[Federal Register Volume 65, Number 84 (Monday, May 1, 2000)]
[Rules and Regulations]
[Pages 25241-25265]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-10393]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG 31


List of Approved Spent Fuel Storage Casks: Holtec HI-STORM 100 
Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to add the Holtec HI-STORM 100 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 May 31, 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 Holtec HI-STORM 100 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, 
Holtec International submitted an application for NRC approval with the 
Safety Analysis Report (SAR) entitled ``Topical Safety Analysis Report 
for the HI-STORM 100 Cask System.'' The NRC evaluated the Holtec 
International submittal and issued a preliminary Safety Evaluation 
Report (SER) and a proposed Certificate of Compliance (CoC) for the 
Holtec HISTORM 100 cask system. The NRC published a proposed rule in 
the Federal Register (64 FR 51271; September 22, 1999) to add the 
Holtec HI-STORM 100 cask system to the listing in 10 CFR 72.214. The 
comment period ended on December 6, 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, including its 
appendices, the Technical Specifications (TSs), and the Approved 
Contents and Design Features, for the Holtec HI-STORM 100 cask system. 
The NRC staff has also modified its preliminary SER. Finally, comments 
were received from other industry organizations suggesting changes to 
the TSs and the Approved Contents and Design Features. Some of these 
were editorial in nature, others provided clarification and 
consistency, and some reflected final refinements in the cask design. 
The NRC staff agrees with many of these suggested changes and has 
incorporated them into the final documents, as appropriate. The NRC 
staff has also modified the rule language by changing the word 
``Certification'' to ``Certificate'' to clarify that it is actually the 
Certificate that expires.
    The NRC finds that the Holtec International HI-STORM 100 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 Holtec HI-STORM 100 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 Holtec HI-STORM 100 cask system under the general license in 10 
CFR Part 72, Subpart K, by holders of power reactor operating licenses 
under 10 CFR Part 50. Simultaneously, the NRC is issuing a final SER 
and CoC that will be effective on May 31, 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 a industry users group, two members of the public, 
and a State. 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 Holtec HI-STORM 100 Cask System

    The comments and responses have been grouped into eleven areas: 
General, radiation protection, accident analysis, criticality, design, 
welds, structural, materials, thermal, technical specifications, and 
miscellaneous. Several of the commenters provided specific comments on 
the draft CoC, the NRC staff's preliminary SER, the TSs, and the 
applicant's SAR. Some of the editorial comments have been grouped.

[[Page 25242]]

To the extent possible, all of the comments on a particular subject are 
grouped together. The listing of the Holtec HI-STORM 100 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 expressed concern over the number of 
cask designs being certified because there would be more problems and a 
lack of standardization and integration in the country's total waste 
system. The commenter stated that this amendment would change existing 
environmental concerns as it would add one more design, complicating 
the waste system for workers at a plant. The commenter asked how many 
designs would be certified by the NRC and how many designs could be 
used at one plant. Additional designs add to more mistakes and human 
error because each design has different fabrication criteria and 
handling procedures.
    Response: These comments are beyond the scope of this rule that is 
focused solely on whether to add a particular cask design, the Holtec 
HI-STORM 100 cask system, to the list of approved casks. Pursuant to 
the general license, each licensee must determine whether or not the 
reactor site parameters are encompassed by the cask design bases 
considered in the cask SAR and SER. Further, each general licensee must 
document this determination in accordance with 10 CFR 72.212.
    Comment A.2: One commenter stated that the tiered environmental 
impact statement (EIS) is outdated for current dry cask design and 
should be redone, particularly looking at terrorism and sabotage at an 
independent spent fuel storage installation (ISFSI).
    Response: The NRC disagrees with the comment. The environmental 
assessment (EA) and finding of no significant impact (FONSI) prepared 
as required by 10 CFR Part 51 conform to National Environmental Policy 
Act (NEPA) procedural requirements. Tiering on past EISs and EAs is a 
standard process under NEPA. As stated in the Council on Environmental 
Quality's 40 Frequently Asked Questions, the tiering process makes each 
EIS/EA of greater use and meaning to the public as the plan or program 
develops without duplication of the analysis prepared for the previous 
impact statement.
    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 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 the public health and safety.
    Comment A.3: One commenter questioned whether the NRC was including 
interim storage away from reactors in the EA, such as at a Federal or 
private storage site in Nevada or Utah. The commenter further 
questioned whether it was the NRC's intent to include transfer and 
storage at a second site in the EA. The commenter asked if the 
certification covered use at an interim site in Nevada or Utah.
    Response: The EA supports the generic use of the Holtec HI-STORM 
100 cask system under a general license. The storage could occur at any 
site that meets the definition of a general licensee under 10 CFR Part 
72. 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). The EA does not cover 
transportation from one site to another.
    Comment A.4: One commenter questioned whether the NRC claims to 
have done research on the condition of spent fuel after 20 to 50 years 
of storage at a reactor in pools and dry casks, after being unloaded 
twice and being transported across the country. The commenter stated 
that a detailed analysis of what can happen to spent fuel before it 
gets to Nevada or Utah should be conducted by the NRC. The commenter 
asked what the spent fuel will be like and what the potential 
environmental impacts will be after the fuel is unloaded and 
transported.
    Response: The NRC staff has reviewed numerous research reports 
regarding the long term condition of spent fuel in wet and dry storage. 
Additionally, the NRC has ongoing confirmatory research with spent fuel 
removed from dry storage after 10 to 20 years. Analysis of spent fuel 
has included the loads from routine shipping; and the effects, 
primarily due to vibration, were found to be negligible.
    The HI-STORM 100 MPC is a dual-purpose canister. Once loaded in the 
MPC, the fuel is not intended to be unloaded and reloaded as the 
questioner suggests. The lid welding and testing requirements and the 
structural and thermal analyses in the SAR give the NRC staff 
reasonable assurance that cask confinement and fuel integrity will be 
maintained under design basis normal, off-normal, or accident events. 
Therefore, fuel unloading should not be necessary. Regardless of 
whether unloading may be necessary, each cask user is required to 
develop detailed site-specific unloading procedures. Proper unloading 
does not cause any particular degradation to occur to the fuel.
    Comment A.5: One commenter stated that the no action alternative 
was acceptable because the NRC should not be certifying numerous 
designs. The commenter stated that other agencies such as NWTRB, EPA, 
OCRWM, and DOE should be contacted for their views on what happens to 
the whole waste system as more designs are certified.
    Response: The NRC disagrees with the comment. The NRC found no 
inherent design features that would result in significant environmental 
impacts and that the HI-STORM 100 design meets regulatory requirements. 
Therefore, there is no basis for denial of the application. The NRC 
does not limit the number or types of casks that may be certified. The 
NRC is not required to contact the agencies mentioned by the commenter 
and we have not specifically solicited their input. The commenter may 
contact these other agencies if interested in their views.
    Comment A.6: One commenter recommended finding a reference 
(reference 1 on page 3-16 of the SER) that is more recent than 1962.
    Response: The NRC disagrees with this comment. This reference 
refers to the change of the coefficient of friction from static to 
dynamic condition. The rational behind this engineering principle has 
not changed with time.
    Comment A.7: One commenter stated that the NRC should request 
simpler designs because of material interactions instead of approving 
designs with new materials that have never received long term testing 
for material interactions.
    Response: The NRC staff disagrees with this comment. The materials 
used in casks are selected upon the basis of the needed properties. 
Casks are constructed from a limited number of materials. The materials 
used in the Holtec HI-STORM design have a long history of use in the 
nuclear industry and the performance of those materials is well known.
    Comment A.8: One commenter objects to site specific changes that 
are made to generic designs.
    Response: This comment is beyond the scope of this rule that is 
focused solely on whether to place the HI-STORM 100 cask system on the 
list of approved casks. Section 72.48 permits

[[Page 25243]]

changes to the spent fuel storage cask as described in the FSAR and 
defines the conditions under which these changes may be made without 
prior NRC approval.
    Comment A.9: One commenter stated that it appeared that Holtec 
split what appears to be one generic system into two separate rules and 
asked why the system was not certified together. Systems should be 
complete when they are proposed for rulemaking. The commenter further 
stated that vendors should apply for storage and transport at the same 
time and that NRC should not allow loading until the transportation 
portion is certified.
    Response: The NRC disagrees with the comment. The HI-STAR 100 Cask 
System and HI-STORM 100 Cask System are two separate spent fuel storage 
cask systems. Each is a complete spent fuel storage cask system that 
satisfies the requirements of 10 CFR Part 72. Regarding the dual-
purpose (storage and transportation) use of a cask system or its 
components, separate certifications are required for approval of a cask 
design (or individual components such as a canister) under the 
provisions of use for 10 CFR Parts 71 and 72. There is no regulatory 
requirement that the certification be simultaneous.
    Comment A.10: One commenter asked a number of site-specific 
questions related to Private Fuel Storage's plans to use the Holtec HI-
STAR and HI-STORM cask systems at the Utah site. These issues related 
to cask handling, dry transfer, sabotage scenarios, infrastructure for 
unloading, etc. One commenter stated that they understood that Private 
Fuel Storage plans to use the HI-STAR system for storage and transport 
with the HI-STORM as a companion concrete overpack, that the metal HI-
STAR overpack would be used as a backup, and that the commenter 
objected to these plans.
    Response: The comment is beyond the scope of this rule that is 
focused solely on whether to add a particular design, the Holtec HI-
STORM 100 cask system, to the list of approved casks. The rule will 
enable licensees to use this cask system under the general license 
provisions of 10 CFR Part 72. The rule does not address site-specific 
issues related to potential users.
    Comment A.11: One commenter objected to calling the cask a multi-
purpose cask (MPC) because that stands for storage, transport, and 
disposal, and stated that the cask is not approved for these functions 
which can cause confusion when real MPCs are certified.
    Response: The NRC disagrees with the comment. The name or model 
number given to the cask design is developed by the applicant. The CoC 
for the Holtec HI-STORM 100 is intended for the interim storage of 
spent fuel. The use of MPC in a dry storage cask application or an NRC 
SER/CoC is not a certification under 10 CFR Part 71 for the transport 
of radioactive materials or an approval for disposal at a high-level 
waste repository.
    Comment A.12: One commenter stated that Holtec should not be 
allowed to approve its own suppliers and that the suppliers should be 
ASME-approved.
    Response: The NRC disagrees with the comment. NRC regulations do 
not require an ASME stamp for a cask or the use of ASME-approved 
suppliers. The design and fabrication requirements for a certified dry 
cask storage system are described in 10 CFR Part 72 and the NRC staff's 
Standard Review Plan, NUREG-1536, ``Standard Review Plan for Dry Cask 
Storage Systems'' (SRP). Applicant submittals are reviewed to the 
criteria in the SRP. Cask fabrication activities are audited by the 
licensees and inspected by the NRC staff to ensure that components are 
fabricated as designed. The CoC holder and licensee are responsible for 
verifying that fabricators are qualified. The CoC holder and licensee 
must have a Quality Assurance (QA) Program that has been approved by 
the NRC as part of the licensing or CoC issue process. This QA program 
must meet the requirements of 10 CFR 72.148 and 10 CFR 72.154 for the 
selection of fabricators. Also, the procurement documents issued to the 
fabricator must comply with 10 CFR 21.31. The licensee/CoC holder is 
required to verify that all regulations and CoC conditions applicable 
to the container are met. The NRC inspects the licensee/CoC holders and 
fabricators to verify compliance. Additionally, many storage cask 
fabricators are certified by the American Society of Mechanical 
Engineers and are N-Stamp Certificate holders.
    Comment A.13: One commenter stated that issues should not be 
resolved in telephone conferences but in public meetings with a record 
in the public document room.
    Response: The NRC disagrees with the comment. Telephone conferences 
are an important mode of communication with applicants and licensees 
and enable the NRC staff to conduct its official business efficiently. 
If, in these telephone conferences, the NRC staff receives information 
that would form the basis for its regulatory decision, that information 
is documented and made available for public inspection under 10 CFR 
Parts 2 and 9.
    Comment A.14: One commenter stated that all details of the design 
should be finalized and open for public comment.
    Response: The NRC disagrees that all design details need to be 
finalized and open for public comment before a design is approved. The 
NRC staff focuses its review on those design details that are 
significant with respect to the health and safety of the public and/or 
are required to make a regulatory finding. Design details that are 
pertinent to the NRC staff's findings are finalized and made available 
for public inspection and comment under 10 CFR Parts 2 and 9.

B. Radiation Protection

    Comment B.1: One commenter objected to the use of less shielding 
for the 100-ton transfer cask and allowing the utilities to perform a 
cost-benefit analysis to justify the use of the 100-ton transfer cask 
at the expense of the worker. The workers should receive the minimum 
achievable dose and not the maximum allowable dose. The NRC should not 
allow the use of the 100-ton transfer cask because the dose is 3 times 
higher and workers should not be treated as guinea pigs. The commenter 
stated that the utilities should be required to use the 125-ton 
transfer cask which is safer and modify their facilities to accommodate 
the transfer cask or choose a cask that works for their specific site 
limitations because the utilities shouldn't limit the shielding for 
workers.
    Response: NRC disagrees with this comment. Each cask user will 
operate the HI-STORM 100 under a 10 CFR Part 20 radiological protection 
program. ALARA means making every reasonable effort to maintain 
exposures to radiation as far below the dose limits while taking in 
account the state of technology, the economics of improvements in 
relation to the state of technology, and the economics of improvements 
in relation to benefits to the public health and safety. As stated in 
Section 2.0.3 of the SAR, the general licensee should utilize the 125-
ton transfer cask provided it is capable of using it. However, 
licensees not capable of using the more shielded design may employ 
ALARA considerations when evaluating whether to modify its plant or use 
the 100-ton transfer cask. The NRC found this acceptable as discussed 
in Section 10.2 of the SER.
    Comment B.2: One commenter asked why the specific dose rate 
criteria for the HI-TRAC was not given and indicated that the criteria 
should be included.

[[Page 25244]]

    Response: The applicant did not provide explicit dose rate values 
as design criteria for the transfer cask designs, but stated that the 
radiological requirements of 10 CFR Parts 72 and 20 as the overall 
shielding design objectives for the cask system. The NRC found this 
acceptable. The TSs in Appendix A of the CoC specify dose rate limits 
for the transfer casks that are based on the applicant's shielding 
calculations.
    Comment B.3: One commenter questioned the bounding analysis for 
cobalt impurities, asked how much cobalt is really in the fuel, and if 
the quantity had been tested and verified for the real thing.
    Response: The applicant's analysis of cobalt impurities is 
discussed in Section 5.2.1 of the SER. The applicant showed that the 
cobalt impurity values that are assumed in its shielding analyses were 
appropriate based on industry data and analysis of post-irradiation 
cooling of older fuel. The NRC found this acceptable. The cask user is 
not required to measure the actual quantity of cobalt in its spent 
fuel. The cask user will operate the cask under a 10 CFR Part 20 
radiological protection program and verify that the cask system meets 
the dose rate limits specified in the TSs.
    Comment B.4: One commenter asked why backscattering was not 
considered for all cask designs.
    Response: This comment is beyond the scope of this rule that is 
focused solely on whether to add a particular cask design, the Holtec 
HI-STORM 100 cask system, to the list of approved casks. Note that 
backscatter was considered for the Holtec HI-STORM 100 cask system.
    Comment B.5: One commenter asked what are the various array 
configurations allowed and what are the differences between them. The 
commenter asked if the cask array is limited to two rows and for the 
applicable NRC criteria.
    Response: The use of the HI-STORM design is not limited to two 
rows. The NRC requires the applicant to perform off-site dose 
calculations from a typical ISFSI array to demonstrate that radiation 
shielding features are sufficient to meet the radiological requirements 
of 10 CFR Parts 72.104 and 72.106. As discussed in Section 5.3.1 of the 
SER, the applicant used a two-row cask array model as part of its 
methodology to estimate off-site dose rates. The values obtained by 
this method can be applied to dose rate calculations for typical cask 
arrays that may consist of multiple rows. NRC found the dose estimates 
to be acceptable. Each general licensee will identify an ISFSI 
configuration and perform a site-specific dose evaluation to 
demonstrate compliance with Part 72 radiological requirements.
    Comment B.6: One commenter asked why the dose rate for the bottom 
of the MPC-68 was higher than for the MPC-24 when the dose rates at the 
side and top were higher for the MPC-24. The commenter stated that the 
trunnion doses showed that extreme care needs to be taken in those 
areas and that the bottom doses are really high and don't get enough 
attention.
    Response: The applicant appropriately assumed design basis fuel 
loadings for each canister and estimated dose rates at various 
locations. The NRC notes that dose rates at the bottom of the canister 
depend on several factors such as the fuel hardware characteristics, 
irradiation and cooling history, and the relative position of each fuel 
type within the cask system. The NRC found that the applicant 
appropriately addressed these and other factors, and that the 
calculated dose rates at the bottom and at the trunnions of the 
transfer cask were acceptable. In addition, each cask user will operate 
the HI-STORM 100 under a 10 CFR Part 20 radiological protection program 
and monitor dose rates during loading and unloading.
    Comment B.7: One commenter asked what the dose for the 2x5 cask 
array was at 100 meters.
    Response: Figure 5.1.3 of the SAR indicates that the dose rate for 
a 2x5 array at 100 meters is approximately 600 to 700 mrem/yr assuming 
a design basis fuel loading and 100 percent occupancy. Each general 
licensee will identify an ISFSI configuration and perform a site-
specific dose evaluation, based partly on site-specific 
characteristics, to demonstrate compliance with Part 72 radiological 
requirements.
    Comment B.8: One commenter asked why other cask designs do not 
account for approximate atmospheric conditions. The commenter also 
asked the conditions of weather or location for which the air density 
decreases.
    Response: Atmospheric density changes daily. The measure of the 
density is provided by local weather forecasters through the barometric 
pressure. When a high pressure front passes an area, the air density is 
greater than when a low pressure weather front passes the same 
location.
    The comment concerning other cask designs is beyond the scope of 
this rule that is focused solely on whether to place the Holtec HI-
STORM 100 cask system on the list of approved casks. For the HI-STORM 
100, each general licensee should consider atmospheric conditions 
relevant to its ISFSI as indicated in Section 5.4.2 of the SER.
    Comment B.9: One commenter asked how much the releases from dry 
storage add to the effluent from a reactor site and the duration of a 
release, and what happens to the cask and fuel during the release.
    Response: Specific effluent releases from reactors operated by 
general licensees are beyond the scope of this rule. However, NRC does 
not expect any effluent release from the HI-STORM 100 under credible 
conditions. Design basis public exposures from direct radiation and 
hypothetical releases are discussed in SER Sections 10.4 and 10.5.
    Comment B.10: One commenter approved of the condition in Appendix B 
of the CoC regarding the evaluation of engineering features (e.g. berm) 
that are used for radiological protection by the user.
    Response: No response is necessary.
    Comment B.11: One commenter stated that average surface dose rates 
in TS 3.2.1 for transfer cask dose rates should not be used, that the 
highest value should be used, and the limit should not be exceeded. The 
commenter also asked why the side dose rates are measured along the 
middle of the flat surface section of the neutron shield rather than on 
the radial steel fins where dose rates are assumed by the commenter to 
be higher.
    Response: The NRC disagrees with the comment. The specification of 
surface average dose rates and the measuring locations on the side of 
the neutron shield are consistent with health physics methods that are 
used to characterize radiation fields around a cask. The measuring 
locations are also consistent with the dose rate calculations presented 
in the applicant's shielding analysis. The cask user will operate the 
HI-STORM 100 under a 10 CFR Part 20 radiological protection program. 
NRC has reasonable assurance that the general licensee's radiological 
protection and ALARA program will detect and mitigate exposures from 
the radiation fields that are expected during operation of the HI-STORM 
100 system.
    Comment B.12: One commenter asked why the dose rate for the bottom 
of the transfer cask is not provided in TS 3.2.1 and what is that dose 
rate.
    Response: Dose rate limits for the bottom of the transfer casks are 
not needed because they would not provide a significant benefit in 
ensuring compliance with regulatory limits on occupational dose and 
dose to the public. The dose limits at the top and side of the transfer 
casks are adequate to help ensure that the cask system is

[[Page 25245]]

safely operated in compliance with 10 CFR Part 20 and Part 72. 
Calculated dose rates at the bottom of the transfer casks are reported 
in Sections 5.1 and 5.4 of the SAR.
    Comment B.13: One commenter recommended that Section 5.1.2 of the 
SER be revised to clarify that overpack surface dose rates are design 
objectives and are shown to be met by analysis, and that the TSs are 
equal to or more conservative than the design objectives.
    Response: The NRC disagrees with this comment. The NRC staff does 
agree that the vent dose rates calculated by the applicant are 
significantly less than the applicant's proposed design criteria. 
However, the differences between the calculated vent dose rates and the 
proposed design criteria are not relevant to the bases and findings in 
the SER. The TSs in Appendix A of the CoC specify vent dose rate limits 
for the overpack that are based on the applicant's shielding 
calculations. Therefore, a revision to the SER to reflect the dose rate 
difference is not necessary.
    Comment B.14: One commenter recommended that Section 5.4.11 and 
Table 5.4-1 of the SER be clarified to indicate that the dose rates are 
not peak or maximum values.
    Response: The NRC agrees with the comment. The SER has been 
clarified to state the vent dose rates are average over the area of the 
vent opening. A footnote has been added to Table 5.4.1 to clarify 
values are average over surface detector areas.
    Comment B.15: One commenter recommended that Section 10.5.1 of the 
SER be revised to indicate that the maximum MPC leak rate is utilized 
in the calculations.
    Response: The NRC agrees with the comment. The SER text has been 
revised accordingly.
    Comment B.16: One commenter indicated there was an inconsistency 
between the accident condition whole body and thyroid dose values 
referenced in Chapter 11 of the draft SER and the dose values 
calculated in Section 7 of the applicant's SAR.
    Response: The NRC agrees with the comment. The SER has been revised 
to indicate the correct whole body and thyroid dose values calculated 
by the applicant. The accident condition whole body total effective 
dose equivalent (TEDE) is 44.1 mrem and the thyroid dose is 4.1 mrem.
    Comment B.17: One commenter objected to the use of a 30-day 
duration of a radiological release during an accident. The commenter 
noted that this assumption is stated in Interim Staff Guidance 5 but 
that it is not justified in the guidance or any accompanying report. 
The commenter pointed out that NRC regulations for ISFSIs do not 
require offsite emergency planning, or planning for the ingestion 
pathway zone, and therefore, there is no basis for assuming that 
something happens within 30 days to stop the release.
    Response: The NRC disagrees with the comment. As indicated in ISG-
5, Rev.1, the 30-day assumption is consistent with the time period that 
is used to demonstrate compliance with radiological dose requirements 
associated with reactor facilities that operate under 10 CFR Part 50. 
The applicant specified corrective actions for each accident in Chapter 
11 of the SAR. NRC believes that these corrective actions can be 
reasonably achieved within 30 days. Although NRC does not expect 
effluent release from the HI-STORM 100 under credible accident 
scenarios, the 30-day assumption in the analysis is acceptable because 
the NRC staff has reasonable assurance that in the 30-day timeframe 
adequate protective measures can and will be taken for the public in 
the event of a radiological emergency. These protective measures 
include implementation of the general licensee's Part 50 emergency 
plan, evacuation of the surrounding public, and mitigation of 
radiological ingestion pathways.
    Comment B.18: One commenter objected to the assumption that a 
person at the fence post (500 meters) would be exposed for only 2000 
hours/year which is the number of working hours in a year. The 
commenter stated that 8,760 hours/year should be used because a 
licensee can not control who would be in the area outside the fence or 
how long they would be there. For conservatism, the applicant should 
have assumed that people, such as mothers with pre-school aged 
children, the elderly, ranchers, and farmers are present at the fence 
post day-long and year-round.
    Response: The NRC agrees that 8,760 hours/year should be used and 
notes that Section 7.2.9 of the HI-STORM SAR explicitly states that: 
``The individual at the site boundary is exposed for 8,760 hours 
[7.0.2].'' The NRC staff's independent calculations confirmed Holtec's 
calculated results, as stated in the NRC staff's SER. In addition, 
Section 7.2.9 also assumed in its calculations that: ``The distance 
from the cask to the site boundary is 100 meters.'' With respect to 
hypothetical individual exposed at the site boundary, the methods used 
in the dosage calculations cover children, the elderly, ranchers, 
farmers, etc. The overall public dose limit is protective of all 
individuals because the variation of sensitivity with age and gender 
was accounted for in the selection of the lifetime risk limit, from 
which the annual public dose limit was derived.
    The NRC continues to believe that the existing regulations and 
approved methodologies adequately address public health and safety. The 
issue of dose rates to children was addressed in the Federal Register 
on May 21, 1991 (56 FR 23387).
    Comment B.19: One commenter stated that the dose due to direct 
gamma and ingestion of radionuclides should be considered in the dose 
calculation because to ignore these pathways underestimates the dose. 
The commenter further objected to the NRC staff stating (in the Holtec 
HI-STAR 100 final rule) that these pathways would be addressed in the 
general licensee's site-specific review. The commenter stated that 
there is no regulatory requirement for these actions to be taken by the 
general licensee. The commenter stated that it is misleading for the 
applicant to do a calculation that provides a reassuring result, based 
on assumptions that have nothing to do with the real requirements of 
the regulations because licensees tend to rely heavily on the generic 
analyses that have been performed by cask manufacturers.
    Response: The NRC disagrees with the comment. Although the NRC does 
not expect effluent release from the HI-STORM 100 under credible 
conditions, the applicant's method used to determine design basis dose 
rates from a hypothetical release are adequate to demonstrate that the 
confinement features are sufficient to meet the radiological 
requirements of 10 CFR 72.106. The NRC staff believes the methods 
applied by the applicant conservatively bound hypothetical dose rates 
to the general public. Further, 10 CFR 72.212(b)(6) requires the 
general licensee to review its reactor emergency plan and radiation 
protection program to determine its effectiveness and make changes if 
necessary when using a cask listed in 10 CFR Part 72, Subpart L.
    Comment B.20: One commenter stated that the thyroid and whole body 
doses should consider chlorine-36 (Cl-36) because it will be present in 
the irradiated fuel and will significantly contribute to the dose. The 
commenter points out that the Department of Energy acknowledges that 
Cl-36 is one of the significant radionuclides in Appendix A, of the 
Yucca Mountain Draft EIS.
    Response: The NRC disagrees with the comment. The NRC staff's 
independent analysis of the thyroid and whole body dose was based on 
independent

[[Page 25246]]

calculations using the ORIGEN computer code, as referenced by the 
commenter. The calculated contribution of the chlorine gas was below 
the truncation limit used in the calculation. Cl-36 has an 
inconsequential contribution on the total dose to an individual.

C. Accident Analysis

    Comment C.1: One commenter asked if lead could be a missile strike 
barrier from a tornado or from current weapons. The commenter asked if 
missiles could penetrate the transfer cask and canister inside, and 
when the missile strike is assumed to occur (i.e. when a loaded 
transfer cask is on top of the overpack.) The commenter stated that 
this needs to be updated and evaluated.
    Response: The lead backed outer shell of HI-TRAC has been evaluated 
for the required tornado missile strike. The analysis shows that there 
is no penetration consequence that would lead to a radiological 
release. The threat of missiles from weapons is beyond the scope of 
this rule.
    Comment C.2: One commenter expressed concern that the transfer cask 
is a real target on top of the storage cask and asked if it had been 
fully evaluated for terrorism and sabotage, particularly when it was on 
top of the storage cask. The commenter asked if the overpack was put in 
place while on the pad; the commenter felt that this would be a target 
for terrorists. The commenter asked if the transfer cask, with inner 
cannister inside, could be knocked off by a terrorist blast and fall, 
crash, or roll into other casks or be upended so that the fuel is 
upside down.
    Response: The performance of the transfer cask in a sabotage or 
terrorist event was not evaluated. The threat of terrorism or sabotage 
is beyond the scope of this rule. See also the response to C.8.
    Comment C.3: One commenter asked if the seismic event was based on 
the actual pad analysis and not the reactor building seismic analysis 
because the conditions between the reactor building and pad location 
could significantly differ.
    Response: The storage pad is a site-specific issue and is beyond 
the scope of this cask design rule. Under 10 CFR 72.212, the cask 
operators are required to perform written evaluations to ensure that 
storage pads have been designed to adequately support the stored casks. 
The licensee using a particular cask design has the responsibility 
under the general license to evaluate the match between reactor site 
parameters and the range of site conditions (i.e. the envelope) 
reviewed by the NRC for an approved cask.
    Comment C.4: One commenter asked how a full cask array would behave 
in a seismic event. The commenter asked what buildings or equipment are 
allowed on the pad that could crash into the casks during a seismic 
event, such as the transfer equipment. The commenter asked if a crack 
or ``push up'' of the pad could cause the cask to roll (down an incline 
or into water).
    Response: The SAR indicates that the HI-STORM 100 overpack will 
neither slide nor tip over due to a seismic event with the design-basis 
earthquake input listed in Section 3.4.2 of the SER. The use of a 
general licensed cask by a utility requires that the user ensure that 
the site is not subject to any potential accident that has not been 
analyzed for the general license. This would include any potential 
design basis earthquakes that were not enveloped by the NRC SER for the 
cask or any site conditions associated with the actual pad and cask 
locations that could affect the cask design.
    Comment C.5: One commenter asked what the design-basis earthquake 
on top of the surface pad was and where it occurred. The commenter 
questioned why the bottom surface was not evaluated because the ground 
can push up and crack or cause heaving in the concrete and how the 
condition of the bottom surface is known.
    Response: The design basis earthquake is the most severe earthquake 
that has been historically reported for a particular site and 
surrounding area, with sufficient margins for the limited accuracy, 
quantity, and period of time in which historical data have been 
accumulated. Structure, systems, and components important to safety are 
designed to withstand the effects of this earthquake without loss of 
capability to perform their safety functions. The design basis 
earthquake is described by an appropriate response spectrum anchored at 
the peak ground acceleration. The response is then amplified through 
the pad to obtain the input response spectrum at the top of the pad (or 
at the bottom of the cask) for cask seismic evaluation. Soil and 
storage pad interaction is a site-specific issue that will be addresses 
in the cask user's 10 CFR 72.212 evaluation and is beyond the scope of 
this rule.
    Comment C.6: One commenter asked what happens if the pad is cracked 
and heaving up as the cask is tipping over because a tornado or seismic 
event will likely affect both the pad and the casks.
    Response: The NRC does not consider the scenario described by the 
commenter to be credible. The evaluation in Section 3 of the SAR shows 
that tipover will not occur. However, as a defense-in-depth measure, 
cask tipover is also evaluated in Section 3 of the SAR and discussed in 
Section 3.4.2 of the SER.
    Comment C.7: One commenter asked if the cask could become upside 
down in a tornado or seismic event and if it happened would the top of 
the fuel hit the underside of the MPC lid with the weight on the 
overpack lid studs.
    Response: The HI-STORM 100 overpack is evaluated for tornado, 
tornado missiles, and seismic events in Section 3 of the SAR. The 
results indicate that the cask will not tip over. Therefore, the cask 
will not become upside down.
    Comment C.8: One commenter stated that an airplane crash with its 
fuel fire should be evaluated, including crash into a full cask array, 
damage to the pad, and a fuel and airplane explosion after the crash. 
The commenter stated that an anti-missile device with an incendiary 
device and a truck bomb should be analyzed for the cask transfer 
facility (CTF).
    Response: The NRC disagrees with the comment. Before using the HI-
STORM 100 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 casks as required by 10 CFR 
72.212(b)(3). The licensee's site evaluation should consider the 
effects of nearby transportation and military activities.
    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). 
NRC regulations in 10 CFR Part 72 establish physical protection 
requirements for an ISFSI located within the owner-controlled area of a 
licensed power reactor site. Spent fuel in the ISFSI is required to be 
protected against radiological sabotage using provisions and 
requirements as specified in 10 CFR 72.212(b)(5). Further, specific 
performance criteria are specified in 10 CFR Part 73. Each utility 
licensed to have an ISFSI at its reactor site is required to develop 
physical protection plans and install systems that provide high 
assurance against unauthorized activities that could constitute an 
unreasonable risk to public health and safety.
    The physical protection systems at an ISFSI and its associated 
reactor are similar in design features to ensure the detection and 
assessment of unauthorized activities. Alarm

[[Page 25247]]

annunciations at the general license ISFSI are monitored by the alarm 
stations at the reactor site. Response to intrusion alarms is required. 
Each ISFSI is periodically inspected by NRC. Also, the licensee 
conducts periodic patrols and surveillances to ensure that the physical 
protection systems are operating within their design limits. The ISFSI 
licensee is responsible for protecting spent fuel in the casks from 
sabotage not the certificate holder. Comments on the existing 
regulations specifying what type of sabotage events must be considered 
are beyond the scope of this rule.
    Comment C.9: One commenter questioned why the tornado missile test 
simulated a pulse impact of a vehicle and stated that a sharp object 
such as a metal pole or other items that might be in the vicinity of a 
real pad would do more penetration damage.
    Response: In addition to the 4,000-pound automobile impacting at a 
126 mph velocity, the SAR also provided analyses for two more missiles 
impacting at 126 mph velocity: a 1-in diameter steel sphere and an 8-in 
diameter rigid cylinder. Results of the analyses show that the 4,000 
pound automobile produces the highest impact force on the cask because 
it has the largest mass. Based on these results, the NRC staff has 
reasonable assurance that the 4,000 pound vehicle bounds the effect of 
other credible types of tornado missiles.
    Comment C.10: One commenter stated that the 15-minute transporter 
fire is not valid. A big plane crash with its fuel should be evaluated 
as well as a sabotage missile penetration with an incendiary device.
    Response: The NRC disagrees with this comment. The basis for the 
4.8-minute fire (not a 15 minute fire, see response to comment C.18) is 
associated with the time it would take to burn 50 gallons of fuel, 
presumably carried by the transporter. The CoC, Appendix B, Section 
3.4, states that ``the on-site transporter fuel tank will contain no 
more than 50 gallons of diesel fuel while handling a loaded OVERPACK or 
TRANSFER CASK.'' Other modes of transport causing the fire (e.g., 
airplanes, trains, delivery trucks or missiles) are not considered as 
plausible and are beyond the scope of this rule. Before using the HI-
STORM 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 that 
analyzed in the SAR and that any cask handling equipment used to move 
the HI-STORM cask to the pad is limited to 50 gallons of fuel (refer to 
CoC, Appendix B, Section 3.4--Site Specific Parameters and Analyses).
    Comment C.11: One commenter asked why there were no calculations 
for the bottom plate, overpack lid, etc. in a fire because the 
temperatures of these plates were important to know and could affect 
the pad or fire fighting equipment.
    Response: The applicant did calculate the temperatures for the 
bottom plate and the overpack lid. However, these temperatures were not 
reported in the SAR. Not all calculated temperatures are reported in 
the SAR. With respect to the impact of fire on the pad or fire fighting 
equipment, a postulated 50 gallon fuel source would have minimal impact 
on those components. The heat generated by the pool of fuel is directed 
upward where the fuel is in a gaseous state. The limiting temperatures 
will occur above the surface of the concrete pad. Because the fuel has 
to vaporize in order to burn, the liquid fuel on the concrete will have 
minimal impact on the bottom plate of the overpack lid (in a liquid 
state, the fuel is cool). The duration of the fire is less than 4 
minutes. The impact on the fire fighting equipment would be minimal, if 
any. Table 4-3 of the SER was modified to indicate that the 
temperatures were not reported.
    Comment C.12: One commenter asked how the 45,000 MWD/MTU for 5 
years related to the sabotage and terrorist evaluation for radiation 
disposal and stated that the evaluation is outdated.
    Response: The comment on the sabotage report is beyond the scope of 
this rule. See the discussion in the response to C.8.
    Comment C.13: One commenter asked if the water jacket could be 
pierced with an anti-missile gun or if a terrorist could shoot the 
jacket full of holes, and what are the consequences if these events did 
occur.
    Response: The specific threat of an anti-missile gun or other small 
arms against the HI-STORM 100 is beyond the scope of this rule. 
However, the resultant dose rate for an assumed complete loss of the 
water jacket is addressed in Section 5.1.2 of the SAR. The analysis 
indicates that the off-site dose at 100 meters will be below the 5 rem 
accident limit in 10 CFR 72.106.
    Comment C.14: One commenter asked why a burial under a landslide 
during a seismic event is not considered.
    Response: Burying a cask due to seismic event, landside, or tornado 
is considered a very unlikely event. Considering the unlikeliness of 
the event coupled with the casks being able to withstand these events 
make burying and any adverse consequence in the opinion of the NRC not 
credible.
    Comment C.15: One commenter asked why a vertical drop of a loaded 
transfer cask is not considered a credible accident, particularly as it 
is perched on top of the concrete overpack to load.
    Response: A vertical drop of a transfer cask is not considered 
credible because vertical lifting of a loaded transfer cask must be 
performed with structures and components designed to prevent a cask 
drop. The criteria for those structures and components are specified in 
Section 3.5 of Appendix B to CoC No. 1014. The restrictions on vertical 
lifting are specified in Section 5.5 of the TSs (Appendix A to the 
CoC).
    Comment C.16: One commenter stated that defense-in-depth is needed 
for sabotage events which could cause a tipover.
    Response: The NRC disagrees with this comment. Sabotage events are 
beyond the scope of this rule. They are considered in Part 73. 
Furthermore, the SAR demonstrates that the HI-STORM 100 overpack will 
not tipover due to a design basis accident. However, as an added 
defense-in-depth measure, cask tipover is evaluated in Section 3 of the 
SAR and discussed in Section 3.4.2 of the SER.
    Comment C.17: One commenter asked why a postulated explosion from a 
truck bomb at the pad fence was not evaluated.
    Response: The specific threat of a truck bomb is beyond the scope 
of this rule. The response to C.8 addresses radiological sabotage of 
storage casks at generally licensed ISFSIs.
    Comment C.18: One commenter asked the basis for the 217-second fire 
for the overpack and the 4.8-minute fire for the transfer cask. The 
commenter also asked if the NRC assumed that nothing on the vehicle or 
in the vicinity (such as grass or trees or other structures) will burn 
and cause the fire to burn longer.
    Response: The duration of a fire burn is based on several factors. 
One factor is the rate at which the fire burns, normally categorized as 
inches of fuel burned per minute. The burn rate (inches per minute) is 
the same for both the overpack and the transfer cask because the source 
of fuel is the same (e.g., diesel fuel). The duration of the burn comes 
from the postulated depth of the pool of fuel. A conservative estimate 
of the time of burn is to assume that the spilled fuel does not extend 
beyond 1 meter of the surface of the overpack or the surface of the 
transfer cask. (In reality, the fuel will spill significantly farther 
than one meter on

[[Page 25248]]

a flat surface, just as spilling a bucket of water on the ground, and 
will not accumulate to any significant depth which creates a shorter 
fire burn time.) Because the outer diameter of the overpack and the 
outer diameter of the transfer cask are different, the postulated depth 
or height that 50 gallons of fuel is postulated to reach will differ 
for the two cases. The case with the higher column of fuel will burn 
longer. Because the surface area of the pool of fuel for the overpack 
is 1.3 times larger than for the transfer cask, the pool of fuel for 
the overpack will be lower (given the same volume of available fuel, 
e.g., 50 gallons). A lower pool of fuel will burn quicker. (Note that 
the burn rate is in inches of fuel per minute, and a smaller column of 
fuel will burn quicker than a higher column of fuel). Therefore, the 
burn time for the overpack is shorter than the burn time for the 
transfer cask.
    With respect to other flammable sources that could catch fire, 
before using the HI-STORM 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(b)(3). Included in this evaluation is the verification that 
the cask handling equipment used to move the concrete cask to the pad 
is limited to 50 gallons of fuel (refer to CoC, Appendix B, Section 
3.4.5) and that the assumptions used in the SAR bound the consequences 
for the proposed site. Additional assessments would have to be 
performed if other sources are identified that could result in a more 
limiting fire.
    Comment C.19: One commenter objected to the use of the leakage rate 
used by Holtec because it is based on an analysis of a transportation 
cask rather than a storage cask, for which the NRC and industry have 
different design and testing requirements. The commenter noted that the 
small assumed leakage rate and calculation methodology in NUREG/CR-6487 
are based on ANSI standard N14.5 for transportation casks. ANSI N14.5 
assumes that casks will be leak-tested periodically before shipment and 
after maintenance and repair. The commenter pointed out that some 
ISFSIs have no design provisions for testing helium leakage during 
storage and no provisions for repairing and maintaining casks and 
testing for leakage after repair and maintenance. Therefore, it is 
inappropriate to assume that these storage casks will have the same 
small leakage rate as transportation casks for which leakage potential 
is designed and planned to be monitored. The commenter stated that 
neither Holtec nor the NRC has any basis for relying on NUREG-1617 to 
assume a small leakage rate in a storage cask breach.
    Response: The NRC disagrees with the comment. The ANSI N14.5 
standard was developed to determine allowable leak rates for shipping 
packages that employ mechanical seals, which typically undergo 
repetitive use. Periodic testing is prescribed for the mechanical seal 
to ensure it has not degraded from repetitive use and/or seal 
maintenance. The analytic technique in ANSI N14.5 that is used to 
determine a leak rate across an assumed leak path is valid for 
determining an assumed leak rate across the confinement boundary of a 
welded canister. An off-site dose can be subsequently calculated using 
standard atmospheric dispersion principles and assuming a partial 
release of the cask constituents at the calculated leak rate. The 
welded closure is leak-tested to a sensitivity equal to the calculated 
leak rate to ensure integrity of the confinement system before storage 
operations. Periodic testing of the confinement boundary is not 
applicable because the welded confinement boundary is designed to 
remain intact during normal, off-normal, and accident conditions for 
the lifetime of the canister.
    Comment C.20: One commenter stated that the methodology used in 
NUREG/CR-6487 may not apply for accidents that exceed the design basis 
accident. The allowed leak hole size can easily be exceeded in 
accidents involving sabotage such as an impact with a MILAN or TOW-2 
hand held anti-tank device, a jet engine, or military ordnance.
    Response: The NRC disagrees with the comment. Consideration of 
accidents that exceed design basis is not required by 10 CFR Part 72 
and is beyond the scope of the NRC staff's review. The threat of 
accidents involving sabotage is beyond the scope of this rule. Sabotage 
issues are covered by 10 CFR Part 73.
    Comment C.21: One commenter stated that Holtec should consider a 
300 gallon fire or a 6,000 gallon fire. The commenter stated that these 
are credible accidents at an ISFSI and should be considered. A heavy 
haul tractor carries 300 gallons of fuel and is likely to be used at 
ISFSIs. Locomotives that carry casks to ISFSIs may carry 6,000 gallons 
of diesel fuel.
    Response: The NRC disagrees with the comment. The analysis need 
only address the maximum permissible source of fuel at the storage site 
near the HI-STORM 100 system (10 CFR Part 72). Section 3.4 of Appendix 
B to the CoC limits the source of fuel near the HI-STORM 100 system to 
50 gallons. Licensees are required to verify that all conditions of the 
CoC are met.
    Comment C.22: One commenter stated that Holtec's fire analysis is 
deficient because the fire calculations assume that the fire takes 
place outside the concrete storage cask and does not consider the 
possibility of a fuel fire being drawn into the intake vent of the HI-
STORM 100 cask.
    Response: The NRC staff disagrees with the comment. The purpose of 
the fire analysis is to assess the consequences of a postulated fire on 
the HI-STORM system. The elements of interest are the impact of the 
fire on the peak clad temperature and the impact of the fire on the 
system materials. A 50-gallon fuel supply will have a very short burn 
duration. Applying the conservative assumptions of 10 CFR Part 71, a 
50-gallon fuel supply would theoretically result in a pool size of 0.54 
inches if limited to a one-meter spread around the overpack. The burn 
duration of the fuel in this configuration is 3.6 minutes. This burn 
duration will have insignificant impact on the peak clad temperature. 
The heat capacity of the system is too great to have an appreciable 
feedback on the peak clad temperature for a short duration transient. 
The greatest impact of a fire will be felt on the overpack. A bounding 
analysis was performed on the overpack by imposing a maximum burn 
temperature (specified in 10 CFR Part 71) on the entire outer surface 
of the overpack. This maximizes the impact on the steel liner and the 
concrete. In a less conservative calculation, the maximum temperature 
will be limited to the lower portion of the overpack. For additional 
conservatism, the applicant increased the inside temperature of the 
overpack to 300 deg.F to account for heating of the air as it passes 
through the vents. As illustrated in SAR Figures 11.2.1--11.2.5, this 
bounding calculation illustrates that only the outer boundary of the 
concrete exceeds the temperature limit for concrete. At a depth of one 
inch into the concrete the temperature limit is not challenged. If a 
conservative assumption postulates the fire to occur inside the vent, 
similar results would occur because there is only a limited amount of 
energy (BTU) that can be deposited into the massive overpack structure. 
Exceeding the concrete temperature limit only at the concrete surface 
does not lead to a safety concern, and therefore, the SAR analysis is 
acceptable.
    Comment C.23: One commenter stated that the consequences of a hit 
by an anti-tank missile, such as the MILAN or

[[Page 25249]]

TOW-2 missile should be considered. The commenter noted that the 
regulations only require a licensee to install systems that protect 
against unauthorized entry; however, entry to a site is not necessary 
to successfully carry out sabotage using an anti-tank missile. The 
commenter stated that the NRC should place additional conditions in the 
CoC to lower the probability of a sabotage event. The commenter further 
pointed out that the NRC has been inconsistent and arbitrary in 
determining whether to treat sabotage issues as site-specific or 
generic.
    Response: The NRC disagrees with the comment. The threat of an 
anti-tank missile and other sabotage events is beyond the scope of this 
rule. Requirements on radiological sabotage are covered in 10 CFR Part 
73 and apply to both ISFSIs and spent fuel storage cask designs. 
Therefore, comments on a specific threat or mode of attack are beyond 
the scope of this Part 72 rule. See also the response to C.8 addressing 
radiological sabotage of storage casks at reactor site ISFSIs.

D. Criticality

    Comment D.1: One commenter objected to the assumption on the 
continued efficacy of the boral over a 20-year storage period because 
it has never been tested or proven.
    Response: The NRC disagrees with the comment. The NRC staff does 
not consider the loss 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 
HI-STORM 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 HI-STORM SAR describes the neutron absorber and its environment, 
and evaluated boron depletion due to neutron absorption. Section 
9.1.5.3 of the SAR describes the testing procedures for the neutron 
absorber material. The neutron absorber material 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. This material is not unique and is commonly used in other 
spent fuel storage and transportation applications.
    Comment D.2: One commenter asked if Boral had ever been used in any 
dry storage casks before and if it had, how long and had it been 
tested. The commenter asked if this was an experiment with a new 
application. The commenter further asked what proof was available to 
show the continued efficacy of a boral panel. The commenter asked what 
other fuel storage and transport applications utilized Boral and stated 
that it should be documented in the SER.
    Response: As described in SAR section 1.2.1.3.1, Boral has been 
used in environments comparable to those in spent fuel storage casks 
since the 1950s, and in spent fuel shipping casks for Canadian spent 
fuel in the 1960s. In the United States, Boral has been used in 
numerous other spent fuel transportation casks since the early 1980's 
and in storage casks since the early 1990's. Some of the casks that use 
Boral are the NAC-I28 S/T, NAC-S/T, the NUHOMS-24P, NUHOMS MP-187, and 
BMI-1. The NRC disagrees that the HI-STORM SER should include a list of 
other casks that use Boral. Information on other spent fuel casks and 
Boral is publicly available. The response to comment D.1 discusses the 
efficacy of Boral and why testing other than initial fabrication 
testing is not necessary.
    Comment D.3: One commenter stated that a test should be conducted 
to verify the presence and uniformity of the neutron absorber in 
fabrication.
    Response: The presence and uniformity of the neutron absorber is 
verified as described in Section 9.1.5.3 of the SAR. The neutron 
absorber material 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.
    Comment D.4: One commenter asked if water injection in unloading 
reflood could result in large amounts of steam generation and two-phase 
flow conditions inside the MPC cavity causing over pressurization of 
the confinement boundary and a potential criticality.
    Response: As stated in SAR section 6.4.2.1, the HI-STORM system was 
evaluated with various water densities inside the cask. The cask met 
the design criterion of keff less than or equal to 0.95 for 
all credible flooding conditions. The cask is most reactive when filled 
with full density water. As can be seen in SAR Table 6.4.1, the cask 
reactivity decreases when filled with low density water (i.e., steam).
    In addition, Section 4.5.1.1.6 describes the cask cooldown and 
reflood analysis during fuel unloading operation. This section of the 
SAR states that before reflooding the cask with water, the helium 
inside the MPC is cooled to below 200 deg.F which is below the boiling 
point of water. The procedures are outlined in Section 8.3.1 of the SAR 
with reference to TS 3.1.3. These procedures limit steam generation and 
two-phase-flow interactions with the fuel to acceptable levels, thereby 
preventing over pressurization of the MPC.

E. Design

    Comment E.1: One commenter asked if there are three MPCs that are 
NRC-certified for storage and transfer because the SER states that they 
are evaluated and approved.
    Response: As stated in Condition 1.b. of the CoC and in Section 1.1 
of the SER, there are three types of MPCs that can be used in the HI-
STORM 100 Cask System: The MPC-24, the MPC-68, and the MPC-68F. The 
MPC-24 holds up to 24 PWR fuel assemblies that must be intact. The MPC-
68 holds up to 68 BWR fuel assemblies that may be intact or damaged 
(i.e., with known or suspected cladding defects greater than hairline 
cracks or pinholes). The MPC-68F holds up to 68 BWR fuel assemblies 
that may be intact, damaged, or in the form of fuel debris (i.e., with 
known or suspected defects such as ruptured fuel rods, severed fuel 
rods, and loose fuel pellets). All three MPCs have the same external 
dimensions. Section 1.2.1.1 and Table 1.2.1 of the SAR has been revised 
to clarify that there are three types of MPCs.
    Comment E.2: One commenter asked how and to what the trunnion is 
attached, and what it is made of.
    Response: The trunnions are attached by welds to the inner and 
outer shell and to the HI-TRAC top flange. The trunnions are fabricated 
of SB-637-N07718 steel and SA-350-LF3 steel.
    Comment E.3: One commenter stated that the concrete for the 
overpack should be reinforced and asked why it is not reinforced.
    Response: The NRC disagrees with this comment. The main function of 
the concrete encased between the steel shells in the HI-STORM 100 
overpack is shielding. The structural strength of the HI-STORM 100 
overpack is provided by the inner and outer carbon steel shells. The 
concrete, on the other hand, will provide an added benefit to the HI-
STORM 100 overpack because it will increase the stiffness and weight of 
the overpack to resist external forces due to seismic, tornado, and 
tornado missiles.
    Comment E.4: One commenter asked if the pedestal could shift in 
movement and touch the liner or if it could corrode to the carbon steel 
liner or baseplate. The commenter also asked what the

[[Page 25250]]

baseplate was made of and if a ceramic baseplate should be used.
    Response: The pedestal consists of concrete, 17 inches thick, 
encased in a steel shell. This shell is welded to the steel overpack 
baseplate, and the weld is examined according to the ASME Code Section 
V. Stresses in the pedestal have safety factors exceeding 16. The 
pedestal will not shift. The exterior and interior surfaces of the 
overpack are coated with an epoxy paint to prevent corrosion. The 
overpack baseplate is made of carbon steel that meets the design 
criteria.
    Comment E.5: One commenter stated that jamming of parts could be a 
problem because unjamming the parts could cause damage (during both 
loading and unloading). The commenter further asked if the cask had 
been tested for jamming and what the situation would be after 20 to 40 
years in storage.
    Response: Stainless steel shims, depicted in Detail T of drawing 
1495, sheet 5, prevent the MPC from contacting the overpack interior 
and preclude the paint from being scraped during the operational steps. 
The drop accident analyses cause stresses which significantly bound the 
stresses that could occur during normal handling operations. Therefore, 
damage to the MPC during loading and unloading into the overpack is not 
credible.
    The calculation in the SAR demonstrated that there will be no 
jamming of the MPC in the overpack under the most severe stack-up of 
tolerances. The cask has not been tested for jamming; however, a dry 
run of all operational steps is required before use of the system.
    The license life of all overpack and MPC components is 20 years. 
The applicant engineered the overpack, HI-TRAC, and MPC for 40 years of 
design life. More detailed information regarding the service life of 
the overpack, HI-TRAC, and MPC can be found in Sections 3.4.11 and 
3.4.12 of the SAR.
    Comment E.6: One commenter stated that the clearances were not 
adequate. The commenter asked if the wet fuel would be inserted into 
the overpack in the same way as the dry run and what happens if the 
crane does not have the MPC completely vertical when inserting it in 
the overpack or if the HI-TRAC or pad is not level.
    Response: There is no adverse tolerance stack-up that would prevent 
the insertion of the MPC into the overpack. Additionally, the dry run 
will verify that the MPC can be inserted into the HI-TRAC and overpack. 
All cell plates of the MPC are constructed of stainless steel that is 
not effected by immersion in water; therefore, the tolerances for the 
dry run would not change and the wet fuel will go in the same way as in 
the dry run.
    Comment E.7: One commenter is concerned that the manufacturer's 
tolerances are not clear if fabrication is the minimum margin of safety 
or minimum clearance allowed.
    Response: The most severe ``stack-up'' of manufacturer's tolerances 
provides sufficient clearance for insertion of the MPC into the HI-
TRAC. The minimum clearance allowed is thus met. The cask could be 
manufactured to the minimum allowed clearances, but this would not 
reduce the minimum margin of safety.
    Comment E.8: One commenter asked if there would be a problem if the 
radial clearance of the HI-TRAC MPC is at a maximum and the radial 
clearance of the MPC overpack is at the minimum allowed. The commenter 
asked how much leeway is allowed in fabrication in both of these radial 
clearance measurements.
    Response: No operational problem exists if the radial clearance of 
HI-TRAC/MPC is at a maximum tolerance ``stack-up'' and the radial 
clearance of the MPC overpack is at the minimum tolerance ``stack-up.'' 
These tolerances have been evaluated for all manufacturer's design 
criteria requirements and for all design temperatures. The largest 
allowable radial dimension of the HI-TRAC is greater than the smallest 
allowable radial dimension of the overpack. Fabrication requirements, 
including tolerances, are stated on the drawings. These tolerances 
provide sufficient clearance for operations.
    Comment E.9: One commenter expressed concern over the \13/16\-inch 
difference in the maximum MPC diameter and minimum overpack internal 
diameter because it was a minuscule amount for fabrication. The 
commenter also asked what was meant by average radial clearance of 
about 0.4 inches and stated that it was not a lot of clearance.
    Response: The \13/16\-inches is the minimum clearance accounting 
for tolerances between the MPC diameter and the channels/shims that are 
attached to inner shell of overpack. The channels/shims provide 
guidance for MPC insertion, position MPC within the overpack, and allow 
the cooling air flow to circulate through the overpack. The minimum 
clearance between the MPC and overpack inner shell is approximately 5 
inches without the channels. Both the clearance between the MPC and 
channels/shims and between the MPC and overpack inner shell are 
considered to be acceptable. The SER has been changed to clarify that 
\13/16\-inches is the clearance between the maximum MPC diameter and 
the channels/shims that are attached to inner shell of overpack rather 
than between the MPC and overpack inner shell. The average radial 
clearance is diametral clearance divided by two.
    Comment E.10: One commenter asked what the computed decrease (page 
3-9 of the SER) was related to. The commenter expressed concern that 
these were very small calculation amounts (0.11 inches) to depend on 
computer accuracy.
    Response: The computed decrease of 0.11 inches is the calculated 
maximum decrease in the inner diameter of the overpack shell due to a 
tipover accident. The 0.11 inches decrease in the inner diameter of the 
overpack shell is not computed by computer simulation. Rather, it is 
computed by using a standard text book equation for deformation 
calculation. The deformation due to tipover is expected to be small. 
This calculation has been evaluated by the NRC staff and found to be 
acceptable.
    Comment E.11: One commenter asked if the base under the pads would 
be the same at all sites and asked what is under the pad. The commenter 
is concerned that the pad evaluation has not received adequate 
attention because it is a crucial part of the tipover and drop 
evaluation.
    Response: Each user is required to meet the site parameters in CoC, 
Appendix B, Section 3.4 that include specific requirements for the pad. 
Site characteristics will be investigated by each cask user and 
addressed in the cask user's 10 CFR 72.212 evaluation. The pad is a 
site-specific issue. Site-specific issues are beyond the scope of this 
rule that is focused solely on whether to add the HI-STORM 100 cask 
system to the list of approved casks.
    Comment E.12: One commenter asked why there were two different 
weights for the transfer cask.
    Response: As discussed in Section 1.2.1.2.2 of the SAR, the 100-ton 
transfer cask weighs less than the 125-ton transfer cask because it has 
a reduced thickness in lead and water. The 100-ton transfer cask is 
designed for facilities not capable of handling the heavier 125-ton 
transfer cask.
    Comment E.13: One commenter asked why the bottom pool lid supported 
the weight of a loaded MPC plus water.
    Response: During lifting of the transfer cask from the fuel pool 
there is water in the MPC and the annulus. Therefore, the structural 
evaluation of the bottom pool lid of the transfer cask must consider 
all the applicable weights

[[Page 25251]]

supported by the pool lid, including the water.
    Comment E.14: One commenter stated that the cask should be up on 
something to air out the area under the cask to prevent rusting. The 
commenter questioned if the baseplate rusted if that could cause the 
cask to tipover or lean. The commenter is concerned that if the 
canister ended up leaning against the inner liner of the concrete 
shell, it would cause blockage of the venting annulus and create a 
hotspot in the concrete.
    Response: The NRC disagrees with this comment. The baseplate is 
coated with an epoxy type coating to prevent corrosion. Some rusting 
may occur at scratches in the coating. However, even a postulated 
extreme case, assuming no coating present, would not result in 
sufficient corrosion to cause an amount of leaning that would be 
significant.
    Comment E.15: One commenter asked if there is any leeway for the 
pressure in the concrete encasement between the two carbon steel outer 
and inner liners, if the concrete had room to move, and if the concrete 
could split the outer carbon steel encasing it, particularly at the 
welds.
    Response: The coefficient of thermal expansion of steel is only 
slightly greater than that of concrete, and the thermal gradient 
through the overpack wall, experienced during the extreme temperature 
criteria, was calculated to be approximately 40 deg.F. This temperature 
difference and thermal coefficient of expansions do not cause the inner 
steel to apply significant force to the concrete in the overpack. The 
outer steel shell expands somewhat more than the concrete; therefore, 
the concrete has room for expansion and exerts no force on the outer 
steel plates.
    Comment E.16: One commenter asked what a bottom pool lid is and how 
it is replaced by the heavier shielded transfer lid and if it has been 
tested.
    Response: The bottom pool lid is described in Section 1.2.1.2.2 of 
the SAR. The lids are interchanged with a transfer slide device as 
described in Section 8.1.1 of the SAR. The NRC did not require test 
results for lid changing operations. The NRC found the pool and 
transfer lid design to be acceptable for the HI-STORM 100 system.
    Comment E.17: One commenter asked if the 17.000 inches of concrete 
for the overpack baseplate was a typo and if the number of significant 
figures was correct.
    Response: The value in the SER has been revised to state 17.0 
inches that reflects the thickness assumed in the shielding analysis.
    Comment E.18: One commenter stated that the configuration 
discussion in Section 6.4.1 of the SER is not clear because the HI-STAR 
doesn't have a transfer cask.
    Response: As stated in SER section 6.4.1, the HI-STORM system has a 
transfer cask, not the HI-STAR system. The transfer cask, the HI-STORM 
overpack, and the HI-STAR overpack are constructed of different 
materials. The effectiveness of these materials to reflect neutrons 
affects the criticality safety of the system; therefore, each was 
explicitly evaluated. The other parameters affecting the criticality 
safety of the HI-STORM system, including the transfer cask, are 
identical to the HI-STAR system.
    Comment E.19: One commenter asked if the closure ring was a ring or 
a lid.
    Response: The closure ring is a ring. In the MPC, the lid and the 
closure ring are two different components.
    Comment E.20: One commenter asked how many rings are included in 
the design.
    Response: There is one closure ring included in the design.
    Comment E.21: One commenter asked why voids in the installation of 
the lead shield are only minimized instead of being disallowed 
completely, if the shield was composed of lead bricks or poured, and 
which was more prone to voids. The commenter asked if lead bricks could 
be used and then have lead poured into the cracks between the bricks, 
and how the lead shield is installed.
    Response: The HI-STORM 100 must be fabricated and tested in 
accordance with the drawings specified in the SAR and under a quality 
assurance program that meets the requirements of 10 CFR Part 72, 
Subpart G. The proper fabrication of the lead shield, including 
potential voids, will be evaluated under this quality assurance 
program. As discussed in Section 9.1.5.2 of the SAR, effectiveness of 
the lead pours are verified during fabrication by performing gamma 
scanning on all accessible surfaces of the transfer cask in the lead-
pour regions. Installation of the lead shields is discussed in Section 
9.1.5.1 of the SAR. The SAR specifies the use of poured lead and does 
not allow the installation of lead bricks.
    Comment E.22: One commenter asked what the relief valve was and 
what type of maintenance it received.
    Response: A relief valve is a mechanical device that opens when 
pressure inside a system exceeds the actuation pressure of the valve 
(pressure that will open the valve). Relief valves are common pressure 
limiting devices. Relief valves are placed on water heaters in homes to 
ensure that the water pipes in a house will not fail due to excessive 
pressure. Similarly, relief valves are attached to the radiator in a 
car to ensure that the coolant hoses do not burst from excessive 
pressurization of the engine coolant system. Maintenance of the relief 
valves are discussed in SAR Section 9.2.4. The relief valves are 
calibrated annually to ensure that their pressure relief setting is 
correct or they are replaced with factory-set relief valves.
    Comment E.23: One commenter asked if there were holes in the shield 
jacket to add and drain things and indicated that holes would be a 
potential sabotage threat for someone to drain the jacket or add 
something dangerous to the water.
    Response: There are drain holes in the water jacket end plate. The 
125-ton HI-TRAC has two 1\1/2\-inch drain holes and the 100-ton HI-TRAC 
has four \3/4\-inch drain holes. The resultant dose rate for an assumed 
loss of the water jacket is addressed in Section 5.1.2 of the SAR. The 
analysis indicates that the off-site dose at 100 meters will be below 
the 5 rem accident limit in 10 CFR 72.106. In addition, NRC regulations 
in 10 CFR Part 72 have established physical protection and security 
requirements for an ISFSI located in the owner-controlled area of a 
licensed power reactor site.
    Comment E.24: One commenter stated that Conditions 1a and 1b of the 
CoC should both state that the cask system has two transfer casks.
    Response: The NRC disagrees with the comment. Condition 1b of the 
certificate of compliance specifies that there are two types of 
transfer cask options: the 125-ton HI-TRAC and the 100-ton HI-TRAC. It 
is not necessary to repeat that information in Condition 1a.
    Comment E.25: One commenter stated that there should be a drawing 
of the damaged fuel container in the CoC because the structure is not 
explained.
    Response: The NRC disagrees with this comment. A drawing of the 
damaged fuel container is included in Chapter 1 of the SAR and is 
available to the public. This level of detail is not necessary in the 
CoC.
    Comment E.26: One commenter asked what the screens are made of, how 
the screens are attached, if the screens can deteriorate or come loose 
over time, and what happens if the screens fall out.
    Response: As shown on the drawings in Chapter 1 of the SAR, the 
damaged fuel container, including the screen, is constructed of 
stainless steel. The damaged fuel container is an additional structural 
component that will make the MPC fuel basket even stronger. The screen 
is placed between two steel plates welded together with a 0.06 inch,

[[Page 25252]]

continuously 360 degree, all around fillet weld. It is not considered 
credible for the screens to fall off or fail. However, if a screen 
failed, there would be no release of radioactive material because the 
MPC is sealed. Small amounts of loose debris in the MPC have been 
considered during unloading operations, as described in SAR Section 
8.3.4.
    Comment E.27: One commenter stated that damaged fuel and intact 
fuel should not be placed in the same cask because it can cause 
potential problems in unloading.
    Response: The NRC disagrees with the comment. Damaged fuel can be 
stored safely with undamaged fuel. If damaged fuel is stored with 
undamaged fuel, then CoC, Appendix B, Section 2.1.1.c requires all fuel 
assemblies in the cask to meet the more restrictive heat generation 
requirements for the damaged fuel. Additionally, damaged fuel must be 
loaded into damaged fuel containers to enable safe handling during cask 
loading and unloading operations.
    Comment E.28: One commenter asked what the basis is for putting the 
hotter fuel in the center of the cask. The commenter also asked if the 
doses would be accurate if the lower dose fuel is placed at the 
periphery positions. The commenter stated that it would be better to 
have a more even heat and dose distribution in the MPC and asked if 
dose was more important than the heat.
    Response: The design of the HI-STORM cask considered both the 
thermal and radiological effects of the fuel. If one assumes the same 
enrichment and burnup (time that the fuel was left in the reactor to 
produce power), the fuel that is left longer to decay in the spent fuel 
pool (e.g., ``cooler fuel'') will generate less heat and high-energy 
radiation than the fuel that is removed sooner from the pool (e.g., 
``hotter fuel''). For the method used in the HI-STORM design, cooler 
fuel assemblies are stored on the periphery of the cask for two 
reasons. First, the ``cooler fuel'' assemblies have lower allowable 
peak clad temperature limits and the temperature of the assemblies on 
the periphery is cooler. Second, storing the ``cooler fuel'' on the 
periphery of the MPC provides some additional radiological shielding 
from the hotter fuel assemblies in the center.
    Comment E.29: One commenter asked if BPRAs and thimble plugs could 
be stored in the cask. The commenter stated that they should not be 
because they add weight.
    Response: BPRAs and thimble plugs have not been analyzed for this 
cask system and, therefore, are not authorized for storage in the HI-
STORM 100 at this time.
    Comment E.30: One commenter asked what the cask transfer station is 
and whether it had been designed yet. The commenter asked if it is 
constructed of reinforced concrete. The commenter stated that more 
explanation was necessary and that a drawing should be included. The 
commenter asked how and why an impact limiter is used. The commenter 
asked what the basis is and if an evaluation had been completed. The 
commenter was concerned with the use of terms ``if'' and ``shall be 
designed'' because this implies the CTF hasn't been designed. The 
design should have specific criteria.
    Response: The term ``cask transfer station'' in CoC, Appendix B, 
Section 3.5.1, is a typographical error and has been corrected to 
``cask transfer facility.'' A cask transfer facility (CTF) is a 
facility used for transferring the MPC between the transfer cask and 
the overpack. The CTF does not include 10 CFR Part 50 controlled 
structures such as the fuel handling building or reactor building. The 
NRC disagrees that a drawing of the CTF or more design details are 
necessary. The HI-STORM 100 Cask System is approved for use under the 
general license provisions of 10 CFR Part 72. Therefore, the cask may 
be used in any nuclear power reactor site licensed under 10 CFR Part 
50, provided that the site parameters are enveloped by the cask design 
bases. The specific design and operation of the CTF will be dictated by 
site-specific needs. Because of the varied needs of each reactor site, 
the NRC found it impractical and unnecessary to review and approve a 
specific CTF design, including the specific materials of construction. 
The NRC reviewed and approved the criteria for the design, 
construction, and operation of the CTF. These criteria are specified in 
CoC, Appendix B, Section 3.5, and SAR Section 2.3.3.1.
    The impact limiter is a possible CTF design feature whose function 
would be a defense-in-depth measure because the lifting equipment used 
in the CTF must be designed to preclude a drop. As discussed in the 
response to comment J.14, the specific requirement for an impact 
limiter has been eliminated because other methods may be available to 
prevent a canister breach in case of a canister drop during transfer 
operations.
    Comment E.31: One commenter stated that we should clearly state 
what the CTF is and make sure that every detail of the procedure is 
carefully analyzed because it is vague.
    Response: The CTF is defined in SAR Section 2.3.3.1 and in CoC, 
Appendix B, Section 1.0. Detailed design and operational requirements 
for the CTF are also specified in SAR Section 2.3.3.1, as well as in 
CoC, Appendix B, Section 3.5. Under the provisions of 10 CFR 72.212 and 
CoC Condition 2, each licensee that elects to use a CTF must develop 
written procedures for operating the CTF. These procedures are subject 
to NRC review during inspection. As required by TS 5.2.h, the licensee 
must conduct a dry run training exercise, prior to first use of a CTF, 
to demonstrate that the procedures can be conducted safely and 
successfully.
    Comment E.32: One commenter recommended that Design Drawing 1495, 
Sheets 4 and 6 and Design Drawing BM-1575, Sheet 2 be revised.
    Response: The NRC agrees with the comment. These changes correct 
drafting errors or provide a level of flexibility that is acceptable to 
the NRC staff. The SAR drawings have been revised accordingly.

F. Welds

    Comment F.1: One commenter asked how use of the trunnions puts 
stress on the weld at the water jacket.
    Response: Use of the pocket trunnion does not put any stress on the 
water jacket. The seal weld between the pocket trunnion and water 
jacket shell is for retaining the water inside the water jacket. The 
pocket trunnion is attached to the outer transfer cask shell by full 
penetration welds all the way around the trunnion. When the pocket 
trunnion is used, the force is transferred to this weld and not to the 
seal weld on the water jacket. The other type of trunnion on the 
transfer cask is the lifting trunnion. The lifting trunnion is not 
connected to the water jacket and, therefore, puts no stress on the 
water jacket.
    Comment F.2: One commenter asked how the welds are checked to be 
leakproof and whether water can enter the trunnion.
    Response: All the structural welds have to be examined and 
inspected according to the applicable ASME code. The welded joint is an 
integral part of the structure and is leak proof. Because the trunnions 
are made of solid steel, water cannot leak into them.
    Comment F.3: One commenter stated that the lid and closure ring of 
the MPC should be full penetration welds and should be ultrasonic 
tested (UT) as this is the basis for qualification as a redundant seal.
    Response: The NRC disagrees with this comment. Full penetration 
welds are unnecessary from the structural and

[[Page 25253]]

containment boundary requirements of the design. Employing 
unnecessarily heavy welds leads to fabrication problems such as 
excessive warpage. UT of heavy section, full or partial penetration, 
austenitic stainless steel welds to ASME Code acceptance criteria is 
not feasible with current technology. The redundant seal concept is 
based upon the use of two welds forming the leak barrier, not the 
inspection method. With redundant welds, one weld could leak and the 
second still provide leak tight integrity.
    Comment F.4: One commenter stated that just because there is no 
known plausible, long-term degradation mechanism to cause seal welds to 
fail doesn't mean that the welds won't fail.
    Response: The NRC disagrees with this comment. The NRC staff has 
examined the plausible mechanisms that would cause failure of the seal 
welds and has determined that those mechanisms are inoperative under 
normal service and design accident conditions for HI-STORM. This gives 
the NRC staff reasonable assurance that the welds will not fail under 
design basis normal, off-normal, and accident conditions.
    Comment F.5: One commenter asked how lid welds are removed in 
unloading and stated that the procedures should be in the documents 
before the casks are loaded.
    Response: The NRC disagrees with this comment. Welded cask lids may 
be removed by one of several methods. The method of removal and the 
detailed procedures (as opposed to the general procedures of the SAR) 
are the responsibility of the ISFSI licensee, subject to NRC review and 
inspection. The TSs require ISFSI licensees to perform lid removal 
method demonstrations on full-size mock-ups as part of their pre-
operational testing and training exercises. The NRC staff has reviewed 
and inspected several methods and their associated procedures. 
Inclusion of such detailed procedures in the SAR is unnecessary.
    Comment F.6: One commenter asked what happens if the water jacket 
welds leak water.
    Response: The resultant dose rate for an assumed loss of the water 
jacket is addressed in Section 5.1.2 of the SAR. The analysis indicates 
that the off-site dose at 100 meters will be below the 5 rem accident 
limit in 10 CFR 72.106.
    Comment F.7: One commenter stated that the penetrant test (PT) is 
unacceptable, that the criteria for layers and time are ``wishy 
washy,'' and that PT tests should not be allowed.
    Response: The NRC disagrees with this comment. The commenter has 
not specified why PT is unacceptable or why it should not be allowed. 
PT is a Code accepted examination method. The progressive PT technique 
is used and accepted in the nuclear industry when a volumetric 
examination by means such as UT is impractical. UT is unsuitable for 
heavy section austenitic stainless steel welds.
    The basis for the structural lid weld examination methods is 
documented in the NRC's Interim Staff Guidance-4, Revision 1, that 
allows the use of a multi-layer (i.e., progressive) PT examination in 
lieu of a volumetric examination.
    Comment F.8: One commenter stated that welds need to be checked 
carefully.
    Response: The NRC agrees with the comment. Welds are important 
which is why they are examined and inspected according to the 
applicable ASME code.
    Comment F.9: One commenter stated that the leak testing procedure 
used to demonstrate MPC closure cannot be performed as described and 
that performance of the test is not generally consistent with ANSI 
N14.5-1997, ``Leakage Tests on Packages for Shipment.'' Consequently, 
containment of the radioactive material to the stated criteria cannot 
be demonstrated. The principal reason provided by the commenter is that 
the nominal concentration of helium in air is 5 parts per million. This 
atmospheric concentration masks any leakage from the MPC using the 
specified test conditions. In addition, the commenter noted that there 
is no direct reference to definitions, equations, formula, methodology, 
or criteria of the standard in the text. The commenter further noted 
that when terminology from the standard is given, it is (for the most 
part) used incorrectly.
    Response: The NRC disagrees with the comment. These welds are 
multipass stainless steel welds that are dye penetrant examined 
multiple times during the weld process. The multiple dye penetrant 
examinations provide reasonable assurance of high integrity welds that 
will retain the inert gas and prevent leakage of radioactive material 
into the environment. The leakage testing of these welds provides 
additional insight into the leak-tightness of these welds.
    The NRC staff has reasonable assurance that the leakage test 
procedure outlined in SAR Section 8.1 can be performed as described 
provided that appropriate equipment is used and the leak test method is 
properly qualified. The leakage testing for the lid is to be performed 
with a sniffer type probe; however, the test method for the port and 
drain covers is not specified in the SAR. There are test methods 
discussed in Appendix A of ANSI N-14.5 that could be used to perform 
the port and drain cover leakage testing. Detailed procedures are 
developed by the user who is responsible for ensuring that the TS 
limits are met and therefore, confinement is adequately maintained. The 
leakage testing will require a demonstration that the detector can 
identify an appropriate calibrated leak in the presence of background 
helium. Although sniffer probes detect discrete leaks rather than an 
integrated leakage rate, the typical sniffer probe sensitivity of 
10-\8\ provides reasonable assurance that the TS leakage rate limit of 
5 x 10-\6\ will not be exceeded.
    As stated in the SAR, the leak testing will be performed in 
accordance with ANSI N14.5. It is not necessary to include any more 
level of detail in the SAR; therefore, no change to the SAR is 
necessary. Appropriate detail will be included in the site procedures.
    The SAR has been changed to use terminology consistent with the TS 
and the 1997 revision of ANSI N14.5. The terminology was changed from 
std cc/s to atm cc/s. SAR section 7.3.3 justifies the use of the units 
atm-cc/sec. Also, the SAR was revised to delete the sensitivity of the 
detector. The sensitivity will be addressed in the detailed site 
procedures.

G. Structural

    Comment G.1: One commenter stated that all of the accident level 
events and conditions listed in the SER, particularly a transfer cask 
handling accident or sabotage, should be evaluated for structural 
analysis in a jamming condition on top of the overpack.
    Response: All design basis normal, off-normal, and accident events 
have been evaluated in the structural analyses and are discussed in 
Section 3 of the SER. This includes an evaluation of the transfer cask 
under a 42-inch horizontal drop during transfer operations. A 
horizontal drop from a greater height is not considered because the 
horizontal lifting height limit for the transfer cask is 42 inches. The 
evaluation shows that the HI-TRAC meets all structural requirements and 
there is no adverse effect on the confinement, thermal, or 
subcriticality performance of the contained MPC.
    As discussed in the response to C.15, vertical drop of a transfer 
cask is not considered credible because vertical lifting of a loaded 
transfer cask must be performed with structures and components designed 
to prevent a cask drop. Also, as discussed in the response to E.5, 
jamming is not considered to be

[[Page 25254]]

credible because of the design of the HI-STORM system. The threat of 
sabotage is beyond the scope of this rule and is discussed in the 
response to C.8.
    Comment G.2: One commenter stated that bending of the web and 
pushing of the flanges possibly accompanied by some local weld failures 
sounded feasible and may not result in limited deformation as assumed. 
The commenter asked if a full size cask had been tested in a drop or 
tipover.
    Response: The channels attached to the inner shell of the overpack 
are not classified as important-to-safety and serve no structural 
purpose. Deformation of the channels, whether limited or complete 
collapse, does not affect retrievability of the MPC. On the contrary, 
the deformation of these channels due to a tipover accident absorbs 
energy which reduces the deceleration loadings to the MPC and provides 
a greater opening in the overpack during retrieval. NRC regulations do 
not require full size testing of casks. The applicant can choose the 
method of analysis. Computer analyses have been performed to determine 
the responses of a cask in drop and tipover accidents.
    Comment G.3: One commenter questioned how the structural analysis 
conducted for the 125-ton HI-TRAC transfer cask could bound the 100-ton 
version and indicated that the 100-ton version needs its own analysis.
    Response: All the structural analyses and evaluations of the 125-
ton transfer cask were repeated for the 100-ton transfer cask. However, 
the analytical results of the 125-ton transfer cask are greater than 
that of the 100-ton transfer cask. Therefore, the structural analysis 
of the 125-ton transfer cask bounds the 100-ton transfer cask.
    Comment G.4: One commenter asked what would be the consequences of 
the deformation of the outer shell and lead and water jacket from a 
missile, particularly if the transfer cask was on top of the concrete 
shell.
    Response: The HI-TRAC transfer cask is always held by the handling 
system while in a vertical orientation completely outside of the fuel 
building. Therefore, considerations of instability due to a tornado 
missile strike are not included in the evaluation. However, a 
structural evaluation of the damage to the HI-TRAC transfer cask from 
an intermediate missile strike and a large missile strike is performed. 
The evaluation shows that the outer shell and the water jacket would 
not experience any plastic deformation and will not adversely affect 
the retrievability of the MPC.

H. Materials

    Comment H.1: One commenter questioned why carbon steel was used for 
the inner and outer plate instead of stainless steel because of the 
concern over corrosion. The commenter also asked if the carbon steel 
was coated.
    Response: The materials used in the fabrication of the cask are 
described in Chapters 1 and 3 of the SAR and discussed in Section 3.3 
of the NRC SER. These materials have been found acceptable because they 
meet the requirements for their respective applications in the cask 
system. They are suitable for the expected loading and storage in wet 
and dry environments, including corrosion and galvanic effects. There 
is no requirement for designers to select materials from a given class, 
e.g. stainless steels.
    The carbon steel used in the overpack is protected from corrosion 
by an industrial epoxy coating commonly used for the protection of 
steel.
    Comment H.2: One commenter stated that one alloy should be 
specified for cask fabrication instead of allowing a choice because if 
later problems develop, there are fewer variables.
    Response: The NRC disagrees with the comment. The materials used in 
casks are selected on the basis of the needed properties. Allowing a 
choice of more than one material or alloy for fabrication is acceptable 
provided that each of the options has the appropriate properties. The 
materials chosen for use in the Holtec HI-STORM 100 design have a long 
history of favorable performance in the nuclear industry.
    Comment H.3: One commenter questioned why plain concrete is not 
included in NUREG-1536 and why an exemption was being given to allow 
plain concrete since reinforced concrete is stronger.
    Response: No exemption was given to allow plain concrete to be used 
for structural components. The plain concrete in the HI-STORM 100 
overpack is for shielding only and is not a structural component of the 
overpack. The reinforced concrete included in NUREG-1536 is for 
concrete structures (concrete components that provide structural 
strength) only. The HI-STORM 100 overpack is a welded steel structure, 
not a concrete structure.
    Comment H.4: One commenter asked if the NRC has reviewed the 
manufacturers direction for the carboline 890 and thermaline 450 
coatings. The commenter asked how the coatings are used and applied, 
and if they will wash or flake off in pool water, making the water 
cloudy.
    Response: The NRC staff has reviewed the manufacturer's technical 
information for the coatings mentioned. Both coatings are standard 
coatings employed in industry for immersion service and are applied 
using common industry tools and techniques. No performance problems 
would be expected during intended service.
    Comment H.5: One commenter asked if the carbon steel caused 
reactions that could create loading or unloading problems such as 
reaction products clogging venting or draining equipment with crud or 
flakes or making the pool water cloudy.
    Response: Carbon steel exposed to the cask loading environment 
produces very fine particulates that do not clog equipment. Turbidity 
that may arise from corrosion of uncoated carbon steel can be 
controlled with appropriate water treatment equipment.
    Comment H.6: One commenter asked if temperature or coatings on the 
channel could affect the fit. The commenter also asked if flaking of 
the coating could clog a channel slide or if corrosion in the channels 
could cause problems in unloading.
    Response: The effects of temperature on the channels have been 
calculated and do not affect the fit. Each coat of the epoxy paint 
applied to the exposed surfaces of the inner components of the overpack 
is, at maximum, 0.008 inches thick. Two coats result in a maximum 
diametral reduction in inside diameter of 0.032 inches. This reduction 
will not affect the fit. Both the interior and the exterior of the 
channels are coated to prevent corrosion.
    Comment H.7: One commenter asked if aging was factored into the 
analysis of the pad and stated that the specific site should be 
evaluated for a full cask array.
    Response: Concrete is resistant to environmental conditions, 
including air pollution and moisture. Therefore, the NRC staff expects 
no significant degradation of the pad during the licensed lifetime of 
the ISFSI facility. Each proposed site is subject to a specific 
evaluation to ensure that the design parameters satisfy site-specific 
conditions. In addition, cask users are responsible for inspecting and 
maintaining the pad, and for ensuring that significant degradation is 
not occurring over time.
    Comment H.8: One commenter asked what the condition of the concrete 
is right under the shell and expressed concern that the concrete could 
crack where nobody would see damage needing repair.
    Response: As discussed in the response to E.3, the main function of 
the concrete encased between the steel shells in the HI-STORM 100 
overpack is shielding. The structural strength of

[[Page 25255]]

the HI-STORM 100 overpack is provided by the inner and outer carbon 
steel shells. Cracking of the concrete would not have a significant 
impact on the cask's ability to meet the regulatory dose limits. There 
is no credible mechanism for the concrete to undergo any significant 
damage. Thus, inspection of the concrete is not necessary.
    Comment H.9: One commenter asked if concrete expanded or released 
water or gas when it is superheated.
    Response: Concrete contains some traces of free water. If the water 
is heated, it will evaporate. Concrete will expand upon heating and 
contract upon cooling. The amount is governed by the temperature. These 
expansions/contractions are reversible and not permanent. There are no 
significant effects of expansions/contractions that would occur even if 
the temperature went considerably beyond the design temperature 
parameters.
    Comment H.10: One commenter asked how the bottom face affects the 
supporting surfaces (heat, radiation, weight, stress, pressure etc.).
    Response: As listed in Table 4.4.9 of the SAR, the temperature of 
the bottom lid plate at normal conditions is 183 deg.F. That 
temperature will not have an adverse effect on the concrete. Radiation 
will have minimal impact on the concrete pad due to the shielding 
provided by the pedestal. The weight, stress, and pressure from the 
cask bottom have no adverse effect upon the pedestal or slab because 
they are specifically designed to support all the loads due to the 
casks.
    Comment H.11: One commenter asked how the gas and liquid media that 
escapes from the damaged fuel container interacts with other materials 
in the MPC and if they can cause problems.
    Response: The materials of the cask have been selected to be 
compatible with any constituent or reaction product of the fuel.

I. Thermal

    Comment I.1: One commenter asked if hot spots in the cladding could 
cause lead to sag in the transfer cask if the inner canister is in 
place and the temperature is close to the boiling point of the water 
pack.
    Response: Hot spots in the cladding would not result in sagging or 
melting of the lead. The bounding calculation performed by Holtec 
assumed all the fuel assemblies were at the design basis limit (hottest 
assemblies). The bounding rod cladding temperature occurs at the center 
of the MPC and does not have a direct impact on the lead. The 
assemblies on the periphery of the MPC are significantly cooler because 
they are located near the cooler surface of the MPC. Table 11.2.8 in 
the SAR provides the results from a calculation that assumes no water 
in the water jacket. These results bound the impact of boiling in the 
water pack. Based on those results, it can be concluded that the lead 
temperature remains well below the melting temperature.
    Comment I.2: One commenter asked what happens if the water in the 
transfer pack boils and the steam pressure builds up, and stated that 
this situation should be evaluated.
    Response: As the pressure builds up, the pressure is relieved 
through a safety valve. As water is removed through the safety valve, 
the temperature of the water remains at the saturation temperature. The 
case of water boiling in the HI-TRAC water jacket is bounded by the 
event that assumed no water in the water jacket. This event leads to a 
temperature in the water jacket that is higher than the saturation 
temperature of the water. The impact of loss of water in the water 
jacket is summarized in Table 11.2.8 of the SAR.
    Comment I.3: One commenter asked how, during normal conditions, the 
temperature of the outer shell could be higher than the temperature of 
the concrete because the carbon steel would breathe less than the 
concrete, causing the heat to be retained in the concrete. The 
commenter also asked how the temperature of the concrete could be 
measured since it is encased in the carbon steel.
    Response: The question raised by the commenter is not clear. The 
temperature of the concrete is higher than the temperature of the outer 
shell under normal conditions. Reviewing Table 4.4.9 in the SAR, the 
temperature at the overpack outer shell is not higher than the concrete 
cross sectional average temperature. The temperature distribution 
through the overpack under normal conditions is listed in Table 4.4.9 
of the SAR (e.g., 149  deg.F for the concrete and 131  deg.F for the 
outer shell).
    With regard to the question of measuring the temperature of the 
concrete, the applicant does not measure the temperature of the 
concrete. Bounding calculations are used to assure that the concrete 
temperature limits will not be exceeded.
    Comment I.4: One commenter asked how the pad reacts to the bottom 
plate of a cask from a temperature differential standpoint. The 
commenter asked if the pad would crack and sink under each cask and 
form a concave area that could then collect moisture. The commenter 
further asked if the collected moisture could boil and if the moisture 
could cause the bottom plate to rust.
    Response: The heat transfer between the bottom plate of the 
overpack and the concrete pad is modeled in the thermal computer code 
for the HI-STORM cask system. As listed in Table 4.4.9 of the SAR, the 
temperature of the bottom lid plate at normal condition is 183  deg.F. 
That energy is transmitted to the concrete pad down to the ground, 
which is at the normal soil annual average temperature of 77  deg.F. 
Therefore, the concrete will not experience temperatures above boiling 
(no superheating will occur). In the winter, the concrete will not 
reach freezing temperatures below the cask because it generates heat. 
If the concrete reaches or exceeds boiling or freezing temperatures, 
there is no detrimental effect on the strength or condition of the 
concrete. The pad is specifically designed to support the weight of the 
casks without any cracking or sinking of the pad. The bottom plate of 
the cask is stainless steel and will not rust.
    Comment I.5: One commenter questioned the basis and validity of 
simulating the heat effect of adjacent casks radiating heat back to an 
interior cask and if an analysis of the real situation had been 
conducted.
    Response: The impact of radiation heat transfer from neighboring 
casks was calculated in the HI-STORM thermal evaluations. The method 
used by the applicant was to assume that all of the radiated heat is 
reflected back to the cask. This modeling assumption is equivalent to 
assuming that the cask was totally encircled by other casks. In 
reality, less heat will be radiated back to the cask; therefore, the 
calculations bounded the effects of neighboring casks. The impact of 
neighboring casks was shown to be minimal. The NRC staff does not 
require validation of the analytic method with actual experimental 
data.
    Comment I.6: One commenter asked why the analysis assumed that the 
soil below the overpack was at a constant temperature because the casks 
could cause hot spots.
    Response: The analyses did model the hot spots below the overpack. 
The computer simulation of the overpack modeled the concrete pad that 
the overpack is placed on and the temperature of the soil below the 
concrete pad. The soil is one of several paths for heat to leave the 
cask. The most significant path for heat dissipation is through the air 
passage between the MPC and the overpack. The applicant used the 
highest annual average soil temperature found in the USA. The purpose 
for using the highest average temperature for the soil and air

[[Page 25256]]

in the thermal analyses is to demonstrate fuel retrievability and that 
the cladding is protected during storage against degradation that leads 
to gross ruptures (10 CFR Part 72.122). One acceptable method for 
demonstrating that the cladding will not undergo gross rupture is to 
place a limit on the allowable cladding temperature such that 
reasonable assurance exists that the cladding will not significantly 
degrade. A report by the Pacific Northwest National Laboratory, PNL-
6189, dated May 1987, provides one acceptable approach for establishing 
a temperature limit. The PNL method is conservative when compared to 
the maximum allowable degradation permitted in Part 72 of the 
regulations. This method, in conjunction with the maximum annual 
average temperature, solar heating (e.g., insulation), analytic 
assumptions, etc., provide reasonable assurance that the requirements 
of Part 72 will be met.
    Comment I.7: One commenter asked why an exception was allowed for 
exceeding the short term temperature limit for the fire accident 
scenario and stated that an exception should not be allowed.
    Response: The American Concrete Institute (ACI) establishes 
temperature criteria for concrete. One, but not the only, acceptable 
demonstration that the concrete overpack will maintain its intended 
function is to meet the temperature criteria in ACI 349. However, as 
stated in the NRC staff's Standard Review Plan (NUREG-1536), ``a small 
amount of exterior concrete spalling may result from a fire, the 
application of fire suppression water, rain on heated surfaces or other 
high-temperature condition. The damage from these events is readily 
detectable, and appropriate recovery or corrective measures may be 
presumed. Therefore, the loss of such a small amount of shielding 
material is not expected to cause a storage system to exceed the 
regulatory requirements in 10 CFR 72.106 and, therefore, need not be 
estimated or evaluated in the SAR. The NRC accepts that concrete 
temperatures may exceed the temperature criteria of ACI 349 for 
accidents if the temperatures result from a fire.'' The Holtec analysis 
demonstrated that the amount of concrete that exceeds the ACI 
temperature limit is very limited and would not pose a significant 
safety hazard.
    Comment I.8: One commenter asked for the basis of using an average 
temperature of the gas in the gap and plenum of the limiting rod and 
questioned the validity of the assumption.
    Response: The purpose for evaluating the average of the gas 
temperature in the fuel rod is to calculate the pressure within the 
fuel rod. The computer code used in the analysis calculates the 
temperature profile of the fuel rod, but does not calculate the 
corresponding pressure for that rod. To calculate the pressure, the 
average temperature of the gas is calculated and from the ideal gas 
law, the corresponding pressure is established.
    Comment I.9: One commenter asked what is in the water used for 
forced water circulation under wet transfer of the fuel from the spent 
fuel pool to the location for vacuum drying and if the water could 
chemically affect other materials in the cavity. The commenter asked 
how fast the water flows, if steam could be formed, and if the water 
could physically affect other materials in the cavity, movement of 
rods, flaking of paint, etc.
    Response: The licensee can either use demineralized water or water 
from the spent fuel pool. Neither demineralized nor spent fuel pool 
water would adversely interact with the system. The flow rate of the 
water is based on the heat output of the fuel assemblies and is a site-
specific issue.
    Comment I.10: One commenter asked what the water chiller is used 
for and what material is used as the chilling medium.
    Response: The water chiller is used as the heat sink for cooling 
the helium inside the MPC to below 200 deg.F. The type of water chiller 
used is a site-specific issue and not part of this rulemaking activity.
    Comment I.11: One commenter asked for specific criteria that 
defines clearance around the cask for cooling purposes instead of 
stating a reasonable amount. The commenter also asked how close other 
heat sources may be located and what is considered to be a significant 
heat source.
    Response: The actions identified by the commenter are only valid 
when a breakdown occurs in the helium coolers (LCO 3.1.3). Section 
B3.1.3 of the technical specification bases states that ``if the 
TRANSFR CASK is located in a relatively open area such as a typical 
refuel floor, no additional actions are necessary.'' However, a 
licensee may elect to perform the cooling with the cask located in a 
pit or vault. This is a site-specific activity. The bases identify 
three acceptable options for ensuring adequate heat transfer for the 
TRANSFER CASK. The user may develop other alternatives on a site-
specific basis, considering actual fuel loading and decay heat 
generation within the cask. One of the options is to fill the annulus 
between the MPC and the TRANSFER CASK with water. The second option is 
to remove the TRANSFER CASK from the pit or vault and place it in an 
open area such as the refueling floor with a reasonable amount of 
clearance around the cask and not near a significant source of heat. 
The third option is to supply nominally 1000 SCFM of ambient air to the 
space inside the confined space (e.g., pit or vault). With respect to 
defining an acceptable distance, the licensee could use the analyzed 
event of 15 feet center-to-center storage spacing that corresponds to a 
four foot clearance. Smaller clearances would also be acceptable, given 
the heat load rating of the cask and ambient conditions. With regard to 
defining a significant heat source, this is a site specific 
consideration. For example, if the plant is using a bank of radiant 
heaters near the cask, then an evaluation needs to be performed to 
ensure that those heaters pose no adverse impact on the cask. These 
options are only guidelines to an LCO that a user would have to 
consider.
    Comment I.12: One commenter asked how cool air was provided to the 
space inside the vault at the bottom of the overpack. The commenter 
stated that this needs to be planned out ahead of time for ALARA 
considerations and equipment availability.
    Response: This is a site-specific issue and not part of this 
rulemaking activity. The NRC staff agrees with the comment that the 
user needs to plan this activity considering ALARA and equipment 
availability.
    Comment I.13: One commenter stated that fuel should be adequately 
cooled before it goes into the transfer cask.
    Response: The NRC agrees with the comment that adequate planning is 
needed when performing cask cooldown and reflooding. The purpose of the 
analyses performed in the SAR is to maintain the integrity of the fuel. 
The requirements on the burnup and minimum cooling time serve that 
purpose.
    Comment I.14: One commenter asked if the temperature of the helium 
accurately reflects the internal temperature of the MPC and stated that 
this should be tested.
    Response: The exit temperature of the helium reflects the 
conditions of the fuel rods. After the helium temperature is reduced 
below 200 deg.F, the bulk of the fuel will be at low temperatures, 
minimizing the potential for excessive steaming. Reflooding of a 
canister has been demonstrated without pre-cooling the helium. No 
additional tests are needed.

[[Page 25257]]

    Comment I.15: One commenter objected to the addition of ethylene 
glycol solution to the demineralized water in the water jacket to 
prevent freezing and asked where this had been tested. The commenter 
also asked why the antifreeze was used, how the solution would mix, how 
the NRC knows it will work, what types of effects it could have on the 
inside of the water jacket and the channel walls, if it would add 
weight, and how it is added to the water if the jacket is welded shut.
    Response: Ethylene Glycol is the chemical name for ordinary 
antifreeze. Adding antifreeze to the water jacket, located on the 
outside of the HI-TRAC transfer cask, is an option if the user elects 
to move a loaded MPC in cold weather, down to 0 deg.F. The use of 
antifreeze in the water jacket does not add appreciable weight to the 
HI-TRAC cask. Although the water jacket is a welded system, openings 
are designed to add and remove water from the water jacket. Antifreeze 
has been used in many applications to keep water from freezing. The NRC 
staff believes that the industry has ample experience with antifreeze 
that additional testing and validity is not necessary. Mixing of the 
antifreeze is a site-specific issue that will ensure that the proper 
amount of antifreeze is added to prevent the water from freezing at 
temperatures down to 0 deg.F.
    Comment I.16: One commenter recommended the addition of a note to 
Tables 4.4.20 and 4.4.21 of the SAR to provide clarification for the 
heat loads.
    Response: The NRC disagrees with the comment. SAR Tables 4.4.20 and 
4.4.21 refer to loading the MPC with uniformly aged fuel assemblies 
emitting heat at the design basis maximum rate. Section 4.4.2 
identifies these assemblies as the limiting design basis fuel 
assemblies.
    Comment I.17: One commenter stated that Holtec's use of a two-by-
four block array to be equivalent to an infinite array assumed a 
center-to-center distance between casks of 18.6 feet. The commenter 
stated that this equivalency determination between an infinite array 
and a two-by-four array is invalid where the differences in cask 
spacing do not meet the 18.6-feet center-to-center assumption 
underlying the analysis. The commenter noted that the PSF facility 
design uses a 15-foot center-to-center distance. The commenter stated 
that any CoC issued for this cask system must address this shortcoming.
    Response: The NRC disagrees with the comment. First, the PFS 
facility is outside the scope of this rulemaking activity. Second, as 
identified in Table 1.4.1 of the SAR, the analysis of a 2 by N array 
was performed using a pitch of 13.5 feet, not 18.6 feet. The 18.6-feet 
pitch is used for a square array. When calculating an equivalent 
hydraulic diameter for the square array and taking into account that 
the center-to-center spacing of neighboring casks between the pads is 
38 feet, as described in Figure 1.4.1 of the SAR, the hydraulic 
diameters for the two cases (square array versus 2 by N array) is the 
same.
    Comment I.18: One commenter stated that thermal interaction of 
casks through radiative heat transfer should be considered. The 
commenter also stated that the assumption that individual casks will 
not interfere with cooling air supply of each other may not be correct.
    Response: The NRC agrees that neighboring casks can have an 
influence on each other. However, these influences have a second order 
impact on the results. The analyses performed by Holtec did credit 
radiation heat transfer between the neighboring casks. A bounding 
calculation was performed with an ambient temperature of 125 deg.F. 
That calculation accounted for heat reflected by the hot concrete pad 
and heat generated by neighboring casks. Although not required by the 
NRC staff's review of the SAR submittal, Holtec, in response to other 
inquiries, performed a sensitivity study to quantify the impact of 
neighboring casks and the impact of the sun heating the concrete pad.
    The impact of increasing the spacing between casks by a factor of 
five in the radial direction resulted in a decrease in the peak cask 
surface temperature of 16 oF for an ambient temperature of 100 deg.F 
and 17 deg.F for an ambient temperature of 125 deg.F. The impact on 
peak clad temperature resulted in a decrease of 6 deg.F for an ambient 
temperature of 100 deg.F and a decrease of 8 deg.F for an ambient 
temperature of 125 deg.F. Because the peak clad temperature is on the 
order of 760+ deg.F, the impact of neighboring casks is minimal.
    Comment I.19: One commenter stated that the temperature of the 
reflecting boundary should be taken as the temperature of the cask in 
interaction with the other casks and not the temperature of an isolated 
cask.
    Response: The NRC agrees that one method for calculating the impact 
of neighboring casks is to model the neighboring casks in the array. 
Another acceptable method, that was used by the applicant, is to model 
the limiting (highest temperature) cask and assume that all the 
radiation it emits is reflected back. This analysis bounds the amount 
of radiation that neighboring casks can impose on the center cask. This 
bounding analysis is acceptable. As noted in the response to comment 
I.18, above, the impact of neighboring casks is minimal, given the 
significant margins between the allowable temperatures and the bounding 
calculated temperatures.
    Comment I.20: One commenter stated that the Holtec model does not 
appear to take into account that the heating of the concrete pad is 
likely to diminish the ``chimney effect'' of the intake and outlet 
vents. The commenter stated that if Holtec had taken this effect into 
account, the calculated temperature would be higher in Revision 9 of 
the SAR.
    Response: In a response to other inquires, Holtec performed 
calculations to quantify the effect of concrete pad heating on the cask 
performance. For the bounding 125 deg.F ambient temperature event, 
neglecting the heat reflected by the pad resulted in a reduction of 
cask surface temperature of 10 deg.F and a reduction in peak clad 
temperature of 6 deg.F. These temperature differences illustrate that 
the concrete pad has negligible impact on the cask.
    Comment I.21: One commenter stated that ambient temperature should 
be defined due to the importance of the term. The commenter noted that 
ambient temperature is an important assumption in the thermal 
calculations and an important design element in the CoC. The commenter 
stated that the gross oversimplification of the concept of ambient 
temperature renders the Holtec thermal analysis completely useless. The 
commenter noted that Holtec assumes that the ambient temperature at the 
intake and outlet vents is the same; however, the temperature at ground 
level will be significantly higher than it will be some distance above 
due to the ground absorbing solar energy. The commenter stated that a 
desert may have a surface temperature of 180 deg.F, much higher than 
the 80 deg.F assumed by Holtec as an intake temperature. This would 
reduce the effective buoyancy and air velocity through the cooling 
ducts and result in a higher fuel cladding temperature.
    Response: The NRC disagrees with the comment. The thermal response 
of a cask is very slow. This is due to the large mass of the system. An 
analogy can be reached by observing the buildings constructed in the 
desert. Massive concrete is used to maintain the indoor temperatures at 
reasonable conditions where air conditioners do not exist. The 
temperature in those regions fluctuates over each day. For these 
structures, an estimate of the average conditions can be assessed by 
assuming a bounding average daily temperature. Holtec used such a 
method. In addition, the method

[[Page 25258]]

assumed the maximum solar heating specified in 10 CFR Part 71 averaged 
over a 24-hour period. Holtec used bounding assumptions approved in the 
NRC staff's SER.
    Comment I.22: One commenter stated that NRC should have reviewed 
the inputs and outputs of the FLUENT calculation. The commenter also 
stated that the NRC should have conducted an independent analysis and 
validation of the thermal model employed by Holtec. The commenter 
stated that the HI-STAR analysis cannot be extrapolated to the HI-STORM 
cask because the casks are constructed of different materials, with 
different methods of heat dispersion. The commenter stated that the NRC 
performed a superficial review and had abdicated its role as 
independent regulator and should not issue a CoC for the HI-STORM 100 
cask system because there is no lawful basis.
    Response: The NRC disagrees with the comment. The NRC staff's 
review of the HI-STORM system was not superficial. This is clearly 
demonstrated by the NRC staff's requests for additional information, 
the applicant's many revisions to the SAR to address NRC staff 
concerns, and commitments made by the applicant as outlined in Appendix 
12B of the SAR. The NRC staff does not perform independent confirmatory 
calculations for every analysis submitted in an application, nor does 
the NRC staff routinely review the inputs and outputs of the computer 
calculations without cause. Independent analyses that duplicate the 
extensive computer calculations performed by an applicant may, at 
times, be performed for various reasons. Some reasons include, but are 
not limited to, concern that a major error exists in the calculations; 
allegations that calculations were improperly performed; use of new 
modeling techniques not previously reviewed by the NRC staff; crediting 
heat transfer mechanisms not previously reviewed by the NRC staff; 
concern that the margin in a complex analysis is small; and concern 
that little conservatism exists in the modeling approach.
    For the HI-STORM application, the NRC staff reviewed the basic 
assumptions used in the calculations, as identified in the SAR and in 
the NRC's requests for additional information. A detailed review of 
every number is not warranted. As for performing independent analysis 
and validation, the NRC staff was able to reach its safety findings 
without the need for such calculations. The need for these calculations 
is case specific, as addressed above. For HI-STORM, the applicant used 
computer codes that are employed by the NRC and have been found 
acceptable. The applicant demonstrated its knowledge of the code by 
benchmarking its methodology with a full-scale spent fuel cask 
instrumented with thermocouples, validating its thermal model and 
providing reasonable assurance that its analysts have good working 
knowledge of the code to perform the required calculations. The NRC 
staff's review of the applicant's methods and assumptions indicate 
ample margin and conservatism in the analyses.
    The HI-STORM application review process was conducted under NRC 
policy and guidance, and as required by the regulations in 10 CFR Part 
72. Regarding the reference to the HI-STAR analysis in Section 4.5.4 of 
the preliminary SER, the NRC staff intended to indicate that it was 
aware that Holtec's use of the FLUENT code had been previously found 
acceptable for the HI-STAR application. This reference was not intended 
to imply that the NRC staff relied on the HI-STAR calculations or the 
prior evaluation in its evaluation of the HI-STORM cask. Section 4.5.4 
of the SER has been modified to clarify the description of the NRC 
staff's review. Also, to better illustrate the NRC staff's review of 
the applicant's submittal, Section 4 of the SER was supplemented with 
additional information.
    Comment I.23: One comment indicated that the SER states that the 
ambient temperature under normal conditions must be less than 80 deg.F. 
In addition, the commenter believed Holtec assumed that the ambient 
temperature at the inlet and outlet vents is the same and did not 
consider warming of the air by heat generated by neighboring casks and 
the concrete pad. The commenter stated that calculations indicated that 
a desert may have a surface temperature of 180 deg.F and that the 
temperature 0.5 m above the ground would be 130 deg.F.
    Response: The NRC disagrees with the comment. The SER does not 
state that the ambient temperature under normal conditions must be less 
than 80 deg.F. The applicant evaluated the cask conditions with an 
annual average ambient temperature of 80 deg.F. The use of an annual 
average ambient temperature is used in conjunction with the method 
described in a Pacific Northwest National Laboratory report PNL-6189. 
The method provides one acceptable means for obtaining reasonable 
assurance that the requirements in 10 CFR Part 72 will be met. These 
requirements include protecting the cladding from degradation that 
leads to gross ruptures and designing the storage system to allow ready 
retrieval of the spent fuel or high-level radioactive waste. With 
respect to the 180 deg.F surface temperature in the desert, the SAR 
assumptions used in the 125 deg.F ambient temperature calculation 
credits solar heating (also referred to as solar insolation) and heat 
generated by the casks. Holtec calculated a concrete pad surface 
temperature of 206 deg.F (surrounding the concrete overpack), an 
ambient temperature just above the inlet vent of the overpack of 
136 deg.F, and a concrete temperature at the outlet vent of the 
overpack of 182 deg.F. The NRC staff finds that the Holtec calculation 
adequately models the thermal responses of the cask and its 
environment.

J. Technical Specifications

    Comment J.1: One commenter asked for clarification on the 
conditions for use and the TSs, and if they could be changed without an 
amendment.
    Response: The conditions for cask use are specified in the CoC, and 
includes Appendix A (TSs) and Appendix B (Approved Contents and Design 
Features). These conditions cannot be changed without an amendment to 
the certificate.
    Comment J.2: One commenter stated that the Use and Application 
section of the TSs is confusing and allows too much flexibility for 
completion times and frequencies, and that the TSs should be simple to 
understand and done on time.
    Response: The NRC disagrees with the comment. The Section 1.0, 
``Use and Application'' of the HI-STORM 100 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 HI-STORM 100 Cask System TSs are power reactor 
licensees familiar with the format of the ISTS. The NRC staff believes 
that the format of the HI-STORM 100 TSs will make them easier for 
operators to use and will help to achieve consistency between power 
reactor and spent fuel dry cask storage TSs. The NRC staff disagrees 
that there is too much flexibility for completion times and frequency. 
The NRC staff believes that the specific wording of the TSs clearly 
specifies the allowable time to complete a required action and the 
frequency of any surveillance requirements.

[[Page 25259]]

    Comment J.3: One commenter objected to the use of the term 
``TRANSPORT'' in TS 3.2.2 and indicated that movement to the pad should 
be used because this CoC is for storage only.
    Response: The NRC disagrees with the comment. The term TRANSPORT 
OPERATIONS is specifically defined in Section 1.1 of the Technical 
Specifications and includes all activities involved in moving a loaded 
overpack or transfer cask to and from the ISFSI pad. Further 
clarification of the term is not warranted.
    Comment J.4: One commenter stated that ``Each'' should be in large 
letters in LCO 3.2.2. The commenter also asked why all the removable 
contamination is not removed instead of setting a limit.
    Response: The NRC disagrees with the comment. The capitalization of 
``each'' is consistent with the format of the TSs. As discussed in the 
TS Bases, Section B.3.2.2, the contamination limits for the transfer 
cask are established from guidance in NRC IE Circular 87-01. The limits 
are based on minimum level of activity that can be routinely detected 
under a surface contamination control program using direct survey 
methods. These limits are consistent with levels that prevent the 
spread of contamination to clean areas and are significantly less than 
the levels associated with significant occupational exposure.
    Comment J.5: One commenter stated that the dry run should be 
conducted in sequence and not an alternate step sequence as permitted 
by TS 5.2.
    Response: The NRC disagrees with the comment. The dry runs are 
performed in discrete functional areas to demonstrate the ability to 
perform certain activities as anticipated. The order of performance of 
the functional areas, for the purpose of a dry run, is not directly 
pertinent to a demonstration of a user's capability. The operating 
procedures and technical specifications already control, as necessary, 
functional areas that must be performed sequentially for safe storage. 
The NRC staff considers it important to allow the cask user the 
necessary flexibility to allocate the appropriate resources and 
oversight to the performance of dry runs thatmay involve performing and 
concentrating on certain activities that would be out of sequence with 
a cask loading.
    Comment J.6: One commenter asked why no lifting height limit was 
established for the vertical orientation of the transfer cask in TS 5.5 
and stated that there should be a limit established.
    Response: In the SAR, the design basis drop event analysis is based 
on the horizontal lifting height of 42 inches. Therefore, TS 5.5 only 
specifies the lifting height of the horizontal lifting limit. TS 5.5.c 
permits vertical lifting of loaded transfer cask to any height 
necessary to perform cask handling operations, including the MPC 
transfer. However, the lifts must be made with structures and 
components designed to prevent a drop and in accordance with the 
criteria specified in CoC, Appendix B, Section 3.5 and SAR Section 
2.3.3.1. Therefore, a vertical lift height limit was not established.
    Comment J.7: One commenter asked if the diamond-shaped water rod 
mentioned in note 10 of TS Table 
2.1-3 had been completely analyzed.
    Response: The shape (geometry) of water rods that are part of the 
fuel assembly, is considered in the evaluation.
    Comment J.8: One commenter recommended deleting the words ``For 
OVERPACKS with installed temperature monitoring equipment'' at the 
beginning of the second option under SR 3.1.2.1 because users should 
have the option of using temporary equipment.
    Response: The NRC agrees with the comment. Temperature monitoring, 
a surveillance option permitted in SR 3.1.2.1, could be conducted with 
either temporary or permanently installed equipment. The term 
``installed'' could be interpreted as a requirement that the 
temperature monitoring equipment be permanently fixed. Therefore, the 
beginning of the second option under SR 3.1.2.1 has been reworded as 
follows: ``For OVERPACKS with temperature monitoring equipment'' (i.e., 
the word ``installed'' has been deleted).
    Comment J.9: One commenter recommended several miscellaneous 
editorial changes to the appendices to the CoC.
    Response: The NRC agrees with the comment. The appendices to the 
CoC have been revised to correct typographical errors and incorporate 
minor editorial changes.
    Comment J.10: One commenter recommended that Items 5.2.f and 5.2.j 
in Section 5 of the TSs be revised to insert the phrase ``(for which a 
mock-up may be used)'' at the end of the items for consistency with SAR 
Section 12.2.2.
    Response: The NRC agrees with the comment. Items 5.2.f and 5.2.j of 
the TSs have been revised to indicate that a mock-up may be used for 
those specific dry-run evolutions.
    Comment J.11: One commenter recommended that item 5.5.c in Section 
5 of the TSs be revised to replace the words ``and MPC'' with ``or 
OVERPACK'' because some utilities plan to implement an MPC transfer 
scheme that requires temporary lifting of the loaded OVERPACK above its 
lift height limit.
    Response: The NRC disagrees with the comment. There are no 
evaluations, equipment design criteria, or other information in the SAR 
that support lifting a loaded overpack above its lift height limit.
    Comment J.12: One commenter recommended revising the definitions of 
DAMAGED FUEL ASSEMBLY and PLANAR-AVERAGE INITIAL ENRICHMENT in Section 
1 of Appendix B to the CoC to reflect the evolution of these terms and 
for consistency with those in the HI-STAR 100 CoC.
    Response: The NRC agrees with this comment. The CoC, Appendix B, 
Section 1 has been revised to reflect the new definitions.
    Comment J.13: One commenter recommended revising CoC, Appendix B, 
Section 3.4.6.c to replace the specified yield strength with the 
equivalent ASTM Grade specification.
    Response: The NRC agrees with the comment in part. Storage pad 
design is a site-specific issue that needs to be addressed in the cask 
user's 10 CFR 72.212 evaluation. CoC, Appendix B, Section 3.4.6.c lists 
the design parameters for the storage pads. It is not a list of 
components for fabrication. By using the specific ASTM Grade 
specification as recommended by the commenter, namely, ASTM A615, Grade 
60, the designer of the pad will not have the flexibility to choose 
other reinforcing steels that could also be used (e.g. ASTM A616 or 
A617, Grade 60, etc.). To allow flexibility for the design and still 
ensure adequate reinforcement in the pad CoC, Appendix B, Section 
3.4.6.c has been changed to state that reinforcement shall be 60 ksi 
yield strength ASTM material.
    Comment J.14: One commenter recommended eliminating the requirement 
for impact limiters at the cask transfer facility contained in CoC, 
Appendix B, Item 3.5.2.2.
    Response: The NRC staff assumes that the commenter's reference to 
Section 3.5.2.2 is a typographical error because the requirement for an 
impact limiter is in CoC, Appendix B, Item 3.5.2.1. The NRC agrees in 
part with the comment. The specific requirement for an impact limiter 
has been eliminated from CoC, Appendix B, Section 3.5.2.1.4. The NRC 
determined that this requirement is too restrictive because other 
methods may be available to prevent a canister breach in the event of a 
canister drop during transfer operations. Instead, Item 3.5.2.1.4 has 
been revised to require that

[[Page 25260]]

the CTF be designed, constructed, and evaluated to ensure that if the 
MPC is dropped during an inter-cask transfer operation, its confinement 
boundary would not be breached.
    However, the NRC disagrees with the underlying reason for the 
comment which is: Because a single failure proof crane (or equivalent) 
is required in the CTF, the design features to mitigate the consequence 
of a drop should not be necessary. The NRC staff acknowledges that the 
use of a single-failure proof crane precludes the possibility of a 
heavy load drop event. The requirement for a mitigating feature in the 
CTF design is a defense-in-depth measure that is consistent with the 
overall philosophy and approach of NUREG-0612. This philosophy 
encompasses an intent to prevent as well as to mitigate the 
consequences of postulated accidental load drops. The NRC staff notes 
that, even with a single-failure proof crane, NUREG-0612 still imposes 
a requirement for a safe load travel path ``to minimize the potential 
for heavy loads, if dropped, to impact irradiated fuel in the reactor 
vessel and in the spent fuel pool, or to impact safe shutdown 
equipment.'' The NRC staff views the mitigating feature in the CTF as a 
defense-in-depth measure equivalent to the safe load path. Its function 
is to protect the MPC confinement boundary and the integrity of the 
spent fuel in the MPC in case of a postulated drop.
    Comment J.15: One commenter recommended that CoC, Appendix B, Item 
3.5.2.1.4 be clarified to indicate that the acceptance criterion for 
the impact limiter also applies to the use of mobile cranes.
    Response: The NRC agrees with the comment. As discussed in the 
response to J.14, CoC, Appendix B, Item 3.5.2.1.4 has been revised to 
require that the CTF be designed and evaluated to ensure that if the 
MPC is dropped during an inter-cask transfer operation, its confinement 
boundary would not be breached. Section 3.5.2.1.4 has also been revised 
to specify that this requirement and acceptance criterion apply to both 
stationary and mobile cranes.
    Comment J.16: One commenter recommended that CoC, Appendix B, Item 
3.5.2.1.4 be revised to clarify the scope of drops that require 
evaluation in designing the impact limiter.
    Response: The NRC agrees with the comment. CoC, Appendix B, Item 
3.5.2.1.4 has been revised to clarify that the potential drops that 
require evaluation are those that may occur during inter-cask transfer 
operations.
    Comment J.17: One commenter recommended that the TSs be removed 
from Appendix 12.A of the SAR because they are included in the 
appendices to the CoC.
    Response: The NRC agrees with the comment. The TSs have been 
removed from the SAR.

K. Miscellaneous

    Comment K.1: One commenter expressed approval that movement could 
be conducted at 0 deg.F and above.
    Response: No response is necessary.
    Comment K.2: One commenter stated that the HI-TRAC transfer cask 
must be as safe as the HI-STORM overpack if it is to be used outside 
the reactor security fence.
    Response: The NRC staff reviewed the HI-TRAC transfer cask and 
determined that, like the HI-STORM overpack, it will perform its 
intended safety functions under the design basis normal, off-normal, 
and accident events. It should be noted that the CoC authorizes use of 
the HI-TRAC only within the owner-controlled areas of a licensed power 
reactor.
    Comment K.3: One commenter asked if the inner canister could be 
dropped through, if water could spill out of the overpack, and if the 
water helped to disperse the fuel particles.
    Response: During a canister transfer operation, the transfer cask 
is placed on top of the storage overpack. The canister is then lowered 
through the bottom of the transfer cask into the overpack. It is 
unlikely that a canister drop would occur during this operation because 
the canister must be lifted with equipment (i.e., a single failure 
proof crane or equivalent) that are designed to prevent a drop. In 
addition, the overpack contains only traces of water that is part of 
the concrete material and the canister is dry during cask transfer 
operations.
    Comment K.4: One commenter questioned the assumption that the HI-
TRAC remains static because there are a number of man-made or natural 
causes that could put it in motion, drop, tipover, roll, etc.
    Response: The HI-TRAC is required to be independently secured on 
top of the overpack during the transfer of the MPC.
    Comment K.5: One commenter asked when the measuring equipment (for 
checking tolerances) is calibrated.
    Response: The timing of calibration at the fabricator's facility is 
beyond the scope of this rule. However, the implemented QA program at 
the fabricator's facility provides reasonable assurance that the 
measuring equipment for checking tolerances of fabrication will be 
appropriately calibrated.
    Comment K.6: One commenter asked if the restraint of 11 inches in 
vertical height for overpack handling would actually preclude a corner 
drop situation. The commenter asked how a corner drop could be 
initiated, such as a defective trunnion or lifting lug, etc.
    Response: The 11-inch restriction on lifting height for the 
overpack was calculated to ensure that deceleration loading to the 
loaded MPC would not exceed the design criteria for the confinement 
boundary of the basket. A tipover of the overpack cannot occur if the 
baseplate is limited to 11 inches above a receiving surface.
    Comment K.7: One commenter asked what happens to the inside of the 
cask during a horizontal drop of 50 inches.
    Response: The effect of a 50-inch horizontal drop of a cask was not 
evaluated because the horizontal lifting height limit for the transfer 
cask is 42 inches. The 50-inch carry height specified in the SER was a 
typographical error and has been corrected to 42 inches. There is no 
effect on the confinement function of the MPC as a result of a 
horizontal drop of 42 inches. The structural evaluation shows that all 
stresses are within allowable values and that the confinement boundary 
integrity of the MPC is not impaired.
    Comment K.8: One commenter requested that the SER define what is 
meant by cladding oxide thickness on page 4-1 of the SER.
    Response: The NRC disagrees with the comment. Cladding oxide 
thickness is a measure of corrosion at the clad surface. As water 
interacts with the zirconium clad, the zirconium can interact with the 
oxygen molecules to create zirconium oxide (ZrO2). The 
terminology is commonly used in the spent fuel storage arena and a 
definition in the SER is not necessary.
    Comment K.9: One commenter asked why the internal rod pressure is 
assumed to remain the same. The commenter asked how the gas behaves in 
a dry cask and if it can leak from pinhole leaks and hairline cracks 
over the storage period. The commenter further asked how the lower 
pressure in the rods affects the analysis and heat transfer.
    Response: The internal rod pressure is derived from the initial gas 
inserted during fabrication plus the fission product gases that develop 
during power production within the reactor core. In a closed system 
(e.g., the fuel pin), the pressure is a function of the gases in the 
fuel rod and the average temperature of the gas. As the decay heat 
decreases with time, so does the temperature and the pressure.

[[Page 25261]]

Therefore, the rod temperature does not remain the same. This is 
similar to inflating a balloon with hot air and placing the balloon in 
the refrigerator. As the gases cool, the pressure decreases, as is 
implied by the smaller diameter of the balloon. The lower pressure 
reduces the stress on the cladding and permits a higher allowable 
temperature limit. If the rod experiences a pinhole leak or a hairline 
crack, the gases inside the rod will mix with the helium gas in the 
cask and reduce the internal pressure within the rod.
    Reduction of the internal fuel rod pressure results in added 
assurance that the cladding will remain stable because the internal 
pressure will have equilibrated with that of the cask. The gases from 
the fuel pin mix with the gases in the cask and decreases the thermal 
conductivity of the helium, while at the same time increasing the 
density of the gas. The analyses for accident conditions incorporate 
the impact of reduced conductivity of the helium gases. This impact is 
reduced when crediting cooling that results from natural circulation of 
the gases inside the cask. The use of a maximum allowable temperature 
limit provides assurance that the fuel pins will remain intact 
throughout the storage period. For conservatism, the applicant assumed 
that 1 percent of the cladding experiences a leak under normal 
conditions, a 10-percent leak under off-normal conditions, and a 100-
percent leak under accident conditions.
    Comment K.10: One commenter asked what cask design was tested at 
INEEL (page 4-3 of the SER).
    Response: Several full scale cask designs were tested at the Idaho 
National Engineering and Environmental Laboratory. The cask used by 
Holtec to validate the FLUENT computer code was the TN-24P. The heat 
output of the cask was 23 kW. The NRC staff found the FLUENT computer 
code acceptable for calculating the thermal response of a spent fuel 
cask.
    Comment K.11: One commenter expressed concern over water and debris 
going into cracks on the pad and then freezing and thawing causing 
concrete upheaval and subsequent cask tipover.
    Response: Issues related to cask storage pad will be addressed in 
the cask user's evaluation under 10 CFR 72.212 and is beyond the scope 
of this rule.
    Comment K.12: One commenter asked how moisture and pollution in the 
air could affect the casks and pad over time and if the pad would ever 
need to be replaced.
    Response: The cask can withstand the ambient environmental 
conditions over its 20-year license period with no significant 
degradation. The adequacy of the pad must be addressed by the cask 
users in their 10 CFR 72.212 evaluation and is beyond the scope of this 
rule. Cask users are responsible for inspecting and maintaining the 
pad. With appropriate maintenance, air pollution or moisture would not 
cause significant degradation to the pad.
    Comment K.13: One commenter asked if both the helium and fission 
gases created the pressure inside the rods and for an explanation of 
the fission gases. The commenter also asked why only 30 percent of the 
fission product gas was assumed to be released instead of 100 percent 
because over time 100 percent would likely leak out.
    Response: Fission gases are byproducts of uranium splitting in a 
reactor. These include gases such as hydrogen, krypton, and iodine. The 
gases are contained inside the fuel rod. Data have shown that a 
conservative estimate of 30 percent of the gases generated inside the 
fuel pellet can escape to the gap that exists between the fuel pellet 
and the cladding. This is a conservatively large number used for 
calculating dosage. Experimental data has shown this number to be 
significantly less. The rest of the gases are trapped inside the fuel 
pellet. Therefore, assuming that 100 percent of the gases are released 
from the fuel pellet is not realistic. Helium gas is added to the MPC 
to keep the environment inside the cask inert so it does not promote 
corrosion and to help cool the fuel by transferring heat from the fuel 
rods to the wall of the cask. The impact of helium gas on the pressure 
within a fuel rod is not as significant as the temperature of the gas 
within the fuel rod.
    Comment K.14: One commenter asked what is in the water of the water 
jacket and if the water could affect the carbon steel channels or get 
into the pool through a weld crack or leak and affect the pool. The 
commenter also asked how hot the water and the lead get, and if the 
water could cause pressure buildup in the channels.
    Response: The water used in the water jacket is demineralized water 
as is used in the loading pool, but without boron addition because the 
boron is unnecessary for loading/unloading operations. Carbon steel 
corrodes very slowly in demineralized water; thus, its effect may be 
ignored for the durations experienced in loading operations. If the 
cask is to be loaded in cold weather, antifreeze may be added to the 
jacket water. Antifreeze contains an inhibitor to prevent corrosion. 
There would be no significant effect if the jacket water or water with 
antifreeze leaked into the pool. With regard to the water and lead 
temperatures and pressure buildup in the water jackets, see the 
response to comment I.2.
    Comment K.15: One commenter asked if there was a recent study on 
cladding degradation from creep cavitation.
    Response: Studies on cladding degradation were performed several 
years ago. These studies led to the development of analytic methods to 
calculate the maximum allowable peak clad temperature limits. A report 
developed by the Pacific Northwest National Laboratory (PNL) in May 
1987, PNL-6189, ``Recommended Temperature Limits for Dry Storage of 
Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas'' 
provides an acceptable method for assessing cladding temperature 
limits.
    Comment K.16: One commenter stated that the 100-ton transfer cask 
should not be included in the certification because it is site-specific 
and not made the same as the 125-ton cask.
    Response: NRC disagrees with the comment. The 100-ton and 125-ton 
transfer cask designs have been evaluated and found to meet the 
regulatory requirements of 10 CFR Part 72. The 100-ton transfer cask 
design is not considered site-specific and is approved under this rule 
for use by any general licensee as part of the HI-STROM 100 system as 
described in the SAR. Section 2.0.3 of the SAR provides guidance 
regarding site-specific ALARA objectives that should be considered by 
each user when using either transfer cask design.
    Comment K.17: One commenter asked what does reasonable assurance 
mean in Section 5.1.2 of the SER regarding acceptability of the 
shielding design criteria.
    Response: The finding in Section 5.1.2 is intended to mean that the 
NRC staff believes that the dose rate criteria presented in the SAR are 
acceptable values and that a cask system operating at these values can 
meet the applicable radiological requirements of 10 CFR Parts 20 and 
72. The SAR subsequently demonstrates that the dose rates calculated 
for the HI-STORM system meet the regulatory requirements.
    Comment K.18: One commenter asked if the MOX (mixed oxide) fuel was 
covered by the sabotage report. The commenter asked if MOX fuel had 
been tested and verified to be safe for this design. The commenter 
further questioned how the NRC could include MOX fuel in the SER 
evaluation and stated that storage of MOX fuel should

[[Page 25262]]

not be allowed by the certification. The commenter also asked how we 
know that storage of MOX fuel will work as expected because it has not 
yet been tested in Canada.
    Response: The sabotage report is beyond the scope of this rule. 
However, the design and physical characteristics of a MOX fuel assembly 
are very similar to those of a uranium fuel assembly. The primary 
difference is the fuel pellet constituents and its effects on the 
radiological source term. Testing of MOX fuel is also beyond the scope 
of this rule.
    The HI-STORM design was evaluated for storage of the MOX fuel 
assemblies listed in the Appendix B to the CoC using computer codes and 
models. In lieu of testing, the NRC finds analytic conclusions that are 
based on sound engineering methods and practices to be acceptable. 
Testing is only required if the analytic methods have not been 
validated or assured to be appropriate and/or conservative. The NRC 
staff reviewed the applicant's analyses and found them acceptable. The 
basis of the safety review and findings are identified in the SER and 
the CoC.
    Comment K.19: One commenter asked if all the analysis was based on 
the 100-ton transfer cask or did HI-STORM 100 refer to something else.
    Response: The shielding analysis presented in the SAR evaluated 
both the 100-ton and 125-ton transfer cask designs as part of the HI-
STORM 100 cask system.
    Comment K.20: One commenter asked how the NRC could base its 
evaluation on historical statements when reference documents indicate 
Inconel impurity may be higher than 1000 ppm. The commenter further 
asked what the historical statements were and how we know if the 
statements are valid.
    Response: The applicant's analysis of cobalt impurities are 
discussed in Section 5.2.1 and 5.2.3 of the SER. The applicant showed 
that the cobalt impurity value of 1000 ppm assumed in the shielding 
analyses was appropriate based on industry data and analyses of post-
irradiation cooling of older fuel types that may have had higher cobalt 
impurities for the HI-STORM 100 cask system. As discussed in Section 
5.2.1 of the SAR, historical statements included industry data gathered 
by the applicant from utilities and vendors.
    Cobalt impurities were not necessarily controlled for older fuel 
designs. However, the applicant showed that the post-irradiation 
cooling time that is inherent to these older fuel types significantly 
reduces the HI-STORM 100 dose rates. Therefore, the effects of higher 
impurities are mitigated. Based on historical knowledge of recent 
cobalt reduction programs, the decay effects on older fuel, and its own 
independent evaluations, NRC has reasonable assurance that the 
historical statements referenced in the application are used 
appropriately for the HI-STORM 100. Furthermore, each cask user will 
operate the HI-STORM 100 under a 10 CFR Part 20 radiological protection 
program and will be required to verify dose rates that are specified in 
the TSs. This defense-in-depth approach will mitigate potential 
hardware activation anomalies and ensure compliance with radiological 
requirements.
    Comment K.21: One commenter asked if the steel transport overpacks 
could be reused, how contaminated the overpacks would be after use, the 
number of times an overpack could be reused, and if they would be 
checked after each use.
    Response: This comment that concerns the HI-STAR steel transport 
overpack, is beyond the scope of this rule on the Holtec HI-STORM 100 
cask system.
    Comment K.22: One commenter was pleased that the NRC had evaluated 
uneven flooding.
    Response: No response is necessary.
    Comment K.23: One commenter asked about the chance of one of the 
screens being damaged or loosened in unloading and the debris floating 
out with the cooling water into the pool.
    Response: The damaged fuel container that is placed in the MPC is 
stainless steel and is designed to retain damaged fuel and debris in a 
safe configuration under all normal, off-normal, and accident 
conditions. The damaged fuel container also provides a means to safely 
handle the damaged fuel and debris during loading and unloading. It is 
not considered credible that the screens will fall off or fail. However 
if a screen failed, there would be no release of radioactive material 
during storage since the MPC is sealed. Consideration of loose debris 
during unloading is addressed in SAR Section 8.3 which outlines the MPC 
unloading operations in a spent fuel pool and specifically considers 
loose debris in the MPC. Additionally, the spent fuel pool filtration 
system would capture any debris that remained in the pool.
    Comment K.24: One commenter asked why the volume of water removed 
from the cask is recorded and why this is not done for other cask 
designs.
    Response: The purpose of recording the volume of water removed from 
the canister is to identify the open volume in the canister. This open 
volume is used to calculate the amount of helium to be added to the 
cask following vacuum drying. The procedure and equation used for this 
procedure is discussed on page 8.1-21 in the HI-STORM SAR. The comment 
concerning other cask designs is beyond the scope of this rule.
    Comment K.25: One commenter stated that a detailed procedure on 
mitigating the possibility of fuel crud particulates dispersal should 
be included in the documents and that the procedure should not be site-
specific.
    Response: NRC disagrees with the comment. The generic unloading 
procedures for the HI-STORM 100 system are designed to mitigate crud 
dispersal. However, each cask user will need to develop detailed 
unloading procedures that incorporate the ALARA objectives of its site-
specific radiation protection program. NRC expects the cask user to 
consider the specific characteristics of its fuel, including crud 
phenomena, when developing these procedures.
    Comment K.26: One commenter asked how the utilities are required to 
document that they will not lift the overpack any higher than 11 inches 
and that the receiving surface hardness does not exceed that analyzed 
in the SAR. The commenter stated that the criteria should be clarified 
and which surface should be indicated.
    Response: The receiving surface is the top of the storage pad as 
clearly stated in Sections 3.4.2 and 11.2.3.2 of the SER and described 
in Section 3.4.10 of the SAR. Users of the HI-STORM 100 system are 
required to meet Appendices A and B of the CoC that list the design 
parameters for surface hardness and the restriction for lifting height. 
Furthermore, the cask users are required to develop detailed written 
operating procedures. The restriction on lifting height must be 
incorporated into the operating procedures subject to NRC inspection.
    Comment K.27: One commenter stated that Condition 8 should remain 
in the CoC.
    Response: The NRC agrees with the comment. Condition 8 has not been 
removed from the CoC. Under Condition 8, Certificate holders who wish 
to make changes to the CoC, including Appendices A and B, must submit 
an application for amendment of the Certificate.
    Comment K.28: One commenter asked how upending/downending of the 
transfer cask affected the water in the neutron shield, how the 
licensee knows the shield is full, what happens to the contents of the 
cask when the position changes, what are the stresses and pressures, 
and if the debris in damaged fuel containers goes through the screen.

[[Page 25263]]

    Response: The structural, shielding, and confinement functions of 
the transfer cask are not affected during movement of the cask. The 
neutron shield will normally be filled through the drain valve at the 
bottom of the water jacket and is considered full as water exits the 
vent port at the top of the water jacket. The vent plug is then 
installed to retain the water in the jacket. During the upending and 
downending of the transfer cask, water remains within the neutron 
shield and fuel debris remains within the confinement boundary of the 
MPC. The structural evaluation in the SAR showed all the stresses and 
pressures to remain within allowable values.
    Comment K.29: One commenter stated that exceptions to the codes 
should not be allowed and that the NRC should demand full code 
requirements.
    Response: The NRC disagrees with this comment. Exceptions 
(alternatives) to the ASME Code specifications may be granted by the 
NRC staff on a case-by-case basis. During the NRC staff review of a 
proposed alternative, the applicant must demonstrate that the proposed 
alternative to the Code satisfies one of the following criteria: (1) 
The alternative provides an acceptable level of quality and safety, or, 
(2) compliance with a specific Code requirement would result in 
hardship or unusual difficulty without a compensating increase in the 
level of quality or safety.
    Comment K.30: One commenter stated that videos should not be used 
as a permanent record. The commenter stated that black and white photos 
and negatives should be used and that the negatives should be kept in 
museum qualified storage. The commenter asked what method is best to 
document weld integrity and how the records are stored. The NRC should 
have specific criteria for record keeping requirements.
    Response: The NRC disagrees with the comment. The NRC's regulations 
do not explicitly require specific criteria for record keeping to 
document weld integrity by the applicant. A permanent record of 
completed welds will be made using video, photographic, or other means 
that can provide a retrievable record of weld integrity. As per 
accepted industry practice, the record is typically in color format, in 
order to capture the red dye typically used for PT examinations. The 
general licensee's QA program will specify the types of records and how 
the records are to be stored.
    Comment K.31: One commenter stated that even if the overpack 
baseplates, shell, pedestal shell, and radial plates have large margins 
of safety in the design, they should still be examined to code.
    Response: Holtec has committed to inspect the welds of the overpack 
baseplate to the shell, pedestal shell, and radial plates under ASME 
Code Section V, Article 9. Weld inspection acceptance criteria meet the 
requirements in ASME Section III, Subsection NF-5360.
    Comment K.32: One commenter asked why a mobile lifting device is 
used and why it is not required to meet the requirements of NUREG-0612, 
Section 5.1.6(2) for new cranes. If a new crane is necessary to meet 
the requirements, the utilities should get one and not be allowed to 
lower requirements.
    Response: A mobile lifting device is an alternative option to a 
stationary lifting device that may be used in a CTF. The decision to 
use either a mobile or stationary lifting device would be made by the 
cask users and would be based on their plant's site-specific needs. 
NUREG-0612, Section 5.1.6(2) specifies that new cranes should be 
designed to meet NUREG-0554, ``Single-Failure-Proof Cranes for Nuclear 
Power Plants.'' These requirements are not applicable to mobile lifting 
devices which are not single-failure-proof; therefore, mobile lifting 
devices are exempted from this particular requirement in NUREG-0612. To 
ensure that the mobile lifting device has the equivalent level of 
safety as a single-failure-proof crane, additional conditions in CoC, 
Appendix B, Sections 3.5.2.2.1, 3.5.2.2.2, and 3.5.2.2.4 were imposed.
    Comment K.33: One commenter stated that a discussion on the cask 
transfer facility should be included in the SER, and that the public 
should not have to read the SAR to understand the generic design. The 
commenter requested that this part of the cask transfer facility be 
resubmitted with a complete clear design with specific criteria.
    Response: The NRC disagrees with the comment. SER Section 1.1 
discusses the CTF in a level of detail appropriate for an SER. The 
detailed design and operating criteria for the CTF are given in SAR 
Section 2.3.3.1. This satisfies 10 CFR 72.24, which requires that the 
SAR contain information on structures, systems, and components 
important to safety in sufficient detail for the NRC staff to make its 
regulatory finding. Repeating this information in the SER is not 
necessary. The NRC disagrees that cask transfer facility should be 
resubmitted with a complete clear design with specific criteria. The 
specific criteria for the CTF are already given in CoC, Appendix B, 
Section 3.4, and SAR Section 2.3.3.1. As discussed in the response to 
E.30, NRC found it unnecessary to approve a specific CTF design.
    Comment K.34: One commenter recommended that Section 3.5.7 of the 
SER be revised to reflect that transport of the HI-TRAC transfer cask 
in the vertical orientation is permitted. The comment also recommended 
that ``50 inches'' be changed to ``42 inches'' to be consistent with TS 
Table 5-1.
    Response: The NRC agrees with the comment. The SER has been 
modified to reflect that transport of the HI-TRAC transfer cask in the 
vertical orientation is permitted. The horizontal lifting height per TS 
Table 5-1 will be corrected to 42 inches to correct the typographical 
error.
    Comment K.35: One commenter recommended that Section 9.1.2.2.b of 
the SER be revised to delete ``(either to the fuel pool or the site 
licensee's off-gas system)'' because users may or may not have these 
systems at their plants.
    Response: The NRC agrees with the comment. It is up to the cask 
users to develop the specific procedures for venting the MPC and to 
determine the appropriate location under their plant's waste gas 
handling system design and radiation protection program. Section 
9.1.2.2.b of the SER has been modified as recommended.

Summary of Final Revisions

    As a result of the NRC staff's response to public comments, or to 
rectify issues identified during the comment period, TSs 5.2.f and 
5.2.j have been modified (see comment J.10). The NRC staff has also 
updated the CoC, including Appendix B, and has removed the bases 
section from the TSs attached to the CoC to ensure consistency with 
NRC's format and content. The NRC staff has also modified its SER. In 
addition, the NRC staff has modified the rule language by changing the 
word ``Certification'' to ``Certificate'' to clarify that it is 
actually 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

[[Page 25264]]

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 Holtec 
International HI-STORM 100 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 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 
Holtec International. 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

[[Page 25265]]

rule. Therefore, a backfit analysis is not required.

List of Subjects in 10 CFR Part 72

    Administrative practice and procedure, Hazardous waste, Nuclear 
materials, Occupational safety and health, Penalities, Radiation 
protection, Reporting and recordkeeping requirements, Security 
measures, Spent fuel, Whistleblowing.
    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553; the NRC is adopting the 
following amendments to 10 CFR part 72.

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

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

    Authority:  Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 10d-
48b, sec. 7902, 10b Stat. 31b3 (42 U.S.C. 5851); sec. 102, Pub. L. 
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, 
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 
10155, 10157, 10161, 10168).
    Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), 
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 
10168(c),(d)). Section 72.46 also issued under sec. 189, 68 Stat. 
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub. 
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also 
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2244, (42 U.S.C. 10101, 
10137(a), 10161(h)). Subparts K and L are also issued under sec. 
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 
(42 U.S.C. 10198).


    2. In Section 72.214, Certificate of Compliance 1014 is added to 
read as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1014.
    SAR Submitted by: Holtec International.
    SAR Title: Final Safety Analysis Report for the HI-STORM 100 Cask 
System.
    Docket Number: 72-1014.
    Certificate Expiration Date: June 1, 2020.
    Model Number: HI-STORM 100.

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

    For the Nuclear Regulatory Commission.
Frank J. Miraglia, Jr.,
Acting Executive Director for Operations.
[FR Doc. 00-10393 Filed 4-28-00; 8:45 am]
BILLING CODE 7590-01-P