[Federal Register Volume 64, Number 171 (Friday, September 3, 1999)]
[Rules and Regulations]
[Pages 48259-48274]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-23075]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG17


List of Approved Spent Fuel Storage Casks: (HI-STAR 100) Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to add the Holtec International HI-STAR 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 October 4, 1999.

FOR FURTHER INFORMATION CONTACT: Stan Turel, telephone (301) 415-6234, 
e-mail [email protected] of the Office of Nuclear Material Safety and 
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.

SUPPLEMENTARY INFORMATION:

[[Page 48260]]

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 International HI-STAR 100 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 together 
with the Safety Analysis Report (SAR) entitled ``HI-STAR 100 Cask 
System Topical Safety Analysis Report (SAR), Revision 8.'' 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 International HI-STAR 100 cask system. The NRC 
published a proposed rule in the Federal Register (64 FR 1542; January 
11, 1999) to add the HI-STAR 100 cask system to the listing in 10 CFR 
72.214. The comment period ended on March 29, 1999. Nine comment 
letters were received on the proposed rule.
    Based on NRC review and analysis of public comments, the 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 International HI-STAR 100 cask system. 
The staff has also modified its preliminary SER and has revised the 
title of the SAR in the listing of this cask design in 10 CFR 72.214.
    The title of the SAR has been revised to delete the revision number 
so that in the final rule the title of the SAR is ``HI-STAR 100 Cask 
System Topical Safety Analysis Report.'' This revision conforms the 
title to the requirements of new 10 CFR 72.248, recently approved by 
the Commission.
    The proposed CoC has been revised to clarify the requirements for 
making changes to the CoC by specifying that the CoC holder must submit 
an application for an amendment to the certificate if a change to the 
CoC, including its appendices, is desired. This revision conforms the 
change process to that specified in 10 CFR 72.48, as recently approved 
by the Commission. The CoC has also been revised to delete the proposed 
exemption from the requirements of 10 CFR 72.124(b) because a recent 
amendment of this regulation makes the exemption unnecessary (64 FR 
33178; June 22, 1999). In addition, other minor, nontechnical, changes 
have been made to CoC 1008 to ensure consistency with NRC's new 
standard format and content for CoCs. Finally, extensive comments were 
received from Holtec International and 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. Staff agrees with many of these suggested changes and 
has incorporated them into the final documents, as appropriate.
    The NRC finds that the Holtec International HI-STAR 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 International HI-STAR 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 International HI-STAR 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 
October 4, 1999. 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 nine comment letters on the proposed rule. The 
commenters included the applicant, the State of Utah, an individual 
member of the public, industry representatives, and several utilities. 
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 Direct Final Rule

    As part of the proposed rule, the NRC staff requested public 
comment on the use of a direct final rulemaking process for future 
amendments to the list of approved spent fuel storage casks in 10 CFR 
72.214. The direct final rulemaking process is used by Federal 
agencies, including the Environmental Protection Agency (EPA) and the 
NRC, to expedite rulemaking where the agency believes that the rule is 
noncontroversial and significant adverse comments will not be received. 
Use of this technique in appropriate circumstances has been endorsed by 
the Administrative Conference of the United States (60 FR 43110; August 
18, 1995). Under the direct final rulemaking procedure, the NRC would 
publish the proposed amendment to the 10 CFR 72.214 list as both a 
proposed and a final rule in the Federal Register simultaneously. A 
direct final rule normally becomes effective 75 days after publication 
in the Federal Register unless the NRC receives significant adverse 
comments on the direct final rule within 30 days after publication. If 
significant adverse comments are received, the NRC publishes a document 
that withdraws the direct final rule. The NRC then addresses the 
comments received as comments on the proposed rule and subsequently 
issues a final rule.
    One commenter supported use of the direct final rule process for 
future revisions to the listing in 10 CFR 72.214, stating that it was 
imperative that the regulatory process be streamlined when there is no 
adverse safety concern. Two commenters were opposed to use of a direct 
final rule process stating that a direct final rule would diminish the 
public role in commenting on the approval of spent nuclear fuel casks 
and thereby the public's ability to affect the outcome of rulemaking 
procedures. One of these commenters believed that, given past problems 
with the casks, future approval should be subject to adequate and 
rigorous public scrutiny.

[[Page 48261]]

Those opposed also believed that 30 days (as would be allowed in a 
direct final rule process) is not sufficient time to prepare comments 
that may be significantly adverse so as to cause the NRC to withdraw 
the published final rule. The two commenters did not believe that an 
addition to or revision of the listing is likely to be either 
noncontroversial or routine as evidenced by the number of comments they 
had on the Holtec HI-STAR 100 proposed rule.
    A number of significant adverse comments were received on the NRC's 
proposed listing of the Holtec International HI-STAR 100 cask system 
which are described in subsequent sections of this notice. Therefore, 
it does not appear that the direct final rule approach can be 
implemented at this time for additions to the cask listing. The NRC 
will reassess this issue in the future after experience with more new 
listings to 10 CFR 72.214 has been gained. However, with respect to 
amendments to existing CoCs, the NRC anticipates that, except in 
unusual cases, the direct final rulemaking process can be used because 
the cask design and analysis will have gone through the public comment 
process for the initial CoC listing and the revision will be limited to 
the subject of the amendment. Unless the NRC has reason to believe that 
a particular amendment will be controversial, the NRC plans to use a 
direct final rule for amendments to the cask systems in the 10 CFR 
72.214 listing. The NRC disagrees that use of the direct final 
rulemaking procedure will limit the public's ability to affect the 
outcome of the rulemaking. Receipt of a significant adverse comment 
will cause the direct final rule to be withdrawn and the comment to be 
considered as though received in response to a proposed rule. Further, 
the NRC believes that 30 days is a sufficient amount of time in which 
to submit a comment on an amendment to the CoC for a listed cask since 
most issues related to the cask design will have been resolved in the 
rulemaking conducted to place the design on the 10 CFR 72.214 list.

Comments on the Holtec International HI-STAR 100 Cask System

    The comments and responses have been grouped into five areas: 
general comments, cladding integrity, health impacts, sabotage events, 
thermal requirements, and miscellaneous items. Several of the 
commenters provided specific comments on the draft CoC, the NRC staff's 
preliminary SER, the TSs, and the applicant's Topical SAR. Some of the 
editorial comments have been grouped as well as some of the comments on 
the drawings in the SAR. To the extent possible, all of the comments on 
a particular subject are grouped together. The listing of the Holtec 
International HI-STAR 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:

General Comments

    Comment No. 1: One commenter asked a number of questions about the 
process for review and approval of spent fuel storage cask designs, and 
suggested changes to the process.
    Response: The NRC finds these comments to be beyond the scope of 
the current rulemaking which is focused solely on whether to place a 
particular cask design, the Holtec International HI-STAR 100 cask 
system, on the 10 CFR 72.214 list.
    Comment No. 2: One commenter stated that the cask should be built 
and tested before use at reactors, including the loading and unloading 
procedures. The commenter objected to the use of computer modeling and 
analysis.
    Response: The NRC disagrees with the comment. The HI-STAR 100 
Storage Cask System Design has been reviewed by the NRC. The basis of 
the safety review and findings are clearly identified in the SER and 
CoC. Testing is normally required when the analytic methods have not 
been validated or assured to be appropriate and/or conservative. In 
place of testing, the NRC staff finds acceptable analytic conclusions 
that are based on sound engineering methods and practices. NRC accepts 
the use of computer modeling codes to analyze cask performance. The 
appropriateness of the computer codes and models used by Holtec are 
addressed in the SER and Topical SAR. The NRC staff has reviewed the 
analyses performed by HOLTEC and found them acceptable. No changes to 
the CoC, TSs, SER, or Topical SAR are recommended. These models are 
based on sound engineering sciences and processes.
    Comment No. 3: One commenter requested that a troubleshooting 
manual be prepared that includes information on how many of what type 
cask are loaded, where and how long they have been loaded, and on 
problems that have occurred, and the solutions. The commenter is 
seeking basic information that is periodically updated.
    Response: This comment is beyond the scope of this rulemaking.

Cladding Integrity

    Comment No. 4: One commenter noted that Holtec's conclusion that 
fuel rod integrity will be maintained under all accident conditions is 
based on the fact that the HI-STAR 100 system is designed to withstand 
a maximum deceleration of 60 g, while a Lawrence Livermore National 
Laboratory Report (UCID-21246, Dynamic Impact Effects on Spent Fuel 
Assemblies, Chum, Witt, Schwartz (October 20, 1987)) (LLNL Report) 
shows that the most vulnerable fuel can withstand a deceleration of 63 
g in the most adverse orientation (side drop). The commenter believes 
that Holtec and the NRC staff have not demonstrated a reasonable 
assurance that the cladding will maintain its integrity because 
Holtec's analysis does not take into account the possible increase in 
rate of oxidation of cladding of high burnup fuel, and oxidation may 
cause the cladding to become effectively thinner, decreasing its 
structural integrity and lowering the ``g'' impact force at which fuel 
cladding will shatter. With respect to a possible increase in rate of 
oxidation of cladding, Holtec has not factored the information in 
Information Notice (IN) 98-29, ``Predicted Increase in Fuel Rod 
Cladding Oxidation'' (August 3, 1998) into its calculations. The clear 
implication of IN 98-29, in the commenter's view, is that the lift 
height of the HI-STAR 100 cask must be reduced to lower the ``g'' 
impact forces on the cladding. Also, the commenter provided a table, 
``Effects of Changing Variables in Dynamic Impact Effects on Spent Fuel 
Assemblies,'' which the commenter believes shows that the maximum ``g'' 
impact force, that high burnup fuel with oxidized cladding can 
withstand, approaches 45 g.
    Response: The NRC disagrees with the comment. Information Notice 
98-29 states that high burn-up conditions may increase fuel rod 
cladding oxidation. The increased rate of oxidation is a function of 
the fuel burn-up and will only affect cladding in high burn-up fuel 
applications. In general, fuel with a burn-up exceeding 45,000 MWD/MTU 
is considered to be a high burn-up fuel. However, the Holtec HI-STAR 
100 Storage Cask System is not authorized to contain fuel with a burn-
up exceeding 45,000 MWD/MTU. Fuel cooling and the average burn-up 
approved for the HI-STAR 100 Storage Cask System is: (a) for MPC-24 PWR 
assemblies, the fuel burn-up is limited to 42,100 MWD/MTU; and (b) for 
MPC-68 BWR assemblies, the fuel burn-up is limited to 37,600 MWD/MTU. 
Therefore, the potential for significant amounts of

[[Page 48262]]

oxidized cladding is not a concern for the HI-STAR 100 Storage Cask 
System, and the table provided by the commenter regarding the 
consequences of significantly oxidized fuel cladding is not relevant to 
the approved contents of this cask design.
    Comment No. 5: The same commenter stated that Holtec's SAR for the 
HI-STAR 100 storage cask relies upon the LLNL report for its estimate 
of ``g'' impact force that will damage fuel cladding but that the LLNL 
report fails to take into account the increased brittleness of 
irradiated fuel assemblies. Because the irradiated fuel assemblies may 
have been embrittled, they would also be less resistant to impact. 
During the course of a fuel assembly's life, subatomic particle 
bombardment, including neutron flux, significantly decreases the 
assembly's ductility and increases the assembly's yield stress, thereby 
embrittling the fuel assembly.
    The HI-STAR 100 design cannot rely on LLNL's analysis, in the 
commenter's view, because the LLNL analysis does not account for 
irradiation and embrittlement, which lower the impact resistance of the 
fuel assemblies. These facts are significant when coupled with the 
increased oxidation rate reported in IN 98-29 because increased 
oxidation could tangentially cause an increase in cladding 
embrittlement. Thus, IN 98-29 compounds the LLNL's error in 
disregarding the brittle characteristics of irradiated fuel cladding.
    Response: The NRC disagrees with the comment. The LLNL Report, as 
referred to, considers the effects of irradiation on cladding. Table 3 
of the report delineates irradiated cladding longitudinal tensile tests 
on coupon specimens. These test specimens were machined from the 
cladding. The effects of irradiation will increase the Young's modulus 
and yield stress but decrease the ductility of the cladding. Figure 5 
of the report shows that the total elongation values for zircaloy do 
not change significantly with strain rate and that the ductility 
appears to be independent of the level of the g-loading. Further, 
Figure 5 of the report shows that the yield strength is consistently 
lower than the tensile strength which suggests that significant margin 
exists between yielding of the cladding and gross rupture. The 
allowable ``g'' impact force calculation in the report is based on the 
yield stress. Thus, the approach that is used in the LLNL Report and 
reflected in the SAR is conservative and acceptable.
    Comment No. 6: The same commenter stated that Holtec's calculations 
rely upon the LLNL report's erroneous assumption that the fuel within 
the cladding behaves as a rigid rod. Thus, Holtec merely used a static 
calculation for impact analysis versus a dynamic calculation. This 
assumption is incorrect, in the view of the commenter. Instead of a 
homogenous, rigid rod, the fuel rod consists of fuel pellets stacked 
like coins within thin tubing. In any impact scenario, the fuel 
assembly acts as a dynamic system with the fuel impacting the inside of 
the cladding and creating a greater likelihood of cladding rupture. 
Holtec has not shown that the assumption of a rigid rod is 
conservative. The thinner cladding due to the increased oxidation 
serves to compound this effect because a smaller ``g'' force would be 
required to rupture the assembly.
    Response: The NRC disagrees with the comment. The assertion that 
the fuel rod consists of fuel pellets stacked like coins within thin 
tubing is incorrect for irradiated fuels. The fuel pellets are densely 
packed inside the fuel tubing, and the effects of irradiation will bond 
the pellets to each other and to the fuel cladding. Samples of 
irradiated fuel rods have shown that it is indeed nearly impossible to 
separate the fuel pellets and the cladding.
    It is incorrect to assume the fuel rod acts as a dynamic system 
with the fuel pellets impacting the inside of the fuel rod cladding 
during an accident drop event. The fuel pellets are densely packed 
inside the fuel tube and, for irradiated fuels, the fuel pellets are 
bonded together and to the cladding. The LLNL Report discussed above 
has conservatively neglected the contributions of the fuel pellets to 
fuel rod rigidity. Rather, the report only considers the cladding for 
calculating the allowable g-load. It is true that the LLNL Report used 
static calculations to derive the allowable g-load equivalent to the 
dynamic impact loading. During an accident drop event, the fuel 
assembly is subjected to dynamic impact loading and the equivalent 
static g-load is determined by a dynamic analysis. The equivalent 
static g-load is then shown to be lower than the allowable g-load to 
ensure the fuel cladding integrity is maintained. The approach is well 
established and acceptable. Therefore, the NRC staff has found Holtec's 
accident analysis to be conservative as reflected in SER Chapter 11 and 
is therefore acceptable.
    Comment No. 7: One commenter stated that the calculated health 
impacts under hypothetical accident conditions discussed in Chapter 7 
of Holtec's HI-STAR 100 SAR are not 100 percent conservative. Holtec's 
original hypothetical design basis accident condition assumed that 100 
percent of the fuel rods are nonmechanically ruptured and that the 
gases and particulates in the fuel rod gap between the cladding and 
fuel pellet are released to the multi-purpose canister (MPC) cavity and 
then to the external environment. The accident analysis in the final 
version increased the amount of radioactivity to the MPC cavity by 5 
orders of magnitude in accordance with NUREG-1536, and would have 
placed doses at 100 m over the EPA's limit of 5 rem. An assumed small 
leakage rate by the applicant reduced the amount released from the cask 
cavity to the environment by more than 5 orders of magnitude. This 
design basis accident no longer represents a loss-of-confinement-
barrier accident as originally described.
    Response: The NRC disagrees with the comment. The hypothetical 
accident dose calculation is appropriate. As discussed in Interim Staff 
Guidance (ISG)-5, Rev. 1, ``Normal, Off-Normal, and Hypothetical 
Accident Dose Estimate Calculations for the Whole Body, Thyroid, and 
Skin,'' the hypothetical accident assumes 100 percent fuel rod failure 
within the MPC cavity and release of radioactivity based on factors 
from NUREG/CR-6487. The applicant demonstrated that the HI-STAR 100 
confinement boundary (MPC) remains intact from all credible accidents. 
Therefore, there is not a credible loss-of-confinement-barrier accident 
for the HI-STAR 100. The hypothetical accident leakage is 
conservatively assumed to be equal to that assumed for normal condition 
leakage with corrections for accident pressures and temperatures. The 
normal condition leak rate is specified in TS 2.1.1.
    The NRC believes that there is reasonable assurance that the 
confinement design is adequately rigorous and will remain intact under 
the normal and accident conditions identified by the applicant. 
Therefore, the design basis change has been found to be conservative 
and meets applicable regulations.
    Comment No. 8: One commenter requested the criteria for an intact 
fuel assembly, the number of pinhole leaks, blisters, hairline cracks, 
and crud. The commenter asked if a visual inspection is required and 
stated that just performing visual exam was inadequate.
    Response: As proof that the fuel to be loaded is undamaged, the NRC 
will accept, as a minimum, a review of the records to verify that the 
fuel is undamaged, followed by an external visual examination of the 
fuel assembly before loading to identify any obvious damage. For fuel 
assemblies where

[[Page 48263]]

reactor records are not available, the level of proof will be evaluated 
on a case-by-case basis. The purpose of this demonstration is to 
provide reasonable assurance that the fuel is undamaged or that damaged 
fuel loaded in a storage or transportation cask is confined (canned). 
The criteria for intact assembly are defined in TS Section 1.1 as being 
fuel assemblies without known or suspected cladding defects greater 
than pinhole leaks or hairline cracks and which can be handled by 
normal means. Partial fuel assemblies (fuel assemblies from which fuel 
rods are missing) shall not be classified as intact fuel assemblies 
unless dummy fuel rods are used to displace an amount of water greater 
than or equal to that displaced by the original fuel rods.

Radiation Protection

    Comment No. 9: One commenter stated that Holtec calculated the 
radiation dose to an adult 100 meters from the accident due solely to 
inhalation of the passing cloud without considering other relevant 
pathways, such as direct radiation from cesium and cobalt-60 deposited 
on the ground, resuspension of deposited radionuclides, ingestion of 
contaminated food and water, and incidental soil ingestion, and does 
not reflect 10 CFR 72.24(m).
    Response: The NRC agrees that Holtec calculated the radiation dose 
to an adult 100 meters from the accident due solely to inhalation of 
the passing cloud and did not consider direct radiation and ingestion. 
The NRC staff considers inhalation to be the principal pathway for 
radiation dose to the public, and Holtec has followed NRC staff 
guidance in making conservative assumptions regarding the source term 
and duration of the release. In SER Chapter 10, the NRC staff found 
that the radiation shielding and confinement features of the cask 
design are sufficient to meet the radiation protection requirements of 
10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106. Section 72.106 
addresses postaccident dose limits.
    When a general licensee uses the cask design, it will review its 
emergency plan for effectiveness in accordance with 10 CFR 72.212. This 
review will consider interdiction and remedial actions to monitor 
releases and pathways based on the chosen site conditions and the 
location. Therefore, the pathways identified by the commenter will be 
addressed in the general licensee's site specific review.
    Comment No. 10: One commenter stated that Holtec has not 
specifically calculated potential radiation dose to children, and this 
does not meet NRC regulations. Further, the commenter stated that NRC's 
methodology for calculating the potential dose to children is 
deficient.
    Response: The NRC disagrees with the comments. While Holtec did not 
specifically calculate potential radiation dose to children, the 
international community and the Federal agencies (including EPA and the 
NRC) agree that the overall annual public dose limit, from all sources, 
should be 1 mSv (100 mrem) which is protective of all individuals. The 
purpose of the public dose limit is to limit the lifetime risk from 
radiation to a member of the general public. Variation of the 
sensitivity to radiation with age and gender is built into the 
standards which are based on a lifetime exposure. A lifetime exposure 
includes all stages of life, from birth to old age. For ease of 
implementation, the radiation standards, that are developed from the 
lifetime risk, limit the annual exposure that an individual may 
receive. Consequently, the unrestricted release limit of 0.25 mSv (25 
mrem), a small fraction of the annual public dose limit, is protective 
of children as well as other age groups 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 May 21, 1991, 
Federal Register notice (56 FR 23387).
    Comment No. 11: One commenter asked if the streaming dose rates 
have been measured and if not, will they be measured on the first cask 
loading?
    Response: There is no NRC regulatory requirement to measure 
streaming dose rates at the first cask loading. Further, the applicant 
did not provide measured dose rates from cask streaming in its 
application because it was not required. The applicant did provide 
calculated streaming dose rates in the SAR shielding analysis. The HI-
STAR 100 system is designed to eliminate significant streaming paths, 
and each user is required to operate the HI-STAR 100 under a 10 CFR 
Part 20 radiological program. NRC has reasonable assurance that the 
general licensee's radiological protection and ALARA program will 
detect and mitigate exposures from any significant or unexpected 
radiation fields for each cask loading.
    Comment No. 12: One commenter stated that the applicant should have 
performed a specific analysis for off-normal conditions for confinement 
analysis and should have included an ``85K'' (Kr-85) dose 
calculation to the skin.
    Response: The NRC agrees. The applicant should have done an off-
normal condition confinement analysis; however, the off-normal case 
dose is approximately a factor of 10 greater than normal dose. The 
Holtec normal condition results show acceptable doses when the factor 
of 10 is applied for off-normal conditions and have been found 
acceptable as reflected in the SER. No additional action is necessary 
to meet applicable NRC regulations.
    Comment No. 13: One commenter stated that the licensees' report on 
specific site doses to the public should be included in the PDR.
    Response: The dose for a site-specific location is beyond the scope 
of this rulemaking. Licensees are required to meet the dose restriction 
in 10 CFR Part 20.
    Comment No. 14: One commenter asked for a definition of inflatable 
annulus seal. The commenter further questioned the checks and criteria 
for surface contamination.
    Response: The inflatable annulus seal, which is discussed in 
Sections 1.2.2.1, 8.1, and 10.1.4 of the SAR, is designed to prevent 
radionuclide contamination of the exterior MPC while the cask is 
submerged in a contaminated spent fuel pool. The space between the MPC 
and overpack is filled with clean water and is sealed at the top of the 
MPC with the inflatable annulus seal. After the seal is removed, the 
upper accessible portion of the MPC is examined for contamination to 
verify that the seal remained intact during underwater loading. NRC 
found the seal description and operation to be acceptable. Each general 
licensee will develop site-specific operating procedures that address 
the use of the inflatable annulus seal. Each general licensee will also 
operate the HI-STAR 100 under a 10 CFR Part 20 radiological protection 
program.
    Comment No. 15: One commenter suggested that there should be 
criteria for the distance of dose measuring mechanism from the cask and 
personnel during loading and unloading.
    Response: NRC disagrees with this suggestion because NRC 
regulations do not specifically require these criteria for dose 
measurement. Each general licensee is required to operate the HI-STAR 
100 under a 10 CFR Part 20 radiological program and must develop site-
specific operating procedures that include radiological protection dose 
surveys that must be conducted during loading and unloading operations.

[[Page 48264]]

Sabotage Events

    Comment No. 16: One commenter stated that the current sabotage 
design basis is not a bounding accident and that the NRC should 
consider the effect of a sabotage event with an anti-tank missile. 
There is a lack of a comprehensive assessment of the risks of sabotage 
and terrorism against nuclear waste facilities and shipments. The NRC 
staff could impose additional conditions on dry storage casks and 
Independent Spent Fuel Storage Installations (ISFSIs), e.g., the CoC 
could require that an ISFSI be designed with an earthen berm to remove 
the line-of-sight.
    The commenter stated that since the early 1980s, the NRC has relied 
on and poorly interpreted an outdated set of experiments carried out by 
Sandia National Laboratory and Battelle Columbus Laboratories that 
measured the release of radioactive materials as a result of cask 
sabotage. The NRC has never estimated the economic and safety 
implications of a sabotage event at a fixed storage facility. Following 
the publication of these Sandia study results, the NRC proposed 
elimination of a number of safety requirements for shipments of spent 
fuel. At least 32 parties submitted more than 100 pages of comments in 
response to the notice, to which the NRC never publicly responded. The 
NRC suspended action on the rulemaking but inappropriately continues to 
use the unrevised conclusions in the proposed rule as a basis for its 
policies on terrorism and sabotage of nuclear shipments.
    Response: The NRC disagrees with the comment. The NRC reviewed 
potential issues related to possible radiological sabotage of storage 
casks at reactor site ISFSIs in the 1990 rulemaking 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 
the 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 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, and the licensee conducts periodic 
patrols and surveillances to ensure that the physical protection 
systems are operating within their design limits. It is the ISFSI 
licensee who is responsible for protecting spent fuel in the casks from 
sabotage rather than the certificate holder. Comments on the specific 
transportation aspects of the cask system and existing regulations 
specifying what type of sabotage events must be considered are beyond 
the scope of this rulemaking.
    Comment No. 17: One commenter asked whether an evaluation for a 
truck bomb sabotage event has been conducted.
    Response: The staff has evaluated the effects of a truck bomb 
located adjacent to storage casks. Spent fuel in the ISFSI is required 
to be protected against radiological sabotage using provisions and 
requirements as specified in 10 CFR 72.212(b)(5). Each utility licensed 
to have an ISFSI at its reactor site is required to develop physical 
protection plans and install a physical protection system that provides 
high assurance against unauthorized activities that could constitute an 
unreasonable risk to the public health and safety. The physical 
protection systems at an ISFSI and its associated reactor are similar 
in design to ensure the detection and assessment of unauthorized 
activities. Response to intrusion alarms is required. Each ISFSI is 
periodically inspected by NRC, and the licensee conducts periodic 
patrols and surveillances to ensure that security systems are operating 
within their design limits. The NRC believes that the inherent nature 
of the spent fuel and the spent fuel storage cask provides adequate 
protection against a vehicle bomb, and has concluded that there are no 
safety concerns outside the controlled area.

Thermal Requirements

    Comment No. 18: One commenter stated that the CoC temperature 
limits for the storage cask are deficient because they do not take into 
account a minimum pitch or center-to-center distance between casks to 
be stored in the ISFSI. Further, Holtec has not performed rigorous 
calculations to support the assigned pitch of 12-foot or 4-foot spacing 
between casks based on the amount of detail in its nonproprietary 
version of its analyses.
    Response: The NRC disagrees with the comment. In Section 4.4.1.1.7 
of the SAR, Holtec addressed the heat transfer interaction between the 
overpacks for a cask array at an ISFSI site. No forced convection was 
assumed (e.g. stagnant ambient conditions which would maximize the 
interaction heat effect). The applicant further adjusted the heat 
transfer in accordance with ANSYS methodology and applied it in the 
calculations. Further, in SER Section 4.5.2.1, the NRC staff noted that 
the applicant considered in its temperature calculations that multi-
purpose cask baskets were loaded at design basis maximum heat loads, 
and systems were considered to be arranged in an ISFSI array and 
subjected to design basis normal ambient conditions with insulation. 
The NRC staff concluded in the SER that it has reasonable assurance 
that the spent fuel cladding will be protected against degradation by 
maintaining the clad temperature below maximum allowable limits.

Miscellaneous Items

    Comment No. 19: One commenter asked why a coating without zinc was 
not required for the VSC-24 cask design. The commenter further 
questioned why NRC allowed coatings to be applied to casks because it 
will create problems for future DOE waste disposal.
    Response: NRC regulations do not prohibit the use of coatings in a 
cask design. An applicant must provide information in its safety 
analysis report to support use of coatings. The applicant should 
describe the near and long term effects of the coatings on systems 
important to safety including the benefits and potential impacts of 
coating use. Based on the applicant's analysis, the NRC reviews and 
assesses the use and adequacy of the coatings. Specific comments 
relating directly to VSC-24 are beyond the scope of this rulemaking.
    Comment No. 20: One commenter asked why the current HI-STAR 100 is 
not an ASME stamped component.
    Response: NRC regulations do not require an ASME stamp for a cask. 
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.'' Applicant submittals are reviewed to the criteria in 
the Standard Review Plan. Cask fabrication activities are inspected by 
the licensees and the NRC staff to

[[Page 48265]]

ensure that components are fabricated as designed.
    Comment No. 21: One commenter asked a number of questions related 
to the Boral and NS-4-FR concerning (1) Whether it has been used ``over 
time'' in a cask, (2) the amount of ``creep or slump'' that has 
occurred over time, (3) how the testing is conducted, and (4) how the 
Boral content is tested in the panels. The commenter further asked if 
fabrication is inspected and why no surveillance or monitoring program 
is required to check the Boral content.
    Response: The questions and comments on the Boral neutron absorber 
are addressed in Sections 6.4.2 and 9.1.4 of the SER and Sections 
1.2.1.3.1, 6.3.2, and 9.1.5.3 of the SAR. The NRC routinely accepts the 
use of Boral as a neutron absorber for storage cask applications, and 
it has been used in casks. NRC has approved both storage and 
transportation cask designs that use Boral. Section 1.2.1.3.1 of the 
SAR describes the historical applications and service experience of 
Boral. This information indicates that Boral has been used since the 
1950's and used in baskets since the1960's. Several utilities have also 
used Boral for nuclear applications such as spent fuel storage racks. 
Based on industry experience, no credible mechanism for ``creep or 
slump'' of Boral in the cask has been identified.
    Sections 1.2.1.3.1 and 9.1.5.3 of the SAR describe the testing 
procedures for Boral. Boral 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. 
A statistical sample of each manufactured lot of Boral is tested by the 
manufacturer using wet chemistry procedures and/or neutron attenuation 
techniques.
    The Boral is designed to remain effective in the HI-STAR 100 system 
for a storage period greater than 20 years and there are no credible 
means to lose the Boral. Further, the NRC accepts the use of NS-4-FR as 
a neutron absorber for storage cask applications, and it has been used 
in other casks. Therefore, surveillance and monitoring are not needed.
    Comment No. 22: One commenter provided a discussion on the VSC-24 
design. The issues included materials, the use of coatings, the use of 
March Metalfab as a fabricator, calculations being performed when 
problems are being solved, testing of soils and pads, and cask handling 
temperatures.
    Response: These comments are beyond the scope of the current 
rulemaking.
    Comment No. 23: One commenter asked how the prepossession or 
anodization of aluminum surfaces is checked and what the criteria were 
for the inspection.
    Response: The NRC disagrees that an inspection is necessary. The 
only aluminum used in the MPC-24 or MPC-68 is for the Boral neutron 
absorbers. Aluminum forms a very thin, adherent film of aluminum oxide 
whenever a fresh cut surface is exposed to air or water, becoming 
thicker with increasing temperatures and in the presence of water 
(Source: ``Corrosion Resistance of Aluminum and Aluminum Alloys,'' 
Metals Handbook, Desk Edition, American Society for Metals, 1985). 
Thus, no inspection or acceptance criteria are necessary.
    Comment No. 24: One commenter requested clarification on whether 
the helium will be pure and not mixed with krypton or xenon that would 
have an effect on internal pressure or temperature. The commenter also 
asked whether the helium had to be dry.
    Response: Only pure helium will be used to backfill the cask; no 
krypton or xenon gasses will be added during backfill. Technical 
Specification Table 2-1, Footnote 1, specifies that helium used for 
backfill of MPC shall have a purity of 99.995%. Acceptable 
helium purity for dry spent fuel storage was defined by R. W. Knoll et 
al. at Pacific Northwest Laboratory (PNL) in ``Evaluation of Cover Gas 
Impurities and Their Effects on the Dry Storage of LWR Spent Fuel,'' 
PNL-6365, November 1987. Helium purity is addressed in SAR Section 
8.1.4, MPC Fuel Loading, Step 28, and SER Section 8.1.3.
    Comment No. 25: One commenter asked whether leakage of gases, 
volatiles, fuel fines, and crud was considered credible and whether the 
analysis addressed this concern.
    Response: The applicant has calculated the postulated annual dose 
at 100 meters assuming a realistic leakage rate consistent with ANSI 
N14.5 Standard ``Leakage Tests on Packages for Shipment for Radioactive 
Materials'' (1997) and has reflected the results in SAR Chapter 7. The 
applicant's analysis addresses the commenter's concern, and the 
calculated dose had been found to be within regulatory guidelines 
(limits) and acceptable to the NRC staff.
    Comment No. 26: One commenter was concerned that the cask could 
drop or tip over in the loading area of the plant and whether this has 
been evaluated. The commenter was also concerned about a drop or tip 
over during transfer from the pad or during transport and that all of 
the analysis seemed to be for the pad.
    Response: The tipover, end drops, and horizontal drop analyses form 
part of the structural design basis for the HI-STAR 100 cask design. 
Holtec described drops and tipover analyses in SAR Section 3.4.9. The 
NRC's evaluation of the vendor's analyses is described in SER Sections 
3.2.3.1 and 3.2.3.2. The NRC found the results of these analyses to be 
satisfactory in that the calculated stresses were within the allowable 
criteria of the American Society of Mechanical Engineers (ASME) Code. 
Before using the HI-STAR 100 casks, the general licensee must evaluate 
the foundation materials to ensure that the site characteristics are 
encompassed by the design bases of the approved cask. The events listed 
in the comment are among the site-specific considerations that must be 
evaluated by the licensee using the cask.
    Comment No. 27: One commenter asked whether the design has been 
evaluated for a seismic event during loading and unloading.
    Response: The HI-STAR 100 casks can only be wet loaded and unloaded 
inside the fuel handling facility. Generally, these activities take 
place in a segregated under-water cask loading pit which would limit 
cask movement during a seismic event. The cask will be supported for a 
seismic event during loading and unloading. General procedure 
descriptions for these operations are summarized in Sections 8.1 and 
8.3 of the SAR. Detailed loading and unloading procedures are developed 
and evaluated on a site-specific basis by the licensee using the cask.
    Comment No. 28: One commenter questioned whether the method for 
cooling has been tested with a real cask.
    Response: The NRC regulations and guidance in the Standard Review 
Plan require the review and approval of the design criteria. No testing 
is required for approval of the design under this current rule. The 
cask user is required to perform preoperational testing to determine 
the effectiveness of the cooling methods.
    Comment No. 29: One commenter questioned whether the manufacturer's 
literature for the ``high emissivity'' paint on the overpack had been 
evaluated and tested, how the testing was done, and what the results 
were. The commenter also questioned whether/how the painted components 
were safely stored. The commenter further stated that the paint on the 
surfaces of the overpack should be a specified paint, not just a 
requirement of ``an emissivity of no less than 0.85.''
    Response: The manufacture and application of high-emissivity paints 
is

[[Page 48266]]

not a new technology. Several manufacturers provide paints with 
specified emissivity ratings. Thermal tests are required to confirm the 
heat transfer capabilities of the inner and intermediate shells and 
radial channels. Annual cask inspection will check the exterior surface 
conditions at which time the paint will be examined and touched up in 
local areas as necessary. The NRC does not believe that identifying a 
specific brand name of paint is required. There are several suppliers 
who manufacture paints with the specified emissivity. The NRC has 
reviewed the applicant's analysis and found that paints with an 
emissivity greater than 0.85 are acceptable.
    Comment No. 30: One commenter questioned the drain down time and 
asked how frequently the water is checked. The commenter requested 
information on what happens if the MPC can't be vacuum dried 
successfully and when the fuel needs to be put back in the spent fuel 
pool.
    Response: The drain down time is not specified in the TSs but is 
part of the vacuum drying procedure. The TSs state that the vacuum 
drying must be completed within 7 days. There is not a specific 
procedure in the application to monitor the water content; however, 
that will be addressed by the cask user on a site-specific basis and is 
beyond the scope of this rulemaking. If the drying process is 
unsuccessful and the TS requirements cannot be met within 30 days, the 
fuel assemblies must be moved from the cask and be placed in the spent 
fuel pool.
    Comment No. 31: One commenter requested information on the cask 
storage array on the pad and the radiation affect from other casks in a 
full cask array. The commenter further requested information on how the 
applicant/certificate holder/licensee will examine and/or test the HI 
STAR 100 and who was actually responsible for the test. The commenter 
questioned whether a domed cask cover would be better for runoff and 
sky shine concerns.
    Response: The applicant performed a shielding analysis that 
included a three-by-three cask array (square) model to simulate the 
average dose contribution from the center cask, which is partially 
shielded by the surrounding periphery casks. This value is applied in 
an offsite dose formula used to estimate offsite doses from every cask 
in the array. The center-to-center cask pitch was assumed to be 12 feet 
in the shielding analyses. Testing of the actual as-installed 
configuration will be performed by the cask user and will be evaluated 
at that time. Offsite dose estimates for a typical ISFSI array, 
including the affects of multiple casks and skyshine, are discussed in 
Sections 5.4.3 and 10.4.1 of the SAR. NRC found the dose estimates to 
be acceptable. As required in 10 CFR 72.212, each general licensee will 
perform a site-specific dose evaluation to demonstrate compliance with 
Part 72 radiological requirements. The general licensee will identify 
an ISFSI configuration and may elect to use additional engineered 
features of its choosing, such as shield walls, a domed cover, or 
berms, to ensure compliance with radiological requirements. Section 
1.4.7 of Appendix B to the CoC requires that any such engineered 
feature be considered important to safety and evaluated to determine 
the applicable quality assurance category.
    Comment No. 32: One commenter questioned what the criteria were for 
the polyester resin ``poured'' into radial channels, how they were 
tested, handled and inspected, and whether they had been tested in a 
real cask. The commenter questioned whether a ``poured'' neutron shield 
was really safe and whether uncontrolled voids caused a problem with 
occupational dose requirements. The commenter stated that poured 
neutron shields should not be used.
    Response: The NRC has reviewed Holtec's application that described 
the neutron shielding to be used to meet the requirements of 10 CFR 
72.104 and 72.106. The NRC found the Holtec approach acceptable. The 
methods for testing, handling, and inspecting installation of the 
shielding are beyond the scope of this rulemaking. However, poured 
neutron shielding has been successfully used in other cask designs.
    Comment No. 33: One commenter stated that appropriate limits for 
burnup should be specified in the CoC. The commenter is concerned that 
the SAR analysis assumed significantly higher burnups than allowed and 
significantly higher initial uranium loading than specified in the 
table.
    Response: Burnup, cooling time, initial uranium loading, and 
initial enrichment are parameters that affect the total source term 
(radioactivity) of spent fuel. The applicant's source term analysis 
assumed higher uranium loadings and higher burnups than those specified 
in TSs of the CoC. Therefore, the radiological source term is 
conservative relative to the allowed burnups and uranium loadings.
    As discussed in Section 5.2.1 of the preliminary SER, for the same 
level of burnup, neutron source terms typically increase as initial 
enrichment decreases. Therefore, the source term analysis employed 
lower-than-average enrichment values. Based on the SAR analyses, 
conditions of the CoC, and other requirements in Parts 20 and 72, the 
NRC has determined that minimum enrichment is not warranted as an 
additional operating control for the HI-STAR 100. Specific reasons for 
this determination include the following: (1) the enrichments bound a 
significant portion of spent fuel, and the source terms are calculated 
for burnups significantly higher than those allowed in the CoC; (2) the 
radiological source terms are adequately controlled in the CoC by 
limits on maximum burnup, minimum cooling time, maximum initial uranium 
loading, and maximum decay heat; (3) dose rates are controlled in the 
CoC by specific dose limits for the top and side of the cask that are 
based on values calculated in the shielding analysis; (4) each general 
licensee will perform a site-specific dose evaluation to demonstrate 
compliance with Part 72 radiological requirements; and (5) each general 
licensee will operate the ISFSI under a Part 20 radiological protection 
program.
    NRC agrees with the comment that the preliminary SER term of ``low 
probability'' may not provide definite criteria for general license 
cask users regarding limitations on minimum enrichment. Therefore, 
Chapter 5 of the SER has been revised to clarify that minimum 
enrichment is not an operating control for the HI-STAR 100.
    Comment No. 34: One commenter asked what has been considered as 
credible ways to lose the fixed neutron poisons.
    Response: The NRC staff does not consider the loss of fixed neutron 
poisons to be credible after they are installed into the cask because 
the poisons are fixed in place and contained.
    Comment No. 35: A commenter questioned how the welds of the MPC lid 
and closure ring are tested and asked for the acceptance criteria.
    Response: Information on the welds is contained in SAR Tables 
9.1.1, 9.1.2, and 9.1.3.
    Comment No. 36: One commenter asked whether shims are used and 
stated that shims or gaps were not acceptable.
    Response: There are no shims used in the closure weld of the HI-
STAR 100 casks. The only shims used are located between the canister 
and the overpack at basket support locations to provide additional 
support for the basket supports. The actual thickness of the shim will 
depend on the gaps between the cask and the inside cavity of the 
overpack at the basket support locations. Gaps between separate 
components such as the cask and the

[[Page 48267]]

overpack are unavoidable and are necessary to ensure that there will be 
no physical interferences and to allow free thermal expansions.
    Comment No. 37: One commenter stated that all welds should be 
monitored unless they have been tested.
    Response: NRC accepts welded closure of casks. The regulations do 
not require monitoring or testing of welds because there are no 
expected degradation mechanisms identified during the cask usage life. 
However, both the fabricator and cask user will examine and inspect all 
welds as appropriate.
    Comment No. 38: One commenter stated that the detailed loading and 
unloading procedures developed by each cask user should be put in the 
PDR.
    Response: Loading and unloading procedures are site-specific issues 
not required for design approval and are beyond the scope of this 
rulemaking.
    Comment No. 39: One commenter asked how long before an ultrasonic 
testing examination is conducted should the equipment be calibrated.
    Response: Comments on the site-specific examination techniques and 
associated calibration are beyond the scope of rulemaking for the HI-
STAR 100 system.
    Comment No. 40: One commenter was concerned over the possibility 
that the bolts could rust and crack over time or become brittle and 
crack because water, ice, and frost could get into the bolt holes over 
the years.
    Response: The NRC disagrees with this concern over the integrity of 
the bolting material. The 54, 1\5/8\-inch-diameter, closure plate bolts 
are made from ASME SB-637-N07718 material per SAR BM-1476. N07718, a 
nickel-chromium alloy, does not become brittle at colder temperatures. 
N07718 is a high strength, corrosion resistant material used in 
applications with a temperature range from -423  deg.F (-253  deg.C) to 
1300  deg.F (704  deg.C) (Source: Inconel Alloy 718, Inco Alloys 
International, fourth edition, 1985). This material will not rust, 
unlike carbon steels in corrosive environments. In addition, the 
material retains significant ductility down to -320  deg.F (-196 
deg.C) as shown by impact test results (Source: Inconel Alloy 718, 
Table 27). Therefore, the NRC has no concerns about the bolting 
material.
    Comment No. 41: One commenter asked what type of radiographic exam 
is applicable and where it would be conducted.
    Response: SAR Tables 9.1.1, 9.1.2, and 9.1.3 describe which 
radiographic exams are to be performed and when they are required to be 
performed.
    Comment No. 42: One commenter disagreed with allowing the use of a 
penetrant test in lieu of volumetric examination on austenitic 
stainless steels because flaws in these are ``not expected'' to exceed 
the thickness of the weld head. The commenter believes that volumetric 
welds should be required because if you don't know for sure the real 
size of the actual weld, how can you accept a certain flaw size? The 
commenter asked how the permanent record is kept and stated that black 
and white photographs should be used as a permanent record.
    Response: NRC disagrees with this comment. The NRC position on 
inspection of closure welds is contained in ISG-4, ``Cask Closure Weld 
Inspections.'' Actual cask welds are examined in accordance with site-
specific procedures that are beyond the scope of rulemaking for the HI-
STAR 100 system. Nondestructive Examination (NDE) methods are specified 
in accordance with Section III ``Rules for Construction of Nuclear 
Power Plant Components,'' and Section V ``Nondestructive Examination,'' 
of the ASME Code and are already described in SAR Tables 9.1.1, 9.1.2, 
and 9.1.3. 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.
    Comment No. 43: One commenter believed that the marking material 
for the casks should be designated and that the mark needed to be 
permanent.
    Response: NRC agrees with the comment. The storage marking 
nameplate is made from a 4-inch by 10-inch, 14-gauge Type 304 stainless 
steel sheet and welded to the outside of the HI-STAR 100 Overpack. 
Lettering will be etched or stamped on the plate. Details are shown in 
SAR Drawing 1397, Sheet 4 of 7, and described in SER Section 9.1.6. The 
nameplate will provide appropriate cask identification that will last 
well beyond the design life of the HI-STAR 100 system. No nonpermanent 
marking will be used.
    Comment No. 44: One commenter requested information on ``rupture 
disc replacements,'' how they are tested for replacement, what the time 
criteria are, and what is considered a rupture.
    Response: The rupture disc is located in the neutron shield tank of 
the HI-STAR 100 casks. The purpose of the rupture disc is to limit 
pressure build-ups to a precalculated level within the neutron shield 
tank during the fire accident condition. When the pressure build-up 
exceeds the precalculated design pressure, the disc will rupture to 
relieve the pressure. The rupture disc is tested and certified by the 
manufacturer. There is no regulatory requirement for the replacement of 
rupture discs. The SAR has arbitrarily set a replacement schedule for 
every 5 years to assure functionality.
    Comment No. 45: One commenter asked if the casks are checked in 
winter for ice and snow loads or ice around the base and if the pads 
will be kept clean.
    Response: Casks are designed for the worst ice and snow loads 
possible. Ice build-ups around the cask base are not allowed, and the 
pad will be kept clean. Site-specific procedures will address these 
items.
    Comment No. 46: One commenter questioned if there was an evaluation 
for a plane crash, with a fuel fire, into a cask or full cask array 
conducted and whether there is a stipulation as to putting a pad in an 
area where planes regularly fly.
    Response: Before using the HI-STAR 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 cask as 
required by 10 CFR 72.212(b)(3). The licensee's site evaluation should 
consider the effects of nearby transportation and military activities. 
Generally, a cask's inherent design will withstand tornado missiles and 
collision forces imposed by light general aviation aircraft (i.e., 
1500-2000 pounds) that constitute the majority of aircraft in operation 
today. The events listed in the comment are among the site-specific 
considerations that must be evaluated and are beyond the scope of this 
rulemaking.
    Comment No. 47: One commenter questioned why Holtec stated that the 
HI-STAR 100 could be part of the final geologic disposal system.
    Response: The NRC is not reviewing this design for use in a final 
geologic disposal system, but only for interim storage under Part 72.
    Comment No. 48: One commenter asked where the MPC shell weld is 
located and if the pocket trunnions at the bottom of the overpack have 
been analyzed specifically for tipovers and falls.
    Response: The MPC shell has multiple welds located both 
longitudinally on the side of the MPC and circumferentially on the top 
and bottom of the MPC. The pocket trunnions at the bottom overpack have 
been analyzed by the applicant for tipovers and falls. The NRC reviewed 
the design for normal, off-normal, and

[[Page 48268]]

accident conditions, and found it acceptable.
    Comment No. 49: One commenter stated that the lifting and pocket 
trunnions should be checked over the years for cracking or brittleness 
and for debris accumulation and should be kept ready for use over the 
years.
    Response: The NRC agrees with this comment. As shown in SAR Table 
9.2.1, lifting trunnion and pocket trunnion recesses are visually 
inspected before the next handling operation after HI-STAR 100 casks 
are placed on the ISFSI pad. The trunnion material has been evaluated 
for brittle fracture and found to be satisfactory for the operating 
temperature range. In addition, the trunnions are load tested in 
accordance with ANSI N14.6, ``American National Standard for 
Radioactive Materials--Special Lifting Devices for Shipping Containers 
Weighing 10000 Pounds (4500 kg) or More.'' Thus, there is no credible 
reason to suspect undetected cracking or brittleness. The pocket 
trunnion recess is closed by a pocket trunnion plug during storage. 
There is no possibility of animal and bird access and nesting in the 
recess.
    Comment No. 50: One commenter requested information on the criteria 
for the critical flaw size.
    Response: The criteria for critical flaw size are included in ISG 
No. 4, ``Cask Closure Weld Inspections.'' The NRC review determined 
that Holtec's proposed methodology is consistent with this ISG.
    Comment No. 51: One commenter asked how subcontractors are to be 
audited and inspected.
    Response: This comment is beyond the scope of this rulemaking.
    Comment No. 52: One commenter believed that the first cask for each 
utility should be tested at a full heat load and asked what is meant by 
the ``First System In Place'' requirement.
    Response: The heat transfer characteristics of the cask system will 
be recorded by temperature measurements for the first HI-STAR 100 
systems (MPC-24 and MPC-68) placed into service with a heatload greater 
than or equal to 10 kW. An analysis shall be performed by the cask user 
that demonstrates that the temperature measurements validate the 
analytical methods and the predicted thermal behavior described in 
Chapter 4 of the SAR.
    The cask user will perform validation tests for each subsequent 
cask system that has a heat load that exceeds a previously validated 
heat load by more than 2 kW (e.g., if the initial test was conducted at 
10 kW, then no additional testing is needed until the heat load exceeds 
12 kW). No additional testing is required for a system after it has 
been tested at a heat load greater than or equal to 16 kW.
    The cask user will provide a letter report to the NRC in accordance 
with 10 CFR 72.4 summarizing the results of each of these validation 
tests. Cask users may also satisfy these testing and reporting 
requirements by referencing validation test reports submitted to the 
NRC by other cask users with identical designs and heat loads.
    Comment No. 53: One commenter asked how much water is to be drained 
under the MPC lid before welding and how the temperature enters into 
the calculations.
    Response: Chapter 8 of the SAR directs the operators to pump 
approximately 120 gallons of water from the MPC before commencing 
welding operations. The water level is lowered to keep moisture away 
from the weld region. Under these conditions, ample water remains 
inside the MCP to maintain cladding temperatures well below their short 
term limits. This operating condition has been evaluated by the NRC. 
The resulting temperature increase is much less than any previously 
analyzed accident condition might produce.
    Comment No. 54: One commenter asked how lifting height should be 
verified and stated that the height should be recorded.
    Response: The maximum lifting height maintains the operating 
conditions of the Spent Fuel Storage Cask (SFSC) within the design and 
analysis basis. It is the general licensee's responsibility to limit 
the SFSC lifting height to allowable values. The lift height 
requirements are specified in TS LCO 2.1.7 for the vertical and 
horizontal orientations. Surveillance requirements require verification 
that SFSC lifting requirements are met after the SFSC is either 
suspended or secured in the transporter and prior to moving the SFSC 
within the ISFSI.
    Comment No. 55: One commenter questioned how the MPC closure ring, 
lid, vent, and drain covers are removed during unloading and what 
precautions are taken.
    Response: The specific procedures for removal of the closure ring, 
lid, vent, and drain covers are to be developed by the cask user. These 
procedures will be evaluated by the licensee and by the NRC during 
inspections to address adequacy and implementation and, therefore, are 
beyond the scope of this rulemaking.
    Comment No. 56: One commenter questioned that if the MPC gas 
temperature is not met, what additional actions are required and have 
they been evaluated (TS B3.1.8-3)?
    Response: The NRC staff has evaluated this condition. The TSs 
require that if the MPC gas temperature is exceeded during unloading, 
no additional operational actions may be conducted until the 
temperature is restored to below the TS limit.
    Comment No. 57: One commenter asked if ``dry'' unloading operations 
are considered.
    Response: A dry unloading operation was not requested or explicitly 
described in the SAR and thus is not currently allowed for the HI-STAR 
100 system and is beyond the scope of this rulemaking.
    Comment No. 58: One commenter questioned if crud disposal is a 
problem and how it can be mitigated.
    Response: Dispersal of crud is beyond the scope of this rulemaking 
and is a site-specific issue. Experience with wet unloading of some 
fuel types after transportation has involved handling significant 
amounts of crud. However, the NRC notes that the HI-STAR generic 
unloading procedures mitigate crud dispersal. As discussed in Section 
8.3.1 of the SAR, these procedures include gas sampling of the MPC 
internal atmosphere and specific cool-down steps. Each cask user will 
develop additional site-specific unloading procedures based on its 
radiological protection program to further address and mitigate crud 
dispersal.
    Comment No. 59: The applicant made comments relevant to the helium 
backfill pressure of the cask. After discussions with the NRC staff, 
Holtec withdrew this comment during a telephone conversation on 5/7/99.
    Response: Not applicable.

Comments on Proposed TSs

    Upon review of the public comments received on the proposed TSs for 
the HI-STAR-100 Storage Cask, particularly comments received from EXCEL 
Corporation and the Holtec Users Group, the NRC staff has determined 
that several structural changes to the TSs were in order. These changes 
result in a clearer set of TSs and move the TSs from the new generation 
of dual-purpose cask systems toward a standardized format.
    Comment No. 60: It was suggested that controlling the bases for the 
TSs as part of the CoC would result in administrative burdens to all 
involved. These bases are not controlled as part of power reactor 
licenses.
    Response: The NRC staff agrees. Therefore, the bases have been 
relocated to an appendix to the SAR.

[[Page 48269]]

    Comment No. 61: A number of commenters also raised concerns with 
the inclusion of the extensive fuel specifications (formerly Section 
2.0) and a very lengthy design specification section (formerly Section 
4.0).
    Response: The NRC staff agrees that placement of much of this 
information in the TSs is unwarranted. Therefore, much of the 
information regarding fuel specifications and some of the design and 
codes information were moved from the TSs to a separate appendix to the 
CoC. However, the NRC staff did maintain some of the information 
regarding requirements for bases controls by adding it to a revised 
Section 3.0, ``Administrative Controls and Programs,'' of the TSs.
    Upon consideration of public comments and further consideration 
within the NRC, the NRC staff has determined that the structure of TS 
Section 2.1, ``SFSC INTEGRITY,'' did not provide appropriately clear 
guidance. Therefore, the NRC staff has revised this section of the TSs 
to reflect a more logical and focused approach. The number of limiting 
conditions for operations (LCOs) in this section has been reduced to 
four. The NRC staff believes that this will enhance the usefulness of 
the TSs.
    Comment No. 62: One commenter stated that if surface contamination 
exceeds 2200 dpm/100 cm2 from gamma and beta emitting sources, and 
smearable contamination limits cannot be reduced to acceptable levels, 
the TSs require actions up to and including removal of the MPC from the 
HI-STAR 100 overpack after removing the spent fuel from the MPC. The 
commenter stated that the proposed Skull Valley ISFSI in Utah does not 
have facilities for decontaminating casks and, therefore, these TSs 
could not be met.
    Response: The NRC agrees in part. The revised version of the TSs 
(TS 2.2.2) requires verification that removable contamination is within 
limits during loading operations and provides up to 7 days to restore 
the contamination within limits. The specifications no longer list MPC 
or spent fuel removal actions. Further, comments on the proposed site-
specific Skull Valley ISFSI currently under review are beyond the scope 
of this rulemaking. Decontamination requirements will be reviewed as 
part of the site-specific licensing provisions under Part 72 Subpart B 
for the Skull Valley ISFSI.
    Comment No. 63: One commenter stated that the definition of 
``TRANSPORT OPERATIONS'' needs to be revised to reflect that the drop 
analysis is not limited to drops from the transporter, and that lifting 
of a cask with other devices is not prohibited. The commenter 
recommended similar changes to the definition of ``LOADING OPERATIONS'' 
and ``UNLOADING OPERATIONS.''
    Response: The NRC disagrees. The definitions of the three terms in 
question do not prohibit lifting of a cask with other devices (the 
revised note in TS 2.1.3 clarifies this issue), nor do the definitions 
affect the lifting requirements contained in TS 2.1.3.
    Comment No. 64: One commenter stated that it would increase the 
standardization of the TSs by relocating the explanatory information of 
the defined terms in TS Section 1.0 to the TS Bases.
    Response: The NRC disagrees with the comment. The terms defined in 
TS Section 1.0 are important in the understanding of the TS 
requirements. These definitions need to be contained within the TSs. 
This practice is consistent with the standard TSs developed for the 
U.S. nuclear power reactors.
    Comment No. 65: One commenter stated that in Examples 1.3-2 and 
1.3-3, the word ``action'' should be capitalized.
    Response: The NRC agrees. The word ``action'' has been capitalized.
    Comment No. 66: One commenter recommended the removal of portions 
of Table 2.1-1 and all of Table 2.1-2 and Table 2.1-3 from the TSs.
    Response: The NRC agrees, in part, that this information should be 
moved. This design information is crucial to the conclusions reached by 
the NRC staff in its SER; therefore, the design information contained 
in these tables has been relocated (and renumbered) to a separate 
appendix to the CoC, along with other critical design information.
    Comment No. 67: One commenter recommended a change to the format of 
the Titles of Tables 2.1-1, 2.1-2, 2.1-3, and 2.1-4.
    Response: The NRC agrees with the comment. The format has been 
changed.
    Comment No. 68: One commenter recommended a wording change in TS 
Section 3.0 from ``not applicable to an SFSC'' to ``not applicable.''
    Response: The NRC agrees with this comment and has made the 
indicated change.
    Comment No. 69: One commenter stated that there is no need to 
create two specifications for TS 3.1.1, MPC Cavity Vacuum Drying 
Pressure, and TS 3.1.2, OVERPACK Annulus Vacuum Drying Pressure. In 
addition, the commenter indicated there is no need to create two 
specifications for TS 3.1.5, MPC Helium Leak Rate, and TS 3.1.6, 
OVERPACK Helium Leak Rate.
    Response: The NRC agrees with the comment. Section 2.1 of the TSs 
has been revised based on these and similar comments received to 
combine these TSs.
    Comment No. 70: One commenter stated that the frequency of SR 
3.1.7.1 should be revised because, as written, the frequency would 
apply only when a cask is being moved to or from the ISFSI and would 
not apply at other times, such as when moving casks within the ISFSI. 
However, the drop analysis applies any time the cask is suspended. The 
frequency should be revised similar to ``Prior to movement of an 
SFSC.''
    Response: The NRC agrees with the comment. The frequency of SR 
3.1.7.1 has been revised.
    Comment No. 71: One commenter recommended that TS Sections 4.1 and 
4.2 be eliminated because they contain no unique information.
    Response: NRC agrees with the comment. Sections 4.1 and 4.2 have 
been eliminated.
    Comment No. 72: One commenter recommended relocating the 
information contained in TS Sections 4.3 and 4.5 to the SAR, and 
recommended eliminating TS Section 4.4, stating that this section is a 
duplication of existing regulatory requirements.
    Response: The NRC agrees in part. The NRC staff agrees that these 
sections do not belong in the TSs. This design information has been 
relocated to Appendix B to the CoC. The NRC staff disagrees with the 
commenter's proposal to eliminate or relocate these sections to the 
SAR. The NRC has relocated these sections to Appendix B to the CoC due 
to the importance of the design information contained in these 
sections. The NRC staff also disagrees with the comment that TS Section 
4.4 is a duplicate of existing regulations, since this section contains 
the acceptance criteria for the site-specific design parameters.
    Comment No. 73: A commenter recommended relocating the information 
contained in TS Sections 4.6 and 4.8 to an Administrative Controls 
chapter due to their content and relocating Section 4.7 to the SAR 
because it is a one-time administrative task.
    Response: The NRC agrees in part. The NRC staff agrees that these 
sections belong in the administrative section of the TSs and has placed 
this information in a new TS Chapter 3.0, ``Administrative Controls and 
Programs.'' The NRC staff disagrees with the commenter on the proper 
location of Section 4.7 (now TS Section 3.2), because it is established 
NRC staff

[[Page 48270]]

practice to place important administrative requirements, even one-time 
requirements, in the TSs.
    Comment No. 74: A commenter stated that TS 3.1.8 contains conflicts 
because the APPLICABILITY statement, and the COMPLETION TIME when the 
condition is not met, are the same statement. The commenter further 
recommended that because of its complexity and rarity of its use, this 
specification be eliminated and the information specified in the SAR.
    Response: The NRC agrees in part. The NRC agrees with the first 
point. TS 2.1.4 has been rewritten to remove this conflict. The NRC 
staff disagrees with the second point and considers this information 
important to the proper operation of the cask system. Further, the 
changes made to this section resolve concerns regarding its complexity.
    Comment No. 75: One commenter recommended relocating the figure 
attached to TS 3.2.1 to the TS Bases, because the purpose of the figure 
is to show where dose measurements should be taken.
    Response: The NRC disagrees with this comment. This figure, now 
attached to TS 2.2.1, is an integral part of the proper implementation 
of this TS and assures that the dose measurements will be taken at the 
proper locations.
    Comment No. 76: The commenter stated that the TSs do not comply 
with 10 CFR 72.44(d) that requires TSs on radioactive effluents.
    Response: The NRC agrees with this comment. TS Section 3.0 has been 
revised to incorporate the requirements of 10 CFR 72.44(b).
    Comment No. 77: One commenter recommended that within TS Section 
1.1, the definition for ``Intact Fuel Assembly'' should be revised to 
state `` * * * an amount of water greater than or equal to * * *,'' 
adding the term ``greater than or'' to allow greater flexibility with 
respect to dummy rod sizing.
    Response: The NRC agrees with the comment and has revised the 
definition.
    Comment No. 78: One commenter recommended that within TS Table 2.1-
1, Item II.B should be reworded for clarification because the current 
wording could be misinterpreted by users that intact fuel assemblies 
are required to be loaded into damaged fuel containers.
    Response: The NRC agrees with the comment. The table, which has 
been relocated to Appendix B, has been revised.
    Comment No. 79: One commenter requested clarification of TS Section 
4. As written, the text does not require a written report of the 
results of the first measurements, only ``each cask subsequently loaded 
with a higher heat load.'' NRC's intent to require a written report for 
the first temperature measurements is not clear. The commenter further 
stated that it is not clear what ``calculation'' is being referred to 
in the last two sentences, whether it is the original design 
calculation or a new calculation generated from the test. The commenter 
further recommended the addition of ``decay heat'' after ``lesser'' and 
before ``loads'' in the last line.
    Response: The NRC agrees with these comments, except for the 
recommendation to add the phrase ``decay heat,'' which the NRC 
considers unnecessary. TS Section 3.3 has been revised to clarify the 
reporting requirements and the calculational comparison required by 
this TS condition.
    Comment No. 80: One commenter recommended some editorial changes to 
revise TS Bases 2.2.2 and 2.2.3 to clarify that 10 CFR 72.75 has 
additional reporting requirements that may need to be met independent 
of these TS requirements.
    Response: The NRC agrees with the comment. A reference to 10 CFR 
72.75 has been added to Appendix B to the CoC.
    Comment No. 81: One commenter recommended adding a new definition 
for fuel building to the TSs.
    Response: The NRC agrees with the comment. A definition for fuel 
building has been added to the TSs.
    Comment No. 82: One commenter recommended editorially revising TS 
LCO 3.1.7, ``SFSC Lifting Requirements'' and the related bases to 
clarify the applicability. The revision is necessary because the LCO is 
not intended to be applicable while the transport vehicle is in the 
fuel building or when the cask is secured on a railcar or heavy haul 
trailer because the cask is not being lifted.
    Response: The NRC agrees with the comment. TS 2.1.3 has been 
revised accordingly.
    Comment No. 83: One commenter recommended a revision to TS Tables 
2.1-2 and 2.1-3, Note 1, for the purposes of clarification and to allow 
for manufacturer tolerances.
    Response: The NRC agrees with the comment. The recommended changes 
to the tables have been made. The table has been relocated to Appendix 
B of the CoC.
    Comment No. 84: One commenter recommended the revision of TS Table 
3-1, Item 1.c, to change the lower helium tolerance to 10 percent 
because the smaller tolerances were associated with convection heat 
transfer, for which no credit is taken in the application.
    Response: The NRC agrees with the comment and has revised 
renumbered TS Table 2-1.
    Comment No. 85: One commenter recommended that TS 4.3.1 be revised 
to allow for changes to codes and standards because it would provide 
both the vendor and the NRC the flexibility to add exceptions/
alternatives to the code without amending the certificate.
    Response: The NRC agrees with the comment. Section 1.3.2 of 
Appendix B has been revised accordingly.
    Comment No. 86: The applicant recommended in TS Section 4.4.6, the 
revision of the soil effective modulus of elasticity from 
``6,000psi'' to ``28,000 psi.'' In addition, the 
commenter recommended an acceptable method for licensees to comply with 
the soil modulus limit.
    Response: The NRC agrees with the comment. The information has been 
added to Appendix B to the CoC.
    Comment No. 87: One commenter recommended the addition of a third 
option to TS LCO 3.1.7 and Bases B3.1.7 (or elsewhere in the TSs) that 
allows general licensees to calculate site-specific lifting 
requirements based on the site-specific pad design and associated drop/
tipover analyses.
    Response: The NRC agrees with the comment. TS LCO 2.1.3 has been 
revised to add this option.
    Comment No. 88: One commenter believed that the 48-hour time limit 
within TSs 3.1.1 through 3.1.6 is overly restrictive.
    Response: The NRC agrees with this comment in part. Accordingly, 
the NRC has reviewed the time limit in each applicable TS. Some of the 
time limits have been extended to provide for a controlled, deliberate 
response to the LCO condition.
    Comment No. 89: One commenter recommended the deletion of the 
Design Features, Section 4.6, Training Module, and Section 4.7, Pre-
Operational Testing and Training Exercise because the review of the 
training program is required by 10 CFR 72.212(b)(6) and the TS 
duplicates the requirement in the regulation.
    Response: The NRC agrees in part. The NRC agrees that there is 
duplication in the TSs and the regulatory requirements. Accordingly, TS 
3.1 (previously Section 4.6) has been modified to reference the general 
licensee's systematic approach to training. However, the NRC staff 
believes that listing the training exercises as a specific requirement 
for proper cask operation is appropriate to

[[Page 48271]]

be included in the TSs, and it has been maintained.
    Comment No. 90: One commenter recommended adding ``diesel'' before 
``fuel'' in TS Section 4.4.5 and in SER Sections 3.1.2.1.8, 4.3.4, and 
4.4.3.4 for clarification.
    Response: The NRC agrees conceptually with the comment. TS Section 
4.4.5 (now 1.4.5 of Appendix B) and SER Sections 3.1.2.1.8, 4.3.4, and 
4.4.3.4 have been revised to refer to combustible transporter fuel.

Comments on the Draft CoC

    Comment No. 91: Two commenters recommended that CoC Condition 10 be 
revised to be consistent with 10 CFR 72.48 for the cask design and 
operating procedures. Another commenter stated that Condition 10 was 
not clear.
    Response: The NRC agrees with the comments. The applicable CoC 
condition has been revised to delete the prescriptive controls for 
making changes to the cask design and operating procedures. The 
condition now reflects 10 CFR 72.48 as recently approved by the 
Commission.
    Comment No. 92: Two commenters recommended that a Bases Control 
Program be added to the TSs or CoC.
    Response: The NRC disagrees with the comment. The proposed TS bases 
are part of the SAR. Because 10 CFR 72.48 provides a change process for 
the SAR for control of the bases, there is no need to incorporate this 
program into the CoC or TSs.
    Comment No. 93: One commenter requested information on the status 
of a petition for rulemaking on the change process in 10 CFR 72.48.
    Response: This comment is beyond the scope of this rulemaking.
    Comment No. 94: One commenter stated that the description of the 
attachment to the CoC was in error.
    Response: The NRC agrees with this comment. The description has 
been corrected.

Comments on the NRC Staff's SER

    Comment No. 95: One commenter asked a question about what is meant 
by the statement included in the NRC SER in Section 9.3 related to the 
examination and/or testing of the HI-STAR 100 by the applicant/
certification holder/licensee.
    Response: The SER refers to Section 9.1 of the applicant's SAR. 
This section summarizes the scope and acceptance criteria for the HI-
STAR 100 test program. It includes fabrication and nondestructive 
examinations, weld inspecting, structural and pressure tests, leakage 
tests, component tests, and shielding and integrity testing and 
controls. The SAR or SER does not specify which entity must perform 
each test. This is because some tests are performed during fabrication, 
while others can only be performed after installation. The quality 
assurance programs implemented by the fabricator, certificate holder, 
or applicant with appropriate oversight will ensure that these SAR 
specified tests are completed and are effective. Further, the NRC 
inspection program also verifies on a sampling basis that tests and 
surveillances are conducted as required.
    Comment No. 96: One commenter recommended revising the last 
sentence of the first paragraph of SER Section 3.1.2.1.6 to read: ``The 
design-basis earthquake accelerations are assumed to be applied at the 
top of the ISFSI concrete pad with the resulting inertia forces applied 
at the HI-STAR 100 mass center.''
    Response: The NRC agrees with the comment. The SER has been 
revised.
    Comment No. 97: One commenter recommended in SER Section 3.1.4.4, 
in the first paragraph, the replacement of ``* * * the fabricator is an 
accredited facility by the ASME for nuclear fabrication work holding 
``N'' and ``NPT'' stamps, * * *'' with ``* * * the HI-STAR 100 System 
is designed in accordance with the ASME Code, as clarified by the 
exceptions to the Code listed in TS Table 4-1.''
    Response: The NRC agrees with the comment. The SER has been 
revised. Note that the table is now in Appendix B.
    Comment No. 98: One commenter recommended that in SER Section 6.3, 
the word ``minimum'' be replaced with ``maximum'' in the third sentence 
of the first full paragraph to match the analysis.
    Response: The NRC agrees with the comment. The SER has been revised 
to correct the error.
    Comment No. 99: One commenter stated that SER Section 8.1.4, which 
discusses the evaluation of welding and sealing procedures, should be 
revised to recognize the option of performing manual welding of the MPC 
lid closure weld in accordance with a user's as low as reasonably 
achievable (ALARA) practices.
    Response: The NRC disagrees with the comment. As discussed in 
Sections 8.1 and 10.1 of the SAR, the use of the Automated Weld System 
provides justification that the HI-STAR 100 is designed in accordance 
with Part 72 radiological requirements and ALARA objectives consistent 
with Part 20. However, the intent of the proposed SER revision is 
already implied in Section 8.1.2 of the SER that states: ``Each cask 
user will need to develop detailed loading procedures that incorporate 
the ALARA objectives of their site-specific radiation protection 
program.'' Therefore, each user can develop site-specific operating 
procedures based on ALARA objectives that would include the use of 
manual welding and make changes to the SAR in accordance with 10 CFR 
72.48.
    Comment No. 100: One commenter recommended that SER Section 8.3.1, 
which discusses the evaluation of cooling, venting, and reflooding 
during cask unloading operations, should be revised to allow the option 
of a once-through purge in lieu of the closed-loop cooling system.
    Response: The NRC disagrees with this comment. An amendment 
application with a specific design and supporting analysis for a once-
through helium cooling system would be required for NRC review and is 
beyond the scope of this rulemaking.
    Comment No. 101: One commenter noted that a more appropriate method 
to implement the thermal test for the overpack had been accepted by the 
NRC for the HI-STAR 100 transportation cask and recommended this method 
be used for this cask design. Appropriate changes were recommended to 
be made to the SER and SAR.
    Response: The NRC agrees that this method should be included in the 
SAR for the HI-STAR 100 storage cask. Appropriate changes have been 
made to Section 9.1.6 of the SAR and Chapter 9 of the SER.
    Comment No. 102: The applicant submitted numerous editorial 
comments on the SAR, SER, and CoC. Comments were intended as 
clarification, restoration of deleted information, grammatical 
corrections, corrections to text, to maintain consistency between 
documents, typographical corrections, format changes, and to correct 
terminology. These editorial changes do not change the design of the 
cask or supporting analysis.
    Response: The NRC agrees with many of the editorial comments 
suggested by Holtec International. The SAR, SER, and CoC have been 
revised to address the comments as appropriate.

Comments on the Applicant's Topical SAR

    Note: In response to comments received, a number of changes to 
the SAR were made by Holtec International, as discussed below.

    Comment No. 103: One commenter proposed a revision to the language 
in Section 8.0 of the SAR to clarify that users will have some 
flexibility to use

[[Page 48272]]

procedures and equipment suitable for site-specific needs and 
capabilities.
    Response: The NRC agrees with the suggested editorial changes. The 
changes to the SAR have been made.
    Comment No. 104: One commenter recommended some editorial changes 
within SAR Section 4.4, because the wording in Subsection 4.1.1.15 may 
be erroneously interpreted to mean that the chilled helium delivered to 
the MPC cavity to cool the internals prior to flooding the cavity with 
water must be at 100  deg.F. The commenter stated that the text of the 
SAR requires clarification to permit each cask user's cooldown system 
to be engineered with the flexibility to cool MPCs containing fuel with 
varying levels of decay heat production.
    Response: The NRC agrees with the comment. The SAR has been 
revised.
    Comment No. 105: In SAR Section 1.5, Drawings 1399, Sheet 3, and 
BM-1476, and in Drawing Section ``N-N,'' one commenter recommended the 
addition of four threaded holes spaced 90 degrees apart as a personnel 
dose reduction enhancement. The new holes would allow the personnel 
attaching the shield to work in an area of lesser exposure to radiation 
within the same time frame. The effect of the shield attachment will 
remain the same.
    Response: The NRC agrees with the comment. Drawings 1399 and BM-
1476 have been revised to reflect the change.
    Comment No. 106: One commenter suggested that in SAR Revision 10, 
the drawings in Chapter 1 be revised to match those approved by the NRC 
in the transportation SAR.
    Response: The NRC agrees with the comment. Seven drawings in SAR 
Section 1 have been revised to match those in the transportation SAR. 
Although four drawings have not been revised to match the 
transportation SAR, this is acceptable to the NRC staff because they 
reflect storage design features.
    Comment No. 107: In the SAR, one commenter (the applicant) 
recommended changing Section 6.1 by replacing ``(20  deg.C-100  deg.)'' 
with ``(i.e., water density of 1.000 g/cc)'' and delete ``(20  deg.C 
assumed)'' to more accurately describe the assumption made in the 
analyses.
    Response: The NRC agrees. The SAR has been revised as suggested by 
the commenter.
    Comment No. 108: The applicant suggested a number of changes to the 
drawings for the HI-STAR 100 Storage Cask. These changes did not 
require a change to the supporting design analyses.
    Response: The NRC agrees that the changes to the drawings were 
appropriate and do not result in any changes to the supporting design 
analyses. The SAR drawings have been revised in accordance with the 
suggested changes.
    Comment No. 109: The applicant suggested using Magnetic Particle 
Examination in lieu of Liquid Penetrant Examination for the overpack 
weld examination and recommended changes to the associated drawing 
notes.
    Response: The NRC agrees with this suggested change. The NRC agrees 
that resolution of this comment will involve a change to the drawings 
which will mean that drawings referencing this examination shall be 
different for the storage and transportation certificates. These 
differences are not significant because the staff finds Magnetic 
Particle Examination to be equally acceptable to Liquid Penetrant 
Examination. Appropriate changes to the drawings have been made.
    Comment No. 110: The applicant suggested a clarification for the 
sequence for the hydrostatic testing and helium leakage testing during 
fabrication of the overpack.
    Response: The NRC agrees with the suggested change. The SAR has 
been revised accordingly.
    Comment No. 111: As it relates to the Radiography and Heat 
Treatment requirements for the containment boundary of the HI-STAR 
overpack, the applicant requested that post weld heat treatment (PWHT), 
after completing nondestructive examination, be used for all overpack 
containment boundary welds which require an exception from the ASME 
code.
    Response: The NRC agrees. The SAR and Appendix B to the CoC have 
been modified appropriately.
    Comment No. 112: The applicant suggested a revision to the drawings 
in the SAR to reflect the localized thinning tolerance in the 
containment shell.
    Response: The NRC staff agrees with the suggested revision. 
However, the applicant did not provide the suggested changes in its 
final revisions to the SAR. The initial drawings remain acceptable.
    Comment No. 113: One commenter (the applicant) recommended that 
changes to Technical Specification Table 4-1, MPC Enclosure Vessel and 
Lid, should be made to replace ``and sufficient intermediate layers to 
detect critical wild flaws'' with ``and at least one intermediate PT 
after approximately \3/8\ inch weld depth.'' The commenter also 
recommended the deletion of ``Flaws in austenitic stainless are not 
expected to exceed the bead''. The commenter further recommended 
several changes to the SER as follows: SER Section 8.1.4 should be 
changed to add ``(or optional multi-layer PT examination),'' after 
``ultrasonic examination (UT)''; the SER should recognize that users 
may choose to perform the MPC void-to-shell weld manually; and SER 
Section 11.4.1.3.1 should be reworded to read ``examined using UT or 
multi-layer PT techniques,'' instead of ``volumetrically examined using 
UT.''
    Response: The NRC agrees and notes that the applicant's comments 
with respect to TS Table 4-1 have been superseded by its latest 
revision to the SAR. Changes have been made to Table 1-3 to Appendix B. 
The SER has been revised as recommended.

Summary of Final Revisions

    The NRC staff modified the listing for the Holtec International HI-
STAR 100 cask system within 10 CFR 72.214, ``List of approved spent 
fuel storage casks,'' with respect to the title of the SAR as well as 
the CoC and its two appendices, the TSs, and the Approved Contents and 
Design Features. The NRC staff has also modified its SER.

Agreement State Compatibility

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

Finding of No Significant Environmental Impact: Availability

    Under the National Environmental Policy Act of 1969, as amended, 
and the Commission's regulations in Subpart A of 10 CFR part 51, the 
NRC has determined that this rule is not a major Federal action 
significantly affecting the quality of the human environment and 
therefore an environmental impact statement is not required. This final 
rule adds an additional cask to the list of

[[Page 48273]]

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 Stan 
Turel, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone (301) 415-6234, 
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 USC 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-STAR 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 
rulemaking 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 rule. Therefore, a backfit analysis is not required.

List of Subjects in 10 CFR Part 72

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

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

[[Page 48274]]

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 1008 is added to 
read as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
Certificate Number: 1008
SAR Submitted by: Holtec International
SAR Title: HI-STAR 100 Cask System Topical Safety Analysis Report
Docket Number: 72-1008
Certification Expiration Date: (20 years after final rule effective 
date)
Model Number: HI-STAR 100

    Dated at Rockville, Maryland, this 23rd day of August, 1999.

    For the Nuclear Regulatory Commission.
William D. Travers,
Executive Director for Operations.
[FR Doc. 99-23075 Filed 9-2-99; 8:45 am]
BILLING CODE 7590-01-P