule that adds a generic cask design to the listing.
    Comment I.23: One commenter asked if the Surry cask design has been 
amended to use increased burnup and enrichment.
    Response: Virginia Power has submitted a request to the NRC to 
amend their cask design to permit storage of fuel with higher 
enrichment and with higher burnup. This request will be reviewed by the 
NRC staff.
    Comment I.24: One commenter asked if ``assembly line methods'' of 
fabrication are causing problems and multiple non-conformance for the 
TN-32 design, and if there are problems that should be resolved.
    Response: The NRC is not aware of any fabrication methods that have 
caused problems or non-conformance with the TN-32 design.

Summary of Final Revisions

    As a result of the staff's response to public comments, or to 
rectify issues identified during the comment period, the following 
items in the TSs have been modified: TS 1.1 (staff initiative), TS 2.1 
(staff initiative), Table 2.1.1-1 (see comment H.2), TS 3.1.3 (see 
comment H.3), TS 3.1.4 (see comment H.3), TS 3.1.6.1 (see comment H.4), 
TS 4.1.3 (see comment H.16), Table 4.1-1 (see comment H.15 and H.16), 
and TS 5.2.3 (staff initiative).
    The proposed CoC has been revised to clarify the requirements for 
making changes to the CoC by specifying that the CoC holder must submit 
an application for an amendment to the certificate if a change to the 
CoC, including its appendices, is desired. The CoC has also been 
revised to delete the proposed exemption from the requirements of 10 
CFR 72.124(b) because a recent amendment of this regulation makes the 
exemption unnecessary (64 FR 33178; June 22, 1999). The staff has also 
updated the CoC, including the addition of explicit conditions 
governing acceptance tests and maintenance program, approved contents, 
design features, and

[[Page 14809]]

authorization, and has removed the bases section from the TSs attached 
to the CoC to ensure consistency with NRC's format and content. In 
addition, other minor, nontechnical changes have been made to CoC 1021 
to ensure consistency with NRC's new standard format and content for 
CoCs. The NRC staff has also modified its SER. The NRC staff has also 
modified the rule language by changing the word ``Certification'' to 
``Certificate'' to clarify that it is the Certificate that expires.

Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the Commission on June 30, 1997, 
and published in the Federal Register on September 3, 1997 (62 FR 
46517), this rule is classified as compatibility Category ``NRC.'' 
Compatibility is not required for Category ``NRC'' regulations. The NRC 
program elements in this category are those that relate directly to 
areas of regulation reserved to the NRC by the Atomic Energy Act of 
1954, as amended (AEA), or the provisions of Title 10 of the Code of 
Federal Regulations. Although an Agreement State may not adopt program 
elements reserved to NRC, it may wish to inform its licensees of 
certain requirements via a mechanism that is consistent with the 
particular State's administrative procedure laws, but does not confer 
regulatory authority on the State.

Finding of No Significant Environmental Impact: Availability

    Under the National Environmental Policy Act of 1969, as amended, 
and the Commission's regulations in Subpart A of 10 CFR Part 51, the 
NRC has determined that this rule is not a major Federal action 
significantly affecting the quality of the human environment and 
therefore, an environmental impact statement is not required. This 
final rule 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 Merri Horn, 
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone (301) 415-8126, 
e-mail [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 is adding the Transnuclear TN-
32 cask system to the list of NRC-approved cask systems for spent fuel 
storage in 10 CFR 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.
    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 above problems and is consistent with previous 
Commission 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 above discussion of the benefits and impacts of the 
alternatives, the NRC concludes that the requirements of the final rule 
are commensurate with the Commission's responsibilities for public 
health and safety and the common defense and security. No other 
available alternative is believed to be as satisfactory, and thus, this 
action is recommended.

Small Business Regulatory Enforcement Fairness Act

    In accordance with the Small Business Regulatory Enforcement 
Fairness Act of 1996, the NRC has determined that this action is not a 
major rule and has verified this

[[Page 14810]]

determination with the Office of Information and Regulatory Affairs, 
Office of Management and Budget.

Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 
605(b)), the Commission certifies that this rule will not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This 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 (10 CFR 50.109 or 10 
CFR 72.62) does not apply to this rule because this amendment does not 
involve any provisions that would impose backfits as defined in the 
backfit rule. Therefore, a backfit analysis is not required.

List of Subjects in 10 CFR Part 72

    Criminal penalties, Manpower training programs, Nuclear materials, 
Occupational safety and health, Reporting and recordkeeping 
requirements, Security measures, Spent fuel.


    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting 
the following amendments to 10 CFR part 72.

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

    1. The authority citation for Part 72 continues to read as follows:

    Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 10d-
48b, sec. 7902, 10b Stat. 31b3 (42 U.S.C. 5851); sec. 102, Pub. L. 
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, 
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 
10155, 10157, 10161, 10168).

    Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), 
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 
10168(c),(d)). Section 72.46 also issued under sec. 189, 68 Stat. 
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub. 
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also 
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2244, (42 U.S.C. 10101, 
10137(a), 10161(h)). Subparts K and L are also issued under sec. 
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 
(42 U.S.C. 10198).


    2. In Sec. 72.214, Certificate of Compliance 1021 is added to read 
as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1021.
    SAR Submitted by: Transnuclear, Inc.
    SAR Title: Final Safety Analysis Report for the TN-32 Dry Storage 
Cask.
    Docket Number: 72-1021.
    Certificate Expiration Date: April 19, 2020.
    Model Number: TN-32, TN-32A, TN-32B.

    Dated at Rockville, Maryland, this 6th day of March, 2000.

    For the Nuclear Regulatory Commission.
William D. Travers,
Executive Director for Operations.
[FR Doc. 00-6630 Filed 3-17-00; 8:45 am]
BILLING CODE 7590-01-P