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List of Approved Spent Fuel Storage Casks: Holtec HI-STORM 100 Addition
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: May 1, 2000 (Volume 65, Number 84)]
[Rules and Regulations]
[Page 25241-25265]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr01my00-3]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 72
RIN 3150-AG 31
List of Approved Spent Fuel Storage Casks: Holtec HI-STORM 100
Addition
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to add the Holtec HI-STORM 100 cask system to the list of
approved spent fuel storage casks. This amendment allows the holders of
power reactor operating licenses to store spent fuel in this approved
cask system under a general license.
EFFECTIVE DATE: This final rule is effective on May 31, 2000.
FOR FURTHER INFORMATION CONTACT: Merri Horn, telephone (301) 415-8126,
e-mail mlh1@nrc.gov of the Office of Nuclear Material Safety and
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.
SUPPLEMENTARY INFORMATION:
Background
Section 218(a) of the Nuclear Waste Policy Act of 1982, as amended
(NWPA), requires that ``[t]he Secretary [of Energy] shall establish a
demonstration program, in cooperation with the private sector, for the
dry storage of spent nuclear fuel at civilian nuclear reactor power
sites, with the objective of establishing one or more technologies that
the [Nuclear Regulatory] Commission may, by rule, approve for use at
the sites of civilian nuclear power reactors without, to the maximum
extent practicable, the need for additional site-specific approvals by
the Commission.'' Section 133 of the NWPA states, in part, ``[t]he
Commission shall, by rule, establish procedures for the licensing of
any technology approved by the Commission under Section 218(a) for use
at the site of any civilian nuclear power reactor.''
To implement this mandate, the NRC approved dry storage of spent
nuclear fuel in NRC-approved casks under a general license, publishing
a final rule in 10 CFR Part 72 entitled, ``General License for Storage
of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990).
This rule also established a new Subpart L within 10 CFR Part 72
entitled, ``Approval of Spent Fuel Storage Casks,'' containing
procedures and criteria for obtaining NRC approval of dry storage cask
designs.
Discussion
This rule will add the Holtec HI-STORM 100 cask system to the list
of NRC approved casks for spent fuel storage in 10 CFR 72.214.
Following the procedures specified in 10 CFR 72.230 of Subpart L,
Holtec International submitted an application for NRC approval with the
Safety Analysis Report (SAR) entitled ``Topical Safety Analysis Report
for the HI-STORM 100 Cask System.'' The NRC evaluated the Holtec
International submittal and issued a preliminary Safety Evaluation
Report (SER) and a proposed Certificate of Compliance (CoC) for the
Holtec HISTORM 100 cask system. The NRC published a proposed rule in
the Federal Register (64 FR 51271; September 22, 1999) to add the
Holtec HI-STORM 100 cask system to the listing in 10 CFR 72.214. The
comment period ended on December 6, 1999. Four comment letters were
received on the proposed rule.
Based on NRC review and analysis of public comments, the NRC staff
has modified, as appropriate, its proposed CoC, including its
appendices, the Technical Specifications (TSs), and the Approved
Contents and Design Features, for the Holtec HI-STORM 100 cask system.
The NRC staff has also modified its preliminary SER. Finally, comments
were received from other industry organizations suggesting changes to
the TSs and the Approved Contents and Design Features. Some of these
were editorial in nature, others provided clarification and
consistency, and some reflected final refinements in the cask design.
The NRC staff agrees with many of these suggested changes and has
incorporated them into the final documents, as appropriate. The NRC
staff has also modified the rule language by changing the word
``Certification'' to ``Certificate'' to clarify that it is actually the
Certificate that expires.
The NRC finds that the Holtec International HI-STORM 100 cask
system, as designed and when fabricated and used in accordance with the
conditions specified in its CoC, meets the requirements of 10 CFR Part
72. Thus, use of the Holtec HI-STORM 100 cask system, as approved by
the NRC, will provide adequate protection of public health and safety
and the environment. With this final rule, the NRC is approving the use
of the Holtec HI-STORM 100 cask system under the general license in 10
CFR Part 72, Subpart K, by holders of power reactor operating licenses
under 10 CFR Part 50. Simultaneously, the NRC is issuing a final SER
and CoC that will be effective on May 31, 2000. Single copies of the
CoC and SER are available for public inspection and/or copying for a
fee at the NRC Public Document Room, 2120 L Street, NW (Lower Level),
Washington, DC.
Summary of Public Comments on the Proposed Rule
The NRC received four comment letters on the proposed rule. The
commenters included a industry users group, two members of the public,
and a State. Copies of the public comments are available for review in
the NRC Public Document Room, 2120 L Street, NW (Lower Level),
Washington, DC 20003-1527.
Comments on the Holtec HI-STORM 100 Cask System
The comments and responses have been grouped into eleven areas:
General, radiation protection, accident analysis, criticality, design,
welds, structural, materials, thermal, technical specifications, and
miscellaneous. Several of the commenters provided specific comments on
the draft CoC, the NRC staff's preliminary SER, the TSs, and the
applicant's SAR. Some of the editorial comments have been grouped.
[[Page 25242]]
To the extent possible, all of the comments on a particular subject are
grouped together. The listing of the Holtec HI-STORM 100 cask system
within 10 CFR 72.214, ``List of approved spent fuel storage casks'' has
not been changed as a result of the public comments. A review of the
comments and the NRC staff's responses follow.
A. General
Comment A.1: One commenter expressed concern over the number of
cask designs being certified because there would be more problems and a
lack of standardization and integration in the country's total waste
system. The commenter stated that this amendment would change existing
environmental concerns as it would add one more design, complicating
the waste system for workers at a plant. The commenter asked how many
designs would be certified by the NRC and how many designs could be
used at one plant. Additional designs add to more mistakes and human
error because each design has different fabrication criteria and
handling procedures.
Response: These comments are beyond the scope of this rule that is
focused solely on whether to add a particular cask design, the Holtec
HI-STORM 100 cask system, to the list of approved casks. Pursuant to
the general license, each licensee must determine whether or not the
reactor site parameters are encompassed by the cask design bases
considered in the cask SAR and SER. Further, each general licensee must
document this determination in accordance with 10 CFR 72.212.
Comment A.2: One commenter stated that the tiered environmental
impact statement (EIS) is outdated for current dry cask design and
should be redone, particularly looking at terrorism and sabotage at an
independent spent fuel storage installation (ISFSI).
Response: The NRC disagrees with the comment. The environmental
assessment (EA) and finding of no significant impact (FONSI) prepared
as required by 10 CFR Part 51 conform to National Environmental Policy
Act (NEPA) procedural requirements. Tiering on past EISs and EAs is a
standard process under NEPA. As stated in the Council on Environmental
Quality's 40 Frequently Asked Questions, the tiering process makes each
EIS/EA of greater use and meaning to the public as the plan or program
develops without duplication of the analysis prepared for the previous
impact statement.
The NRC reviewed potential issues related to possible radiological
sabotage of storage casks at reactor site ISFSIs in the 1990 rule that
added Subparts K and L to 10 CFR Part 72 (55 FR 29181; July 18, 1990).
The NRC still finds the results of the 1990 rule current and
acceptable. In addition, each Part 72 licensee is required by 10 CFR
73.51 or 73.55 to develop a physical protection plan for the ISFSI. The
licensee is also required to install systems that provide high
assurance against unauthorized activities that could constitute an
unreasonable risk to the public health and safety.
Comment A.3: One commenter questioned whether the NRC was including
interim storage away from reactors in the EA, such as at a Federal or
private storage site in Nevada or Utah. The commenter further
questioned whether it was the NRC's intent to include transfer and
storage at a second site in the EA. The commenter asked if the
certification covered use at an interim site in Nevada or Utah.
Response: The EA supports the generic use of the Holtec HI-STORM
100 cask system under a general license. The storage could occur at any
site that meets the definition of a general licensee under 10 CFR Part
72. The general licensee must evaluate the site to determine whether or
not the chosen site parameters are enveloped by the design bases of the
approved cask as required by 10 CFR 72.212(b)(3). The EA does not cover
transportation from one site to another.
Comment A.4: One commenter questioned whether the NRC claims to
have done research on the condition of spent fuel after 20 to 50 years
of storage at a reactor in pools and dry casks, after being unloaded
twice and being transported across the country. The commenter stated
that a detailed analysis of what can happen to spent fuel before it
gets to Nevada or Utah should be conducted by the NRC. The commenter
asked what the spent fuel will be like and what the potential
environmental impacts will be after the fuel is unloaded and
transported.
Response: The NRC staff has reviewed numerous research reports
regarding the long term condition of spent fuel in wet and dry storage.
Additionally, the NRC has ongoing confirmatory research with spent fuel
removed from dry storage after 10 to 20 years. Analysis of spent fuel
has included the loads from routine shipping; and the effects,
primarily due to vibration, were found to be negligible.
The HI-STORM 100 MPC is a dual-purpose canister. Once loaded in the
MPC, the fuel is not intended to be unloaded and reloaded as the
questioner suggests. The lid welding and testing requirements and the
structural and thermal analyses in the SAR give the NRC staff
reasonable assurance that cask confinement and fuel integrity will be
maintained under design basis normal, off-normal, or accident events.
Therefore, fuel unloading should not be necessary. Regardless of
whether unloading may be necessary, each cask user is required to
develop detailed site-specific unloading procedures. Proper unloading
does not cause any particular degradation to occur to the fuel.
Comment A.5: One commenter stated that the no action alternative
was acceptable because the NRC should not be certifying numerous
designs. The commenter stated that other agencies such as NWTRB, EPA,
OCRWM, and DOE should be contacted for their views on what happens to
the whole waste system as more designs are certified.
Response: The NRC disagrees with the comment. The NRC found no
inherent design features that would result in significant environmental
impacts and that the HI-STORM 100 design meets regulatory requirements.
Therefore, there is no basis for denial of the application. The NRC
does not limit the number or types of casks that may be certified. The
NRC is not required to contact the agencies mentioned by the commenter
and we have not specifically solicited their input. The commenter may
contact these other agencies if interested in their views.
Comment A.6: One commenter recommended finding a reference
(reference 1 on page 3-16 of the SER) that is more recent than 1962.
Response: The NRC disagrees with this comment. This reference
refers to the change of the coefficient of friction from static to
dynamic condition. The rational behind this engineering principle has
not changed with time.
Comment A.7: One commenter stated that the NRC should request
simpler designs because of material interactions instead of approving
designs with new materials that have never received long term testing
for material interactions.
Response: The NRC staff disagrees with this comment. The materials
used in casks are selected upon the basis of the needed properties.
Casks are constructed from a limited number of materials. The materials
used in the Holtec HI-STORM design have a long history of use in the
nuclear industry and the performance of those materials is well known.
Comment A.8: One commenter objects to site specific changes that
are made to generic designs.
Response: This comment is beyond the scope of this rule that is
focused solely on whether to place the HI-STORM 100 cask system on the
list of approved casks. Section 72.48 permits
[[Page 25243]]
changes to the spent fuel storage cask as described in the FSAR and
defines the conditions under which these changes may be made without
prior NRC approval.
Comment A.9: One commenter stated that it appeared that Holtec
split what appears to be one generic system into two separate rules and
asked why the system was not certified together. Systems should be
complete when they are proposed for rulemaking. The commenter further
stated that vendors should apply for storage and transport at the same
time and that NRC should not allow loading until the transportation
portion is certified.
Response: The NRC disagrees with the comment. The HI-STAR 100 Cask
System and HI-STORM 100 Cask System are two separate spent fuel storage
cask systems. Each is a complete spent fuel storage cask system that
satisfies the requirements of 10 CFR Part 72. Regarding the dual-
purpose (storage and transportation) use of a cask system or its
components, separate certifications are required for approval of a cask
design (or individual components such as a canister) under the
provisions of use for 10 CFR Parts 71 and 72. There is no regulatory
requirement that the certification be simultaneous.
Comment A.10: One commenter asked a number of site-specific
questions related to Private Fuel Storage's plans to use the Holtec HI-
STAR and HI-STORM cask systems at the Utah site. These issues related
to cask handling, dry transfer, sabotage scenarios, infrastructure for
unloading, etc. One commenter stated that they understood that Private
Fuel Storage plans to use the HI-STAR system for storage and transport
with the HI-STORM as a companion concrete overpack, that the metal HI-
STAR overpack would be used as a backup, and that the commenter
objected to these plans.
Response: The comment is beyond the scope of this rule that is
focused solely on whether to add a particular design, the Holtec HI-
STORM 100 cask system, to the list of approved casks. The rule will
enable licensees to use this cask system under the general license
provisions of 10 CFR Part 72. The rule does not address site-specific
issues related to potential users.
Comment A.11: One commenter objected to calling the cask a multi-
purpose cask (MPC) because that stands for storage, transport, and
disposal, and stated that the cask is not approved for these functions
which can cause confusion when real MPCs are certified.
Response: The NRC disagrees with the comment. The name or model
number given to the cask design is developed by the applicant. The CoC
for the Holtec HI-STORM 100 is intended for the interim storage of
spent fuel. The use of MPC in a dry storage cask application or an NRC
SER/CoC is not a certification under 10 CFR Part 71 for the transport
of radioactive materials or an approval for disposal at a high-level
waste repository.
Comment A.12: One commenter stated that Holtec should not be
allowed to approve its own suppliers and that the suppliers should be
ASME-approved.
Response: The NRC disagrees with the comment. NRC regulations do
not require an ASME stamp for a cask or the use of ASME-approved
suppliers. The design and fabrication requirements for a certified dry
cask storage system are described in 10 CFR Part 72 and the NRC staff's
Standard Review Plan, NUREG-1536, ``Standard Review Plan for Dry Cask
Storage Systems'' (SRP). Applicant submittals are reviewed to the
criteria in the SRP. Cask fabrication activities are audited by the
licensees and inspected by the NRC staff to ensure that components are
fabricated as designed. The CoC holder and licensee are responsible for
verifying that fabricators are qualified. The CoC holder and licensee
must have a Quality Assurance (QA) Program that has been approved by
the NRC as part of the licensing or CoC issue process. This QA program
must meet the requirements of 10 CFR 72.148 and 10 CFR 72.154 for the
selection of fabricators. Also, the procurement documents issued to the
fabricator must comply with 10 CFR 21.31. The licensee/CoC holder is
required to verify that all regulations and CoC conditions applicable
to the container are met. The NRC inspects the licensee/CoC holders and
fabricators to verify compliance. Additionally, many storage cask
fabricators are certified by the American Society of Mechanical
Engineers and are N-Stamp Certificate holders.
Comment A.13: One commenter stated that issues should not be
resolved in telephone conferences but in public meetings with a record
in the public document room.
Response: The NRC disagrees with the comment. Telephone conferences
are an important mode of communication with applicants and licensees
and enable the NRC staff to conduct its official business efficiently.
If, in these telephone conferences, the NRC staff receives information
that would form the basis for its regulatory decision, that information
is documented and made available for public inspection under 10 CFR
Parts 2 and 9.
Comment A.14: One commenter stated that all details of the design
should be finalized and open for public comment.
Response: The NRC disagrees that all design details need to be
finalized and open for public comment before a design is approved. The
NRC staff focuses its review on those design details that are
significant with respect to the health and safety of the public and/or
are required to make a regulatory finding. Design details that are
pertinent to the NRC staff's findings are finalized and made available
for public inspection and comment under 10 CFR Parts 2 and 9.
B. Radiation Protection
Comment B.1: One commenter objected to the use of less shielding
for the 100-ton transfer cask and allowing the utilities to perform a
cost-benefit analysis to justify the use of the 100-ton transfer cask
at the expense of the worker. The workers should receive the minimum
achievable dose and not the maximum allowable dose. The NRC should not
allow the use of the 100-ton transfer cask because the dose is 3 times
higher and workers should not be treated as guinea pigs. The commenter
stated that the utilities should be required to use the 125-ton
transfer cask which is safer and modify their facilities to accommodate
the transfer cask or choose a cask that works for their specific site
limitations because the utilities shouldn't limit the shielding for
workers.
Response: NRC disagrees with this comment. Each cask user will
operate the HI-STORM 100 under a 10 CFR Part 20 radiological protection
program. ALARA means making every reasonable effort to maintain
exposures to radiation as far below the dose limits while taking in
account the state of technology, the economics of improvements in
relation to the state of technology, and the economics of improvements
in relation to benefits to the public health and safety. As stated in
Section 2.0.3 of the SAR, the general licensee should utilize the 125-
ton transfer cask provided it is capable of using it. However,
licensees not capable of using the more shielded design may employ
ALARA considerations when evaluating whether to modify its plant or use
the 100-ton transfer cask. The NRC found this acceptable as discussed
in Section 10.2 of the SER.
Comment B.2: One commenter asked why the specific dose rate
criteria for the HI-TRAC was not given and indicated that the criteria
should be included.
[[Page 25244]]
Response: The applicant did not provide explicit dose rate values
as design criteria for the transfer cask designs, but stated that the
radiological requirements of 10 CFR Parts 72 and 20 as the overall
shielding design objectives for the cask system. The NRC found this
acceptable. The TSs in Appendix A of the CoC specify dose rate limits
for the transfer casks that are based on the applicant's shielding
calculations.
Comment B.3: One commenter questioned the bounding analysis for
cobalt impurities, asked how much cobalt is really in the fuel, and if
the quantity had been tested and verified for the real thing.
Response: The applicant's analysis of cobalt impurities is
discussed in Section 5.2.1 of the SER. The applicant showed that the
cobalt impurity values that are assumed in its shielding analyses were
appropriate based on industry data and analysis of post-irradiation
cooling of older fuel. The NRC found this acceptable. The cask user is
not required to measure the actual quantity of cobalt in its spent
fuel. The cask user will operate the cask under a 10 CFR Part 20
radiological protection program and verify that the cask system meets
the dose rate limits specified in the TSs.
Comment B.4: One commenter asked why backscattering was not
considered for all cask designs.
Response: This comment is beyond the scope of this rule that is
focused solely on whether to add a particular cask design, the Holtec
HI-STORM 100 cask system, to the list of approved casks. Note that
backscatter was considered for the Holtec HI-STORM 100 cask system.
Comment B.5: One commenter asked what are the various array
configurations allowed and what are the differences between them. The
commenter asked if the cask array is limited to two rows and for the
applicable NRC criteria.
Response: The use of the HI-STORM design is not limited to two
rows. The NRC requires the applicant to perform off-site dose
calculations from a typical ISFSI array to demonstrate that radiation
shielding features are sufficient to meet the radiological requirements
of 10 CFR Parts 72.104 and 72.106. As discussed in Section 5.3.1 of the
SER, the applicant used a two-row cask array model as part of its
methodology to estimate off-site dose rates. The values obtained by
this method can be applied to dose rate calculations for typical cask
arrays that may consist of multiple rows. NRC found the dose estimates
to be acceptable. Each general licensee will identify an ISFSI
configuration and perform a site-specific dose evaluation to
demonstrate compliance with Part 72 radiological requirements.
Comment B.6: One commenter asked why the dose rate for the bottom
of the MPC-68 was higher than for the MPC-24 when the dose rates at the
side and top were higher for the MPC-24. The commenter stated that the
trunnion doses showed that extreme care needs to be taken in those
areas and that the bottom doses are really high and don't get enough
attention.
Response: The applicant appropriately assumed design basis fuel
loadings for each canister and estimated dose rates at various
locations. The NRC notes that dose rates at the bottom of the canister
depend on several factors such as the fuel hardware characteristics,
irradiation and cooling history, and the relative position of each fuel
type within the cask system. The NRC found that the applicant
appropriately addressed these and other factors, and that the
calculated dose rates at the bottom and at the trunnions of the
transfer cask were acceptable. In addition, each cask user will operate
the HI-STORM 100 under a 10 CFR Part 20 radiological protection program
and monitor dose rates during loading and unloading.
Comment B.7: One commenter asked what the dose for the 2x5 cask
array was at 100 meters.
Response: Figure 5.1.3 of the SAR indicates that the dose rate for
a 2x5 array at 100 meters is approximately 600 to 700 mrem/yr assuming
a design basis fuel loading and 100 percent occupancy. Each general
licensee will identify an ISFSI configuration and perform a site-
specific dose evaluation, based partly on site-specific
characteristics, to demonstrate compliance with Part 72 radiological
requirements.
Comment B.8: One commenter asked why other cask designs do not
account for approximate atmospheric conditions. The commenter also
asked the conditions of weather or location for which the air density
decreases.
Response: Atmospheric density changes daily. The measure of the
density is provided by local weather forecasters through the barometric
pressure. When a high pressure front passes an area, the air density is
greater than when a low pressure weather front passes the same
location.
The comment concerning other cask designs is beyond the scope of
this rule that is focused solely on whether to place the Holtec HI-
STORM 100 cask system on the list of approved casks. For the HI-STORM
100, each general licensee should consider atmospheric conditions
relevant to its ISFSI as indicated in Section 5.4.2 of the SER.
Comment B.9: One commenter asked how much the releases from dry
storage add to the effluent from a reactor site and the duration of a
release, and what happens to the cask and fuel during the release.
Response: Specific effluent releases from reactors operated by
general licensees are beyond the scope of this rule. However, NRC does
not expect any effluent release from the HI-STORM 100 under credible
conditions. Design basis public exposures from direct radiation and
hypothetical releases are discussed in SER Sections 10.4 and 10.5.
Comment B.10: One commenter approved of the condition in Appendix B
of the CoC regarding the evaluation of engineering features (e.g. berm)
that are used for radiological protection by the user.
Response: No response is necessary.
Comment B.11: One commenter stated that average surface dose rates
in TS 3.2.1 for transfer cask dose rates should not be used, that the
highest value should be used, and the limit should not be exceeded. The
commenter also asked why the side dose rates are measured along the
middle of the flat surface section of the neutron shield rather than on
the radial steel fins where dose rates are assumed by the commenter to
be higher.
Response: The NRC disagrees with the comment. The specification of
surface average dose rates and the measuring locations on the side of
the neutron shield are consistent with health physics methods that are
used to characterize radiation fields around a cask. The measuring
locations are also consistent with the dose rate calculations presented
in the applicant's shielding analysis. The cask user will operate the
HI-STORM 100 under a 10 CFR Part 20 radiological protection program.
NRC has reasonable assurance that the general licensee's radiological
protection and ALARA program will detect and mitigate exposures from
the radiation fields that are expected during operation of the HI-STORM
100 system.
Comment B.12: One commenter asked why the dose rate for the bottom
of the transfer cask is not provided in TS 3.2.1 and what is that dose
rate.
Response: Dose rate limits for the bottom of the transfer casks are
not needed because they would not provide a significant benefit in
ensuring compliance with regulatory limits on occupational dose and
dose to the public. The dose limits at the top and side of the transfer
casks are adequate to help ensure that the cask system is
[[Page 25245]]
safely operated in compliance with 10 CFR Part 20 and Part 72.
Calculated dose rates at the bottom of the transfer casks are reported
in Sections 5.1 and 5.4 of the SAR.
Comment B.13: One commenter recommended that Section 5.1.2 of the
SER be revised to clarify that overpack surface dose rates are design
objectives and are shown to be met by analysis, and that the TSs are
equal to or more conservative than the design objectives.
Response: The NRC disagrees with this comment. The NRC staff does
agree that the vent dose rates calculated by the applicant are
significantly less than the applicant's proposed design criteria.
However, the differences between the calculated vent dose rates and the
proposed design criteria are not relevant to the bases and findings in
the SER. The TSs in Appendix A of the CoC specify vent dose rate limits
for the overpack that are based on the applicant's shielding
calculations. Therefore, a revision to the SER to reflect the dose rate
difference is not necessary.
Comment B.14: One commenter recommended that Section 5.4.11 and
Table 5.4-1 of the SER be clarified to indicate that the dose rates are
not peak or maximum values.
Response: The NRC agrees with the comment. The SER has been
clarified to state the vent dose rates are average over the area of the
vent opening. A footnote has been added to Table 5.4.1 to clarify
values are average over surface detector areas.
Comment B.15: One commenter recommended that Section 10.5.1 of the
SER be revised to indicate that the maximum MPC leak rate is utilized
in the calculations.
Response: The NRC agrees with the comment. The SER text has been
revised accordingly.
Comment B.16: One commenter indicated there was an inconsistency
between the accident condition whole body and thyroid dose values
referenced in Chapter 11 of the draft SER and the dose values
calculated in Section 7 of the applicant's SAR.
Response: The NRC agrees with the comment. The SER has been revised
to indicate the correct whole body and thyroid dose values calculated
by the applicant. The accident condition whole body total effective
dose equivalent (TEDE) is 44.1 mrem and the thyroid dose is 4.1 mrem.
Comment B.17: One commenter objected to the use of a 30-day
duration of a radiological release during an accident. The commenter
noted that this assumption is stated in Interim Staff Guidance 5 but
that it is not justified in the guidance or any accompanying report.
The commenter pointed out that NRC regulations for ISFSIs do not
require offsite emergency planning, or planning for the ingestion
pathway zone, and therefore, there is no basis for assuming that
something happens within 30 days to stop the release.
Response: The NRC disagrees with the comment. As indicated in ISG-
5, Rev.1, the 30-day assumption is consistent with the time period that
is used to demonstrate compliance with radiological dose requirements
associated with reactor facilities that operate under 10 CFR Part 50.
The applicant specified corrective actions for each accident in Chapter
11 of the SAR. NRC believes that these corrective actions can be
reasonably achieved within 30 days. Although NRC does not expect
effluent release from the HI-STORM 100 under credible accident
scenarios, the 30-day assumption in the analysis is acceptable because
the NRC staff has reasonable assurance that in the 30-day timeframe
adequate protective measures can and will be taken for the public in
the event of a radiological emergency. These protective measures
include implementation of the general licensee's Part 50 emergency
plan, evacuation of the surrounding public, and mitigation of
radiological ingestion pathways.
Comment B.18: One commenter objected to the assumption that a
person at the fence post (500 meters) would be exposed for only 2000
hours/year which is the number of working hours in a year. The
commenter stated that 8,760 hours/year should be used because a
licensee can not control who would be in the area outside the fence or
how long they would be there. For conservatism, the applicant should
have assumed that people, such as mothers with pre-school aged
children, the elderly, ranchers, and farmers are present at the fence
post day-long and year-round.
Response: The NRC agrees that 8,760 hours/year should be used and
notes that Section 7.2.9 of the HI-STORM SAR explicitly states that:
``The individual at the site boundary is exposed for 8,760 hours
[7.0.2].'' The NRC staff's independent calculations confirmed Holtec's
calculated results, as stated in the NRC staff's SER. In addition,
Section 7.2.9 also assumed in its calculations that: ``The distance
from the cask to the site boundary is 100 meters.'' With respect to
hypothetical individual exposed at the site boundary, the methods used
in the dosage calculations cover children, the elderly, ranchers,
farmers, etc. The overall public dose limit is protective of all
individuals because the variation of sensitivity with age and gender
was accounted for in the selection of the lifetime risk limit, from
which the annual public dose limit was derived.
The NRC continues to believe that the existing regulations and
approved methodologies adequately address public health and safety. The
issue of dose rates to children was addressed in the Federal Register
on May 21, 1991 (56 FR 23387).
Comment B.19: One commenter stated that the dose due to direct
gamma and ingestion of radionuclides should be considered in the dose
calculation because to ignore these pathways underestimates the dose.
The commenter further objected to the NRC staff stating (in the Holtec
HI-STAR 100 final rule) that these pathways would be addressed in the
general licensee's site-specific review. The commenter stated that
there is no regulatory requirement for these actions to be taken by the
general licensee. The commenter stated that it is misleading for the
applicant to do a calculation that provides a reassuring result, based
on assumptions that have nothing to do with the real requirements of
the regulations because licensees tend to rely heavily on the generic
analyses that have been performed by cask manufacturers.
Response: The NRC disagrees with the comment. Although the NRC does
not expect effluent release from the HI-STORM 100 under credible
conditions, the applicant's method used to determine design basis dose
rates from a hypothetical release are adequate to demonstrate that the
confinement features are sufficient to meet the radiological
requirements of 10 CFR 72.106. The NRC staff believes the methods
applied by the applicant conservatively bound hypothetical dose rates
to the general public. Further, 10 CFR 72.212(b)(6) requires the
general licensee to review its reactor emergency plan and radiation
protection program to determine its effectiveness and make changes if
necessary when using a cask listed in 10 CFR Part 72, Subpart L.
Comment B.20: One commenter stated that the thyroid and whole body
doses should consider chlorine-36 (Cl-36) because it will be present in
the irradiated fuel and will significantly contribute to the dose. The
commenter points out that the Department of Energy acknowledges that
Cl-36 is one of the significant radionuclides in Appendix A, of the
Yucca Mountain Draft EIS.
Response: The NRC disagrees with the comment. The NRC staff's
independent analysis of the thyroid and whole body dose was based on
independent
[[Page 25246]]
calculations using the ORIGEN computer code, as referenced by the
commenter. The calculated contribution of the chlorine gas was below
the truncation limit used in the calculation. Cl-36 has an
inconsequential contribution on the total dose to an individual.
C. Accident Analysis
Comment C.1: One commenter asked if lead could be a missile strike
barrier from a tornado or from current weapons. The commenter asked if
missiles could penetrate the transfer cask and canister inside, and
when the missile strike is assumed to occur (i.e. when a loaded
transfer cask is on top of the overpack.) The commenter stated that
this needs to be updated and evaluated.
Response: The lead backed outer shell of HI-TRAC has been evaluated
for the required tornado missile strike. The analysis shows that there
is no penetration consequence that would lead to a radiological
release. The threat of missiles from weapons is beyond the scope of
this rule.
Comment C.2: One commenter expressed concern that the transfer cask
is a real target on top of the storage cask and asked if it had been
fully evaluated for terrorism and sabotage, particularly when it was on
top of the storage cask. The commenter asked if the overpack was put in
place while on the pad; the commenter felt that this would be a target
for terrorists. The commenter asked if the transfer cask, with inner
cannister inside, could be knocked off by a terrorist blast and fall,
crash, or roll into other casks or be upended so that the fuel is
upside down.
Response: The performance of the transfer cask in a sabotage or
terrorist event was not evaluated. The threat of terrorism or sabotage
is beyond the scope of this rule. See also the response to C.8.
Comment C.3: One commenter asked if the seismic event was based on
the actual pad analysis and not the reactor building seismic analysis
because the conditions between the reactor building and pad location
could significantly differ.
Response: The storage pad is a site-specific issue and is beyond
the scope of this cask design rule. Under 10 CFR 72.212, the cask
operators are required to perform written evaluations to ensure that
storage pads have been designed to adequately support the stored casks.
The licensee using a particular cask design has the responsibility
under the general license to evaluate the match between reactor site
parameters and the range of site conditions (i.e. the envelope)
reviewed by the NRC for an approved cask.
Comment C.4: One commenter asked how a full cask array would behave
in a seismic event. The commenter asked what buildings or equipment are
allowed on the pad that could crash into the casks during a seismic
event, such as the transfer equipment. The commenter asked if a crack
or ``push up'' of the pad could cause the cask to roll (down an incline
or into water).
Response: The SAR indicates that the HI-STORM 100 overpack will
neither slide nor tip over due to a seismic event with the design-basis
earthquake input listed in Section 3.4.2 of the SER. The use of a
general licensed cask by a utility requires that the user ensure that
the site is not subject to any potential accident that has not been
analyzed for the general license. This would include any potential
design basis earthquakes that were not enveloped by the NRC SER for the
cask or any site conditions associated with the actual pad and cask
locations that could affect the cask design.
Comment C.5: One commenter asked what the design-basis earthquake
on top of the surface pad was and where it occurred. The commenter
questioned why the bottom surface was not evaluated because the ground
can push up and crack or cause heaving in the concrete and how the
condition of the bottom surface is known.
Response: The design basis earthquake is the most severe earthquake
that has been historically reported for a particular site and
surrounding area, with sufficient margins for the limited accuracy,
quantity, and period of time in which historical data have been
accumulated. Structure, systems, and components important to safety are
designed to withstand the effects of this earthquake without loss of
capability to perform their safety functions. The design basis
earthquake is described by an appropriate response spectrum anchored at
the peak ground acceleration. The response is then amplified through
the pad to obtain the input response spectrum at the top of the pad (or
at the bottom of the cask) for cask seismic evaluation. Soil and
storage pad interaction is a site-specific issue that will be addresses
in the cask user's 10 CFR 72.212 evaluation and is beyond the scope of
this rule.
Comment C.6: One commenter asked what happens if the pad is cracked
and heaving up as the cask is tipping over because a tornado or seismic
event will likely affect both the pad and the casks.
Response: The NRC does not consider the scenario described by the
commenter to be credible. The evaluation in Section 3 of the SAR shows
that tipover will not occur. However, as a defense-in-depth measure,
cask tipover is also evaluated in Section 3 of the SAR and discussed in
Section 3.4.2 of the SER.
Comment C.7: One commenter asked if the cask could become upside
down in a tornado or seismic event and if it happened would the top of
the fuel hit the underside of the MPC lid with the weight on the
overpack lid studs.
Response: The HI-STORM 100 overpack is evaluated for tornado,
tornado missiles, and seismic events in Section 3 of the SAR. The
results indicate that the cask will not tip over. Therefore, the cask
will not become upside down.
Comment C.8: One commenter stated that an airplane crash with its
fuel fire should be evaluated, including crash into a full cask array,
damage to the pad, and a fuel and airplane explosion after the crash.
The commenter stated that an anti-missile device with an incendiary
device and a truck bomb should be analyzed for the cask transfer
facility (CTF).
Response: The NRC disagrees with the comment. Before using the HI-
STORM 100 casks, the general licensee must evaluate the site to
determine whether or not the chosen site parameters are enveloped by
the design bases of the approved casks as required by 10 CFR
72.212(b)(3). The licensee's site evaluation should consider the
effects of nearby transportation and military activities.
The NRC reviewed potential issues related to possible radiological
sabotage of storage casks at reactor site ISFSIs in the 1990 rule that
added Subparts K and L to 10 CFR Part 72 (55 FR 29181; July 18, 1990).
NRC regulations in 10 CFR Part 72 establish physical protection
requirements for an ISFSI located within the owner-controlled area of a
licensed power reactor site. Spent fuel in the ISFSI is required to be
protected against radiological sabotage using provisions and
requirements as specified in 10 CFR 72.212(b)(5). Further, specific
performance criteria are specified in 10 CFR Part 73. Each utility
licensed to have an ISFSI at its reactor site is required to develop
physical protection plans and install systems that provide high
assurance against unauthorized activities that could constitute an
unreasonable risk to public health and safety.
The physical protection systems at an ISFSI and its associated
reactor are similar in design features to ensure the detection and
assessment of unauthorized activities. Alarm
[[Page 25247]]
annunciations at the general license ISFSI are monitored by the alarm
stations at the reactor site. Response to intrusion alarms is required.
Each ISFSI is periodically inspected by NRC. Also, the licensee
conducts periodic patrols and surveillances to ensure that the physical
protection systems are operating within their design limits. The ISFSI
licensee is responsible for protecting spent fuel in the casks from
sabotage not the certificate holder. Comments on the existing
regulations specifying what type of sabotage events must be considered
are beyond the scope of this rule.
Comment C.9: One commenter questioned why the tornado missile test
simulated a pulse impact of a vehicle and stated that a sharp object
such as a metal pole or other items that might be in the vicinity of a
real pad would do more penetration damage.
Response: In addition to the 4,000-pound automobile impacting at a
126 mph velocity, the SAR also provided analyses for two more missiles
impacting at 126 mph velocity: a 1-in diameter steel sphere and an 8-in
diameter rigid cylinder. Results of the analyses show that the 4,000
pound automobile produces the highest impact force on the cask because
it has the largest mass. Based on these results, the NRC staff has
reasonable assurance that the 4,000 pound vehicle bounds the effect of
other credible types of tornado missiles.
Comment C.10: One commenter stated that the 15-minute transporter
fire is not valid. A big plane crash with its fuel should be evaluated
as well as a sabotage missile penetration with an incendiary device.
Response: The NRC disagrees with this comment. The basis for the
4.8-minute fire (not a 15 minute fire, see response to comment C.18) is
associated with the time it would take to burn 50 gallons of fuel,
presumably carried by the transporter. The CoC, Appendix B, Section
3.4, states that ``the on-site transporter fuel tank will contain no
more than 50 gallons of diesel fuel while handling a loaded OVERPACK or
TRANSFER CASK.'' Other modes of transport causing the fire (e.g.,
airplanes, trains, delivery trucks or missiles) are not considered as
plausible and are beyond the scope of this rule. Before using the HI-
STORM casks, the general licensee must evaluate the site to determine
whether or not the chosen site parameters are enveloped by the design
bases of the approved cask as required by 10 CFR 72.212(b)(3). Included
in this evaluation is the verification that no credible source of an
external explosion that would produce an external pressure above that
analyzed in the SAR and that any cask handling equipment used to move
the HI-STORM cask to the pad is limited to 50 gallons of fuel (refer to
CoC, Appendix B, Section 3.4--Site Specific Parameters and Analyses).
Comment C.11: One commenter asked why there were no calculations
for the bottom plate, overpack lid, etc. in a fire because the
temperatures of these plates were important to know and could affect
the pad or fire fighting equipment.
Response: The applicant did calculate the temperatures for the
bottom plate and the overpack lid. However, these temperatures were not
reported in the SAR. Not all calculated temperatures are reported in
the SAR. With respect to the impact of fire on the pad or fire fighting
equipment, a postulated 50 gallon fuel source would have minimal impact
on those components. The heat generated by the pool of fuel is directed
upward where the fuel is in a gaseous state. The limiting temperatures
will occur above the surface of the concrete pad. Because the fuel has
to vaporize in order to burn, the liquid fuel on the concrete will have
minimal impact on the bottom plate of the overpack lid (in a liquid
state, the fuel is cool). The duration of the fire is less than 4
minutes. The impact on the fire fighting equipment would be minimal, if
any. Table 4-3 of the SER was modified to indicate that the
temperatures were not reported.
Comment C.12: One commenter asked how the 45,000 MWD/MTU for 5
years related to the sabotage and terrorist evaluation for radiation
disposal and stated that the evaluation is outdated.
Response: The comment on the sabotage report is beyond the scope of
this rule. See the discussion in the response to C.8.
Comment C.13: One commenter asked if the water jacket could be
pierced with an anti-missile gun or if a terrorist could shoot the
jacket full of holes, and what are the consequences if these events did
occur.
Response: The specific threat of an anti-missile gun or other small
arms against the HI-STORM 100 is beyond the scope of this rule.
However, the resultant dose rate for an assumed complete loss of the
water jacket is addressed in Section 5.1.2 of the SAR. The analysis
indicates that the off-site dose at 100 meters will be below the 5 rem
accident limit in 10 CFR 72.106.
Comment C.14: One commenter asked why a burial under a landslide
during a seismic event is not considered.
Response: Burying a cask due to seismic event, landside, or tornado
is considered a very unlikely event. Considering the unlikeliness of
the event coupled with the casks being able to withstand these events
make burying and any adverse consequence in the opinion of the NRC not
credible.
Comment C.15: One commenter asked why a vertical drop of a loaded
transfer cask is not considered a credible accident, particularly as it
is perched on top of the concrete overpack to load.
Response: A vertical drop of a transfer cask is not considered
credible because vertical lifting of a loaded transfer cask must be
performed with structures and components designed to prevent a cask
drop. The criteria for those structures and components are specified in
Section 3.5 of Appendix B to CoC No. 1014. The restrictions on vertical
lifting are specified in Section 5.5 of the TSs (Appendix A to the
CoC).
Comment C.16: One commenter stated that defense-in-depth is needed
for sabotage events which could cause a tipover.
Response: The NRC disagrees with this comment. Sabotage events are
beyond the scope of this rule. They are considered in Part 73.
Furthermore, the SAR demonstrates that the HI-STORM 100 overpack will
not tipover due to a design basis accident. However, as an added
defense-in-depth measure, cask tipover is evaluated in Section 3 of the
SAR and discussed in Section 3.4.2 of the SER.
Comment C.17: One commenter asked why a postulated explosion from a
truck bomb at the pad fence was not evaluated.
Response: The specific threat of a truck bomb is beyond the scope
of this rule. The response to C.8 addresses radiological sabotage of
storage casks at generally licensed ISFSIs.
Comment C.18: One commenter asked the basis for the 217-second fire
for the overpack and the 4.8-minute fire for the transfer cask. The
commenter also asked if the NRC assumed that nothing on the vehicle or
in the vicinity (such as grass or trees or other structures) will burn
and cause the fire to burn longer.
Response: The duration of a fire burn is based on several factors.
One factor is the rate at which the fire burns, normally categorized as
inches of fuel burned per minute. The burn rate (inches per minute) is
the same for both the overpack and the transfer cask because the source
of fuel is the same (e.g., diesel fuel). The duration of the burn comes
from the postulated depth of the pool of fuel. A conservative estimate
of the time of burn is to assume that the spilled fuel does not extend
beyond 1 meter of the surface of the overpack or the surface of the
transfer cask. (In reality, the fuel will spill significantly farther
than one meter on
[[Page 25248]]
a flat surface, just as spilling a bucket of water on the ground, and
will not accumulate to any significant depth which creates a shorter
fire burn time.) Because the outer diameter of the overpack and the
outer diameter of the transfer cask are different, the postulated depth
or height that 50 gallons of fuel is postulated to reach will differ
for the two cases. The case with the higher column of fuel will burn
longer. Because the surface area of the pool of fuel for the overpack
is 1.3 times larger than for the transfer cask, the pool of fuel for
the overpack will be lower (given the same volume of available fuel,
e.g., 50 gallons). A lower pool of fuel will burn quicker. (Note that
the burn rate is in inches of fuel per minute, and a smaller column of
fuel will burn quicker than a higher column of fuel). Therefore, the
burn time for the overpack is shorter than the burn time for the
transfer cask.
With respect to other flammable sources that could catch fire,
before using the HI-STORM cask, the general licensee must evaluate the
site to determine whether or not the chosen site parameters are
enveloped by the design bases of the approved cask as required by 10
CFR 72.212(b)(3). Included in this evaluation is the verification that
the cask handling equipment used to move the concrete cask to the pad
is limited to 50 gallons of fuel (refer to CoC, Appendix B, Section
3.4.5) and that the assumptions used in the SAR bound the consequences
for the proposed site. Additional assessments would have to be
performed if other sources are identified that could result in a more
limiting fire.
Comment C.19: One commenter objected to the use of the leakage rate
used by Holtec because it is based on an analysis of a transportation
cask rather than a storage cask, for which the NRC and industry have
different design and testing requirements. The commenter noted that the
small assumed leakage rate and calculation methodology in NUREG/CR-6487
are based on ANSI standard N14.5 for transportation casks. ANSI N14.5
assumes that casks will be leak-tested periodically before shipment and
after maintenance and repair. The commenter pointed out that some
ISFSIs have no design provisions for testing helium leakage during
storage and no provisions for repairing and maintaining casks and
testing for leakage after repair and maintenance. Therefore, it is
inappropriate to assume that these storage casks will have the same
small leakage rate as transportation casks for which leakage potential
is designed and planned to be monitored. The commenter stated that
neither Holtec nor the NRC has any basis for relying on NUREG-1617 to
assume a small leakage rate in a storage cask breach.
Response: The NRC disagrees with the comment. The ANSI N14.5
standard was developed to determine allowable leak rates for shipping
packages that employ mechanical seals, which typically undergo
repetitive use. Periodic testing is prescribed for the mechanical seal
to ensure it has not degraded from repetitive use and/or seal
maintenance. The analytic technique in ANSI N14.5 that is used to
determine a leak rate across an assumed leak path is valid for
determining an assumed leak rate across the confinement boundary of a
welded canister. An off-site dose can be subsequently calculated using
standard atmospheric dispersion principles and assuming a partial
release of the cask constituents at the calculated leak rate. The
welded closure is leak-tested to a sensitivity equal to the calculated
leak rate to ensure integrity of the confinement system before storage
operations. Periodic testing of the confinement boundary is not
applicable because the welded confinement boundary is designed to
remain intact during normal, off-normal, and accident conditions for
the lifetime of the canister.
Comment C.20: One commenter stated that the methodology used in
NUREG/CR-6487 may not apply for accidents that exceed the design basis
accident. The allowed leak hole size can easily be exceeded in
accidents involving sabotage such as an impact with a MILAN or TOW-2
hand held anti-tank device, a jet engine, or military ordnance.
Response: The NRC disagrees with the comment. Consideration of
accidents that exceed design basis is not required by 10 CFR Part 72
and is beyond the scope of the NRC staff's review. The threat of
accidents involving sabotage is beyond the scope of this rule. Sabotage
issues are covered by 10 CFR Part 73.
Comment C.21: One commenter stated that Holtec should consider a
300 gallon fire or a 6,000 gallon fire. The commenter stated that these
are credible accidents at an ISFSI and should be considered. A heavy
haul tractor carries 300 gallons of fuel and is likely to be used at
ISFSIs. Locomotives that carry casks to ISFSIs may carry 6,000 gallons
of diesel fuel.
Response: The NRC disagrees with the comment. The analysis need
only address the maximum permissible source of fuel at the storage site
near the HI-STORM 100 system (10 CFR Part 72). Section 3.4 of Appendix
B to the CoC limits the source of fuel near the HI-STORM 100 system to
50 gallons. Licensees are required to verify that all conditions of the
CoC are met.
Comment C.22: One commenter stated that Holtec's fire analysis is
deficient because the fire calculations assume that the fire takes
place outside the concrete storage cask and does not consider the
possibility of a fuel fire being drawn into the intake vent of the HI-
STORM 100 cask.
Response: The NRC staff disagrees with the comment. The purpose of
the fire analysis is to assess the consequences of a postulated fire on
the HI-STORM system. The elements of interest are the impact of the
fire on the peak clad temperature and the impact of the fire on the
system materials. A 50-gallon fuel supply will have a very short burn
duration. Applying the conservative assumptions of 10 CFR Part 71, a
50-gallon fuel supply would theoretically result in a pool size of 0.54
inches if limited to a one-meter spread around the overpack. The burn
duration of the fuel in this configuration is 3.6 minutes. This burn
duration will have insignificant impact on the peak clad temperature.
The heat capacity of the system is too great to have an appreciable
feedback on the peak clad temperature for a short duration transient.
The greatest impact of a fire will be felt on the overpack. A bounding
analysis was performed on the overpack by imposing a maximum burn
temperature (specified in 10 CFR Part 71) on the entire outer surface
of the overpack. This maximizes the impact on the steel liner and the
concrete. In a less conservative calculation, the maximum temperature
will be limited to the lower portion of the overpack. For additional
conservatism, the applicant increased the inside temperature of the
overpack to 300 deg.F to account for heating of the air as it passes
through the vents. As illustrated in SAR Figures 11.2.1--11.2.5, this
bounding calculation illustrates that only the outer boundary of the
concrete exceeds the temperature limit for concrete. At a depth of one
inch into the concrete the temperature limit is not challenged. If a
conservative assumption postulates the fire to occur inside the vent,
similar results would occur because there is only a limited amount of
energy (BTU) that can be deposited into the massive overpack structure.
Exceeding the concrete temperature limit only at the concrete surface
does not lead to a safety concern, and therefore, the SAR analysis is
acceptable.
Comment C.23: One commenter stated that the consequences of a hit
by an anti-tank missile, such as the MILAN or
[[Page 25249]]
TOW-2 missile should be considered. The commenter noted that the
regulations only require a licensee to install systems that protect
against unauthorized entry; however, entry to a site is not necessary
to successfully carry out sabotage using an anti-tank missile. The
commenter stated that the NRC should place additional conditions in the
CoC to lower the probability of a sabotage event. The commenter further
pointed out that the NRC has been inconsistent and arbitrary in
determining whether to treat sabotage issues as site-specific or
generic.
Response: The NRC disagrees with the comment. The threat of an
anti-tank missile and other sabotage events is beyond the scope of this
rule. Requirements on radiological sabotage are covered in 10 CFR Part
73 and apply to both ISFSIs and spent fuel storage cask designs.
Therefore, comments on a specific threat or mode of attack are beyond
the scope of this Part 72 rule. See also the response to C.8 addressing
radiological sabotage of storage casks at reactor site ISFSIs.
D. Criticality
Comment D.1: One commenter objected to the assumption on the
continued efficacy of the boral over a 20-year storage period because
it has never been tested or proven.
Response: The NRC disagrees with the comment. The NRC staff does
not consider the loss of fixed neutron poisons credible after
installation into the cask because the poisons are fixed in place and
contained. The neutron absorber is designed to remain effective in the
HI-STORM system for a storage period greater than 20 years and there
are no credible means to lose the neutron absorber. Section 6.3.2 of
the HI-STORM SAR describes the neutron absorber and its environment,
and evaluated boron depletion due to neutron absorption. Section
9.1.5.3 of the SAR describes the testing procedures for the neutron
absorber material. The neutron absorber material will be manufactured
and tested under the control and surveillance of a quality assurance
and quality control program that conforms to the requirements of 10 CFR
Part 72, Subpart G. The compositions and densities for the materials in
the computer models were reviewed by the NRC staff and determined to be
acceptable. This material is not unique and is commonly used in other
spent fuel storage and transportation applications.
Comment D.2: One commenter asked if Boral had ever been used in any
dry storage casks before and if it had, how long and had it been
tested. The commenter asked if this was an experiment with a new
application. The commenter further asked what proof was available to
show the continued efficacy of a boral panel. The commenter asked what
other fuel storage and transport applications utilized Boral and stated
that it should be documented in the SER.
Response: As described in SAR section 1.2.1.3.1, Boral has been
used in environments comparable to those in spent fuel storage casks
since the 1950s, and in spent fuel shipping casks for Canadian spent
fuel in the 1960s. In the United States, Boral has been used in
numerous other spent fuel transportation casks since the early 1980's
and in storage casks since the early 1990's. Some of the casks that use
Boral are the NAC-I28 S/T, NAC-S/T, the NUHOMS-24P, NUHOMS MP-187, and
BMI-1. The NRC disagrees that the HI-STORM SER should include a list of
other casks that use Boral. Information on other spent fuel casks and
Boral is publicly available. The response to comment D.1 discusses the
efficacy of Boral and why testing other than initial fabrication
testing is not necessary.
Comment D.3: One commenter stated that a test should be conducted
to verify the presence and uniformity of the neutron absorber in
fabrication.
Response: The presence and uniformity of the neutron absorber is
verified as described in Section 9.1.5.3 of the SAR. The neutron
absorber material will be manufactured and tested under the control and
surveillance of a quality assurance and quality control program that
conforms to the requirements of 10 CFR Part 72, Subpart G.
Comment D.4: One commenter asked if water injection in unloading
reflood could result in large amounts of steam generation and two-phase
flow conditions inside the MPC cavity causing over pressurization of
the confinement boundary and a potential criticality.
Response: As stated in SAR section 6.4.2.1, the HI-STORM system was
evaluated with various water densities inside the cask. The cask met
the design criterion of keff less than or equal to 0.95 for
all credible flooding conditions. The cask is most reactive when filled
with full density water. As can be seen in SAR Table 6.4.1, the cask
reactivity decreases when filled with low density water (i.e., steam).
In addition, Section 4.5.1.1.6 describes the cask cooldown and
reflood analysis during fuel unloading operation. This section of the
SAR states that before reflooding the cask with water, the helium
inside the MPC is cooled to below 200 deg.F which is below the boiling
point of water. The procedures are outlined in Section 8.3.1 of the SAR
with reference to TS 3.1.3. These procedures limit steam generation and
two-phase-flow interactions with the fuel to acceptable levels, thereby
preventing over pressurization of the MPC.
E. Design
Comment E.1: One commenter asked if there are three MPCs that are
NRC-certified for storage and transfer because the SER states that they
are evaluated and approved.
Response: As stated in Condition 1.b. of the CoC and in Section 1.1
of the SER, there are three types of MPCs that can be used in the HI-
STORM 100 Cask System: The MPC-24, the MPC-68, and the MPC-68F. The
MPC-24 holds up to 24 PWR fuel assemblies that must be intact. The MPC-
68 holds up to 68 BWR fuel assemblies that may be intact or damaged
(i.e., with known or suspected cladding defects greater than hairline
cracks or pinholes). The MPC-68F holds up to 68 BWR fuel assemblies
that may be intact, damaged, or in the form of fuel debris (i.e., with
known or suspected defects such as ruptured fuel rods, severed fuel
rods, and loose fuel pellets). All three MPCs have the same external
dimensions. Section 1.2.1.1 and Table 1.2.1 of the SAR has been revised
to clarify that there are three types of MPCs.
Comment E.2: One commenter asked how and to what the trunnion is
attached, and what it is made of.
Response: The trunnions are attached by welds to the inner and
outer shell and to the HI-TRAC top flange. The trunnions are fabricated
of SB-637-N07718 steel and SA-350-LF3 steel.
Comment E.3: One commenter stated that the concrete for the
overpack should be reinforced and asked why it is not reinforced.
Response: The NRC disagrees with this comment. The main function of
the concrete encased between the steel shells in the HI-STORM 100
overpack is shielding. The structural strength of the HI-STORM 100
overpack is provided by the inner and outer carbon steel shells. The
concrete, on the other hand, will provide an added benefit to the HI-
STORM 100 overpack because it will increase the stiffness and weight of
the overpack to resist external forces due to seismic, tornado, and
tornado missiles.
Comment E.4: One commenter asked if the pedestal could shift in
movement and touch the liner or if it could corrode to the carbon steel
liner or baseplate. The commenter also asked what the
[[Page 25250]]
baseplate was made of and if a ceramic baseplate should be used.
Response: The pedestal consists of concrete, 17 inches thick,
encased in a steel shell. This shell is welded to the steel overpack
baseplate, and the weld is examined according to the ASME Code Section
V. Stresses in the pedestal have safety factors exceeding 16. The
pedestal will not shift. The exterior and interior surfaces of the
overpack are coated with an epoxy paint to prevent corrosion. The
overpack baseplate is made of carbon steel that meets the design
criteria.
Comment E.5: One commenter stated that jamming of parts could be a
problem because unjamming the parts could cause damage (during both
loading and unloading). The commenter further asked if the cask had
been tested for jamming and what the situation would be after 20 to 40
years in storage.
Response: Stainless steel shims, depicted in Detail T of drawing
1495, sheet 5, prevent the MPC from contacting the overpack interior
and preclude the paint from being scraped during the operational steps.
The drop accident analyses cause stresses which significantly bound the
stresses that could occur during normal handling operations. Therefore,
damage to the MPC during loading and unloading into the overpack is not
credible.
The calculation in the SAR demonstrated that there will be no
jamming of the MPC in the overpack under the most severe stack-up of
tolerances. The cask has not been tested for jamming; however, a dry
run of all operational steps is required before use of the system.
The license life of all overpack and MPC components is 20 years.
The applicant engineered the overpack, HI-TRAC, and MPC for 40 years of
design life. More detailed information regarding the service life of
the overpack, HI-TRAC, and MPC can be found in Sections 3.4.11 and
3.4.12 of the SAR.
Comment E.6: One commenter stated that the clearances were not
adequate. The commenter asked if the wet fuel would be inserted into
the overpack in the same way as the dry run and what happens if the
crane does not have the MPC completely vertical when inserting it in
the overpack or if the HI-TRAC or pad is not level.
Response: There is no adverse tolerance stack-up that would prevent
the insertion of the MPC into the overpack. Additionally, the dry run
will verify that the MPC can be inserted into the HI-TRAC and overpack.
All cell plates of the MPC are constructed of stainless steel that is
not effected by immersion in water; therefore, the tolerances for the
dry run would not change and the wet fuel will go in the same way as in
the dry run.
Comment E.7: One commenter is concerned that the manufacturer's
tolerances are not clear if fabrication is the minimum margin of safety
or minimum clearance allowed.
Response: The most severe ``stack-up'' of manufacturer's tolerances
provides sufficient clearance for insertion of the MPC into the HI-
TRAC. The minimum clearance allowed is thus met. The cask could be
manufactured to the minimum allowed clearances, but this would not
reduce the minimum margin of safety.
Comment E.8: One commenter asked if there would be a problem if the
radial clearance of the HI-TRAC MPC is at a maximum and the radial
clearance of the MPC overpack is at the minimum allowed. The commenter
asked how much leeway is allowed in fabrication in both of these radial
clearance measurements.
Response: No operational problem exists if the radial clearance of
HI-TRAC/MPC is at a maximum tolerance ``stack-up'' and the radial
clearance of the MPC overpack is at the minimum tolerance ``stack-up.''
These tolerances have been evaluated for all manufacturer's design
criteria requirements and for all design temperatures. The largest
allowable radial dimension of the HI-TRAC is greater than the smallest
allowable radial dimension of the overpack. Fabrication requirements,
including tolerances, are stated on the drawings. These tolerances
provide sufficient clearance for operations.
Comment E.9: One commenter expressed concern over the \13/16\-inch
difference in the maximum MPC diameter and minimum overpack internal
diameter because it was a minuscule amount for fabrication. The
commenter also asked what was meant by average radial clearance of
about 0.4 inches and stated that it was not a lot of clearance.
Response: The \13/16\-inches is the minimum clearance accounting
for tolerances between the MPC diameter and the channels/shims that are
attached to inner shell of overpack. The channels/shims provide
guidance for MPC insertion, position MPC within the overpack, and allow
the cooling air flow to circulate through the overpack. The minimum
clearance between the MPC and overpack inner shell is approximately 5
inches without the channels. Both the clearance between the MPC and
channels/shims and between the MPC and overpack inner shell are
considered to be acceptable. The SER has been changed to clarify that
\13/16\-inches is the clearance between the maximum MPC diameter and
the channels/shims that are attached to inner shell of overpack rather
than between the MPC and overpack inner shell. The average radial
clearance is diametral clearance divided by two.
Comment E.10: One commenter asked what the computed decrease (page
3-9 of the SER) was related to. The commenter expressed concern that
these were very small calculation amounts (0.11 inches) to depend on
computer accuracy.
Response: The computed decrease of 0.11 inches is the calculated
maximum decrease in the inner diameter of the overpack shell due to a
tipover accident. The 0.11 inches decrease in the inner diameter of the
overpack shell is not computed by computer simulation. Rather, it is
computed by using a standard text book equation for deformation
calculation. The deformation due to tipover is expected to be small.
This calculation has been evaluated by the NRC staff and found to be
acceptable.
Comment E.11: One commenter asked if the base under the pads would
be the same at all sites and asked what is under the pad. The commenter
is concerned that the pad evaluation has not received adequate
attention because it is a crucial part of the tipover and drop
evaluation.
Response: Each user is required to meet the site parameters in CoC,
Appendix B, Section 3.4 that include specific requirements for the pad.
Site characteristics will be investigated by each cask user and
addressed in the cask user's 10 CFR 72.212 evaluation. The pad is a
site-specific issue. Site-specific issues are beyond the scope of this
rule that is focused solely on whether to add the HI-STORM 100 cask
system to the list of approved casks.
Comment E.12: One commenter asked why there were two different
weights for the transfer cask.
Response: As discussed in Section 1.2.1.2.2 of the SAR, the 100-ton
transfer cask weighs less than the 125-ton transfer cask because it has
a reduced thickness in lead and water. The 100-ton transfer cask is
designed for facilities not capable of handling the heavier 125-ton
transfer cask.
Comment E.13: One commenter asked why the bottom pool lid supported
the weight of a loaded MPC plus water.
Response: During lifting of the transfer cask from the fuel pool
there is water in the MPC and the annulus. Therefore, the structural
evaluation of the bottom pool lid of the transfer cask must consider
all the applicable weights
[[Page 25251]]
supported by the pool lid, including the water.
Comment E.14: One commenter stated that the cask should be up on
something to air out the area under the cask to prevent rusting. The
commenter questioned if the baseplate rusted if that could cause the
cask to tipover or lean. The commenter is concerned that if the
canister ended up leaning against the inner liner of the concrete
shell, it would cause blockage of the venting annulus and create a
hotspot in the concrete.
Response: The NRC disagrees with this comment. The baseplate is
coated with an epoxy type coating to prevent corrosion. Some rusting
may occur at scratches in the coating. However, even a postulated
extreme case, assuming no coating present, would not result in
sufficient corrosion to cause an amount of leaning that would be
significant.
Comment E.15: One commenter asked if there is any leeway for the
pressure in the concrete encasement between the two carbon steel outer
and inner liners, if the concrete had room to move, and if the concrete
could split the outer carbon steel encasing it, particularly at the
welds.
Response: The coefficient of thermal expansion of steel is only
slightly greater than that of concrete, and the thermal gradient
through the overpack wall, experienced during the extreme temperature
criteria, was calculated to be approximately 40 deg.F. This temperature
difference and thermal coefficient of expansions do not cause the inner
steel to apply significant force to the concrete in the overpack. The
outer steel shell expands somewhat more than the concrete; therefore,
the concrete has room for expansion and exerts no force on the outer
steel plates.
Comment E.16: One commenter asked what a bottom pool lid is and how
it is replaced by the heavier shielded transfer lid and if it has been
tested.
Response: The bottom pool lid is described in Section 1.2.1.2.2 of
the SAR. The lids are interchanged with a transfer slide device as
described in Section 8.1.1 of the SAR. The NRC did not require test
results for lid changing operations. The NRC found the pool and
transfer lid design to be acceptable for the HI-STORM 100 system.
Comment E.17: One commenter asked if the 17.000 inches of concrete
for the overpack baseplate was a typo and if the number of significant
figures was correct.
Response: The value in the SER has been revised to state 17.0
inches that reflects the thickness assumed in the shielding analysis.
Comment E.18: One commenter stated that the configuration
discussion in Section 6.4.1 of the SER is not clear because the HI-STAR
doesn't have a transfer cask.
Response: As stated in SER section 6.4.1, the HI-STORM system has a
transfer cask, not the HI-STAR system. The transfer cask, the HI-STORM
overpack, and the HI-STAR overpack are constructed of different
materials. The effectiveness of these materials to reflect neutrons
affects the criticality safety of the system; therefore, each was
explicitly evaluated. The other parameters affecting the criticality
safety of the HI-STORM system, including the transfer cask, are
identical to the HI-STAR system.
Comment E.19: One commenter asked if the closure ring was a ring or
a lid.
Response: The closure ring is a ring. In the MPC, the lid and the
closure ring are two different components.
Comment E.20: One commenter asked how many rings are included in
the design.
Response: There is one closure ring included in the design.
Comment E.21: One commenter asked why voids in the installation of
the lead shield are only minimized instead of being disallowed
completely, if the shield was composed of lead bricks or poured, and
which was more prone to voids. The commenter asked if lead bricks could
be used and then have lead poured into the cracks between the bricks,
and how the lead shield is installed.
Response: The HI-STORM 100 must be fabricated and tested in
accordance with the drawings specified in the SAR and under a quality
assurance program that meets the requirements of 10 CFR Part 72,
Subpart G. The proper fabrication of the lead shield, including
potential voids, will be evaluated under this quality assurance
program. As discussed in Section 9.1.5.2 of the SAR, effectiveness of
the lead pours are verified during fabrication by performing gamma
scanning on all accessible surfaces of the transfer cask in the lead-
pour regions. Installation of the lead shields is discussed in Section
9.1.5.1 of the SAR. The SAR specifies the use of poured lead and does
not allow the installation of lead bricks.
Comment E.22: One commenter asked what the relief valve was and
what type of maintenance it received.
Response: A relief valve is a mechanical device that opens when
pressure inside a system exceeds the actuation pressure of the valve
(pressure that will open the valve). Relief valves are common pressure
limiting devices. Relief valves are placed on water heaters in homes to
ensure that the water pipes in a house will not fail due to excessive
pressure. Similarly, relief valves are attached to the radiator in a
car to ensure that the coolant hoses do not burst from excessive
pressurization of the engine coolant system. Maintenance of the relief
valves are discussed in SAR Section 9.2.4. The relief valves are
calibrated annually to ensure that their pressure relief setting is
correct or they are replaced with factory-set relief valves.
Comment E.23: One commenter asked if there were holes in the shield
jacket to add and drain things and indicated that holes would be a
potential sabotage threat for someone to drain the jacket or add
something dangerous to the water.
Response: There are drain holes in the water jacket end plate. The
125-ton HI-TRAC has two 1\1/2\-inch drain holes and the 100-ton HI-TRAC
has four \3/4\-inch drain holes. The resultant dose rate for an assumed
loss of the water jacket is addressed in Section 5.1.2 of the SAR. The
analysis indicates that the off-site dose at 100 meters will be below
the 5 rem accident limit in 10 CFR 72.106. In addition, NRC regulations
in 10 CFR Part 72 have established physical protection and security
requirements for an ISFSI located in the owner-controlled area of a
licensed power reactor site.
Comment E.24: One commenter stated that Conditions 1a and 1b of the
CoC should both state that the cask system has two transfer casks.
Response: The NRC disagrees with the comment. Condition 1b of the
certificate of compliance specifies that there are two types of
transfer cask options: the 125-ton HI-TRAC and the 100-ton HI-TRAC. It
is not necessary to repeat that information in Condition 1a.
Comment E.25: One commenter stated that there should be a drawing
of the damaged fuel container in the CoC because the structure is not
explained.
Response: The NRC disagrees with this comment. A drawing of the
damaged fuel container is included in Chapter 1 of the SAR and is
available to the public. This level of detail is not necessary in the
CoC.
Comment E.26: One commenter asked what the screens are made of, how
the screens are attached, if the screens can deteriorate or come loose
over time, and what happens if the screens fall out.
Response: As shown on the drawings in Chapter 1 of the SAR, the
damaged fuel container, including the screen, is constructed of
stainless steel. The damaged fuel container is an additional structural
component that will make the MPC fuel basket even stronger. The screen
is placed between two steel plates welded together with a 0.06 inch,
[[Page 25252]]
continuously 360 degree, all around fillet weld. It is not considered
credible for the screens to fall off or fail. However, if a screen
failed, there would be no release of radioactive material because the
MPC is sealed. Small amounts of loose debris in the MPC have been
considered during unloading operations, as described in SAR Section
8.3.4.
Comment E.27: One commenter stated that damaged fuel and intact
fuel should not be placed in the same cask because it can cause
potential problems in unloading.
Response: The NRC disagrees with the comment. Damaged fuel can be
stored safely with undamaged fuel. If damaged fuel is stored with
undamaged fuel, then CoC, Appendix B, Section 2.1.1.c requires all fuel
assemblies in the cask to meet the more restrictive heat generation
requirements for the damaged fuel. Additionally, damaged fuel must be
loaded into damaged fuel containers to enable safe handling during cask
loading and unloading operations.
Comment E.28: One commenter asked what the basis is for putting the
hotter fuel in the center of the cask. The commenter also asked if the
doses would be accurate if the lower dose fuel is placed at the
periphery positions. The commenter stated that it would be better to
have a more even heat and dose distribution in the MPC and asked if
dose was more important than the heat.
Response: The design of the HI-STORM cask considered both the
thermal and radiological effects of the fuel. If one assumes the same
enrichment and burnup (time that the fuel was left in the reactor to
produce power), the fuel that is left longer to decay in the spent fuel
pool (e.g., ``cooler fuel'') will generate less heat and high-energy
radiation than the fuel that is removed sooner from the pool (e.g.,
``hotter fuel''). For the method used in the HI-STORM design, cooler
fuel assemblies are stored on the periphery of the cask for two
reasons. First, the ``cooler fuel'' assemblies have lower allowable
peak clad temperature limits and the temperature of the assemblies on
the periphery is cooler. Second, storing the ``cooler fuel'' on the
periphery of the MPC provides some additional radiological shielding
from the hotter fuel assemblies in the center.
Comment E.29: One commenter asked if BPRAs and thimble plugs could
be stored in the cask. The commenter stated that they should not be
because they add weight.
Response: BPRAs and thimble plugs have not been analyzed for this
cask system and, therefore, are not authorized for storage in the HI-
STORM 100 at this time.
Comment E.30: One commenter asked what the cask transfer station is
and whether it had been designed yet. The commenter asked if it is
constructed of reinforced concrete. The commenter stated that more
explanation was necessary and that a drawing should be included. The
commenter asked how and why an impact limiter is used. The commenter
asked what the basis is and if an evaluation had been completed. The
commenter was concerned with the use of terms ``if'' and ``shall be
designed'' because this implies the CTF hasn't been designed. The
design should have specific criteria.
Response: The term ``cask transfer station'' in CoC, Appendix B,
Section 3.5.1, is a typographical error and has been corrected to
``cask transfer facility.'' A cask transfer facility (CTF) is a
facility used for transferring the MPC between the transfer cask and
the overpack. The CTF does not include 10 CFR Part 50 controlled
structures such as the fuel handling building or reactor building. The
NRC disagrees that a drawing of the CTF or more design details are
necessary. The HI-STORM 100 Cask System is approved for use under the
general license provisions of 10 CFR Part 72. Therefore, the cask may
be used in any nuclear power reactor site licensed under 10 CFR Part
50, provided that the site parameters are enveloped by the cask design
bases. The specific design and operation of the CTF will be dictated by
site-specific needs. Because of the varied needs of each reactor site,
the NRC found it impractical and unnecessary to review and approve a
specific CTF design, including the specific materials of construction.
The NRC reviewed and approved the criteria for the design,
construction, and operation of the CTF. These criteria are specified in
CoC, Appendix B, Section 3.5, and SAR Section 2.3.3.1.
The impact limiter is a possible CTF design feature whose function
would be a defense-in-depth measure because the lifting equipment used
in the CTF must be designed to preclude a drop. As discussed in the
response to comment J.14, the specific requirement for an impact
limiter has been eliminated because other methods may be available to
prevent a canister breach in case of a canister drop during transfer
operations.
Comment E.31: One commenter stated that we should clearly state
what the CTF is and make sure that every detail of the procedure is
carefully analyzed because it is vague.
Response: The CTF is defined in SAR Section 2.3.3.1 and in CoC,
Appendix B, Section 1.0. Detailed design and operational requirements
for the CTF are also specified in SAR Section 2.3.3.1, as well as in
CoC, Appendix B, Section 3.5. Under the provisions of 10 CFR 72.212 and
CoC Condition 2, each licensee that elects to use a CTF must develop
written procedures for operating the CTF. These procedures are subject
to NRC review during inspection. As required by TS 5.2.h, the licensee
must conduct a dry run training exercise, prior to first use of a CTF,
to demonstrate that the procedures can be conducted safely and
successfully.
Comment E.32: One commenter recommended that Design Drawing 1495,
Sheets 4 and 6 and Design Drawing BM-1575, Sheet 2 be revised.
Response: The NRC agrees with the comment. These changes correct
drafting errors or provide a level of flexibility that is acceptable to
the NRC staff. The SAR drawings have been revised accordingly.
F. Welds
Comment F.1: One commenter asked how use of the trunnions puts
stress on the weld at the water jacket.
Response: Use of the pocket trunnion does not put any stress on the
water jacket. The seal weld between the pocket trunnion and water
jacket shell is for retaining the water inside the water jacket. The
pocket trunnion is attached to the outer transfer cask shell by full
penetration welds all the way around the trunnion. When the pocket
trunnion is used, the force is transferred to this weld and not to the
seal weld on the water jacket. The other type of trunnion on the
transfer cask is the lifting trunnion. The lifting trunnion is not
connected to the water jacket and, therefore, puts no stress on the
water jacket.
Comment F.2: One commenter asked how the welds are checked to be
leakproof and whether water can enter the trunnion.
Response: All the structural welds have to be examined and
inspected according to the applicable ASME code. The welded joint is an
integral part of the structure and is leak proof. Because the trunnions
are made of solid steel, water cannot leak into them.
Comment F.3: One commenter stated that the lid and closure ring of
the MPC should be full penetration welds and should be ultrasonic
tested (UT) as this is the basis for qualification as a redundant seal.
Response: The NRC disagrees with this comment. Full penetration
welds are unnecessary from the structural and
[[Page 25253]]
containment boundary requirements of the design. Employing
unnecessarily heavy welds leads to fabrication problems such as
excessive warpage. UT of heavy section, full or partial penetration,
austenitic stainless steel welds to ASME Code acceptance criteria is
not feasible with current technology. The redundant seal concept is
based upon the use of two welds forming the leak barrier, not the
inspection method. With redundant welds, one weld could leak and the
second still provide leak tight integrity.
Comment F.4: One commenter stated that just because there is no
known plausible, long-term degradation mechanism to cause seal welds to
fail doesn't mean that the welds won't fail.
Response: The NRC disagrees with this comment. The NRC staff has
examined the plausible mechanisms that would cause failure of the seal
welds and has determined that those mechanisms are inoperative under
normal service and design accident conditions for HI-STORM. This gives
the NRC staff reasonable assurance that the welds will not fail under
design basis normal, off-normal, and accident conditions.
Comment F.5: One commenter asked how lid welds are removed in
unloading and stated that the procedures should be in the documents
before the casks are loaded.
Response: The NRC disagrees with this comment. Welded cask lids may
be removed by one of several methods. The method of removal and the
detailed procedures (as opposed to the general procedures of the SAR)
are the responsibility of the ISFSI licensee, subject to NRC review and
inspection. The TSs require ISFSI licensees to perform lid removal
method demonstrations on full-size mock-ups as part of their pre-
operational testing and training exercises. The NRC staff has reviewed
and inspected several methods and their associated procedures.
Inclusion of such detailed procedures in the SAR is unnecessary.
Comment F.6: One commenter asked what happens if the water jacket
welds leak water.
Response: The resultant dose rate for an assumed loss of the water
jacket is addressed in Section 5.1.2 of the SAR. The analysis indicates
that the off-site dose at 100 meters will be below the 5 rem accident
limit in 10 CFR 72.106.
Comment F.7: One commenter stated that the penetrant test (PT) is
unacceptable, that the criteria for layers and time are ``wishy
washy,'' and that PT tests should not be allowed.
Response: The NRC disagrees with this comment. The commenter has
not specified why PT is unacceptable or why it should not be allowed.
PT is a Code accepted examination method. The progressive PT technique
is used and accepted in the nuclear industry when a volumetric
examination by means such as UT is impractical. UT is unsuitable for
heavy section austenitic stainless steel welds.
The basis for the structural lid weld examination methods is
documented in the NRC's Interim Staff Guidance-4, Revision 1, that
allows the use of a multi-layer (i.e., progressive) PT examination in
lieu of a volumetric examination.
Comment F.8: One commenter stated that welds need to be checked
carefully.
Response: The NRC agrees with the comment. Welds are important
which is why they are examined and inspected according to the
applicable ASME code.
Comment F.9: One commenter stated that the leak testing procedure
used to demonstrate MPC closure cannot be performed as described and
that performance of the test is not generally consistent with ANSI
N14.5-1997, ``Leakage Tests on Packages for Shipment.'' Consequently,
containment of the radioactive material to the stated criteria cannot
be demonstrated. The principal reason provided by the commenter is that
the nominal concentration of helium in air is 5 parts per million. This
atmospheric concentration masks any leakage from the MPC using the
specified test conditions. In addition, the commenter noted that there
is no direct reference to definitions, equations, formula, methodology,
or criteria of the standard in the text. The commenter further noted
that when terminology from the standard is given, it is (for the most
part) used incorrectly.
Response: The NRC disagrees with the comment. These welds are
multipass stainless steel welds that are dye penetrant examined
multiple times during the weld process. The multiple dye penetrant
examinations provide reasonable assurance of high integrity welds that
will retain the inert gas and prevent leakage of radioactive material
into the environment. The leakage testing of these welds provides
additional insight into the leak-tightness of these welds.
The NRC staff has reasonable assurance that the leakage test
procedure outlined in SAR Section 8.1 can be performed as described
provided that appropriate equipment is used and the leak test method is
properly qualified. The leakage testing for the lid is to be performed
with a sniffer type probe; however, the test method for the port and
drain covers is not specified in the SAR. There are test methods
discussed in Appendix A of ANSI N-14.5 that could be used to perform
the port and drain cover leakage testing. Detailed procedures are
developed by the user who is responsible for ensuring that the TS
limits are met and therefore, confinement is adequately maintained. The
leakage testing will require a demonstration that the detector can
identify an appropriate calibrated leak in the presence of background
helium. Although sniffer probes detect discrete leaks rather than an
integrated leakage rate, the typical sniffer probe sensitivity of
10-\8\ provides reasonable assurance that the TS leakage rate limit of
5 x 10-\6\ will not be exceeded.
As stated in the SAR, the leak testing will be performed in
accordance with ANSI N14.5. It is not necessary to include any more
level of detail in the SAR; therefore, no change to the SAR is
necessary. Appropriate detail will be included in the site procedures.
The SAR has been changed to use terminology consistent with the TS
and the 1997 revision of ANSI N14.5. The terminology was changed from
std cc/s to atm cc/s. SAR section 7.3.3 justifies the use of the units
atm-cc/sec. Also, the SAR was revised to delete the sensitivity of the
detector. The sensitivity will be addressed in the detailed site
procedures.
G. Structural
Comment G.1: One commenter stated that all of the accident level
events and conditions listed in the SER, particularly a transfer cask
handling accident or sabotage, should be evaluated for structural
analysis in a jamming condition on top of the overpack.
Response: All design basis normal, off-normal, and accident events
have been evaluated in the structural analyses and are discussed in
Section 3 of the SER. This includes an evaluation of the transfer cask
under a 42-inch horizontal drop during transfer operations. A
horizontal drop from a greater height is not considered because the
horizontal lifting height limit for the transfer cask is 42 inches. The
evaluation shows that the HI-TRAC meets all structural requirements and
there is no adverse effect on the confinement, thermal, or
subcriticality performance of the contained MPC.
As discussed in the response to C.15, vertical drop of a transfer
cask is not considered credible because vertical lifting of a loaded
transfer cask must be performed with structures and components designed
to prevent a cask drop. Also, as discussed in the response to E.5,
jamming is not considered to be
[[Page 25254]]
credible because of the design of the HI-STORM system. The threat of
sabotage is beyond the scope of this rule and is discussed in the
response to C.8.
Comment G.2: One commenter stated that bending of the web and
pushing of the flanges possibly accompanied by some local weld failures
sounded feasible and may not result in limited deformation as assumed.
The commenter asked if a full size cask had been tested in a drop or
tipover.
Response: The channels attached to the inner shell of the overpack
are not classified as important-to-safety and serve no structural
purpose. Deformation of the channels, whether limited or complete
collapse, does not affect retrievability of the MPC. On the contrary,
the deformation of these channels due to a tipover accident absorbs
energy which reduces the deceleration loadings to the MPC and provides
a greater opening in the overpack during retrieval. NRC regulations do
not require full size testing of casks. The applicant can choose the
method of analysis. Computer analyses have been performed to determine
the responses of a cask in drop and tipover accidents.
Comment G.3: One commenter questioned how the structural analysis
conducted for the 125-ton HI-TRAC transfer cask could bound the 100-ton
version and indicated that the 100-ton version needs its own analysis.
Response: All the structural analyses and evaluations of the 125-
ton transfer cask were repeated for the 100-ton transfer cask. However,
the analytical results of the 125-ton transfer cask are greater than
that of the 100-ton transfer cask. Therefore, the structural analysis
of the 125-ton transfer cask bounds the 100-ton transfer cask.
Comment G.4: One commenter asked what would be the consequences of
the deformation of the outer shell and lead and water jacket from a
missile, particularly if the transfer cask was on top of the concrete
shell.
Response: The HI-TRAC transfer cask is always held by the handling
system while in a vertical orientation completely outside of the fuel
building. Therefore, considerations of instability due to a tornado
missile strike are not included in the evaluation. However, a
structural evaluation of the damage to the HI-TRAC transfer cask from
an intermediate missile strike and a large missile strike is performed.
The evaluation shows that the outer shell and the water jacket would
not experience any plastic deformation and will not adversely affect
the retrievability of the MPC.
H. Materials
Comment H.1: One commenter questioned why carbon steel was used for
the inner and outer plate instead of stainless steel because of the
concern over corrosion. The commenter also asked if the carbon steel
was coated.
Response: The materials used in the fabrication of the cask are
described in Chapters 1 and 3 of the SAR and discussed in Section 3.3
of the NRC SER. These materials have been found acceptable because they
meet the requirements for their respective applications in the cask
system. They are suitable for the expected loading and storage in wet
and dry environments, including corrosion and galvanic effects. There
is no requirement for designers to select materials from a given class,
e.g. stainless steels.
The carbon steel used in the overpack is protected from corrosion
by an industrial epoxy coating commonly used for the protection of
steel.
Comment H.2: One commenter stated that one alloy should be
specified for cask fabrication instead of allowing a choice because if
later problems develop, there are fewer variables.
Response: The NRC disagrees with the comment. The materials used in
casks are selected on the basis of the needed properties. Allowing a
choice of more than one material or alloy for fabrication is acceptable
provided that each of the options has the appropriate properties. The
materials chosen for use in the Holtec HI-STORM 100 design have a long
history of favorable performance in the nuclear industry.
Comment H.3: One commenter questioned why plain concrete is not
included in NUREG-1536 and why an exemption was being given to allow
plain concrete since reinforced concrete is stronger.
Response: No exemption was given to allow plain concrete to be used
for structural components. The plain concrete in the HI-STORM 100
overpack is for shielding only and is not a structural component of the
overpack. The reinforced concrete included in NUREG-1536 is for
concrete structures (concrete components that provide structural
strength) only. The HI-STORM 100 overpack is a welded steel structure,
not a concrete structure.
Comment H.4: One commenter asked if the NRC has reviewed the
manufacturers direction for the carboline 890 and thermaline 450
coatings. The commenter asked how the coatings are used and applied,
and if they will wash or flake off in pool water, making the water
cloudy.
Response: The NRC staff has reviewed the manufacturer's technical
information for the coatings mentioned. Both coatings are standard
coatings employed in industry for immersion service and are applied
using common industry tools and techniques. No performance problems
would be expected during intended service.
Comment H.5: One commenter asked if the carbon steel caused
reactions that could create loading or unloading problems such as
reaction products clogging venting or draining equipment with crud or
flakes or making the pool water cloudy.
Response: Carbon steel exposed to the cask loading environment
produces very fine particulates that do not clog equipment. Turbidity
that may arise from corrosion of uncoated carbon steel can be
controlled with appropriate water treatment equipment.
Comment H.6: One commenter asked if temperature or coatings on the
channel could affect the fit. The commenter also asked if flaking of
the coating could clog a channel slide or if corrosion in the channels
could cause problems in unloading.
Response: The effects of temperature on the channels have been
calculated and do not affect the fit. Each coat of the epoxy paint
applied to the exposed surfaces of the inner components of the overpack
is, at maximum, 0.008 inches thick. Two coats result in a maximum
diametral reduction in inside diameter of 0.032 inches. This reduction
will not affect the fit. Both the interior and the exterior of the
channels are coated to prevent corrosion.
Comment H.7: One commenter asked if aging was factored into the
analysis of the pad and stated that the specific site should be
evaluated for a full cask array.
Response: Concrete is resistant to environmental conditions,
including air pollution and moisture. Therefore, the NRC staff expects
no significant degradation of the pad during the licensed lifetime of
the ISFSI facility. Each proposed site is subject to a specific
evaluation to ensure that the design parameters satisfy site-specific
conditions. In addition, cask users are responsible for inspecting and
maintaining the pad, and for ensuring that significant degradation is
not occurring over time.
Comment H.8: One commenter asked what the condition of the concrete
is right under the shell and expressed concern that the concrete could
crack where nobody would see damage needing repair.
Response: As discussed in the response to E.3, the main function of
the concrete encased between the steel shells in the HI-STORM 100
overpack is shielding. The structural strength of
[[Page 25255]]
the HI-STORM 100 overpack is provided by the inner and outer carbon
steel shells. Cracking of the concrete would not have a significant
impact on the cask's ability to meet the regulatory dose limits. There
is no credible mechanism for the concrete to undergo any significant
damage. Thus, inspection of the concrete is not necessary.
Comment H.9: One commenter asked if concrete expanded or released
water or gas when it is superheated.
Response: Concrete contains some traces of free water. If the water
is heated, it will evaporate. Concrete will expand upon heating and
contract upon cooling. The amount is governed by the temperature. These
expansions/contractions are reversible and not permanent. There are no
significant effects of expansions/contractions that would occur even if
the temperature went considerably beyond the design temperature
parameters.
Comment H.10: One commenter asked how the bottom face affects the
supporting surfaces (heat, radiation, weight, stress, pressure etc.).
Response: As listed in Table 4.4.9 of the SAR, the temperature of
the bottom lid plate at normal conditions is 183 deg.F. That
temperature will not have an adverse effect on the concrete. Radiation
will have minimal impact on the concrete pad due to the shielding
provided by the pedestal. The weight, stress, and pressure from the
cask bottom have no adverse effect upon the pedestal or slab because
they are specifically designed to support all the loads due to the
casks.
Comment H.11: One commenter asked how the gas and liquid media that
escapes from the damaged fuel container interacts with other materials
in the MPC and if they can cause problems.
Response: The materials of the cask have been selected to be
compatible with any constituent or reaction product of the fuel.
I. Thermal
Comment I.1: One commenter asked if hot spots in the cladding could
cause lead to sag in the transfer cask if the inner canister is in
place and the temperature is close to the boiling point of the water
pack.
Response: Hot spots in the cladding would not result in sagging or
melting of the lead. The bounding calculation performed by Holtec
assumed all the fuel assemblies were at the design basis limit (hottest
assemblies). The bounding rod cladding temperature occurs at the center
of the MPC and does not have a direct impact on the lead. The
assemblies on the periphery of the MPC are significantly cooler because
they are located near the cooler surface of the MPC. Table 11.2.8 in
the SAR provides the results from a calculation that assumes no water
in the water jacket. These results bound the impact of boiling in the
water pack. Based on those results, it can be concluded that the lead
temperature remains well below the melting temperature.
Comment I.2: One commenter asked what happens if the water in the
transfer pack boils and the steam pressure builds up, and stated that
this situation should be evaluated.
Response: As the pressure builds up, the pressure is relieved
through a safety valve. As water is removed through the safety valve,
the temperature of the water remains at the saturation temperature. The
case of water boiling in the HI-TRAC water jacket is bounded by the
event that assumed no water in the water jacket. This event leads to a
temperature in the water jacket that is higher than the saturation
temperature of the water. The impact of loss of water in the water
jacket is summarized in Table 11.2.8 of the SAR.
Comment I.3: One commenter asked how, during normal conditions, the
temperature of the outer shell could be higher than the temperature of
the concrete because the carbon steel would breathe less than the
concrete, causing the heat to be retained in the concrete. The
commenter also asked how the temperature of the concrete could be
measured since it is encased in the carbon steel.
Response: The question raised by the commenter is not clear. The
temperature of the concrete is higher than the temperature of the outer
shell under normal conditions. Reviewing Table 4.4.9 in the SAR, the
temperature at the overpack outer shell is not higher than the concrete
cross sectional average temperature. The temperature distribution
through the overpack under normal conditions is listed in Table 4.4.9
of the SAR (e.g., 149 deg.F for the concrete and 131 deg.F for the
outer shell).
With regard to the question of measuring the temperature of the
concrete, the applicant does not measure the temperature of the
concrete. Bounding calculations are used to assure that the concrete
temperature limits will not be exceeded.
Comment I.4: One commenter asked how the pad reacts to the bottom
plate of a cask from a temperature differential standpoint. The
commenter asked if the pad would crack and sink under each cask and
form a concave area that could then collect moisture. The commenter
further asked if the collected moisture could boil and if the moisture
could cause the bottom plate to rust.
Response: The heat transfer between the bottom plate of the
overpack and the concrete pad is modeled in the thermal computer code
for the HI-STORM cask system. As listed in Table 4.4.9 of the SAR, the
temperature of the bottom lid plate at normal condition is 183 deg.F.
That energy is transmitted to the concrete pad down to the ground,
which is at the normal soil annual average temperature of 77 deg.F.
Therefore, the concrete will not experience temperatures above boiling
(no superheating will occur). In the winter, the concrete will not
reach freezing temperatures below the cask because it generates heat.
If the concrete reaches or exceeds boiling or freezing temperatures,
there is no detrimental effect on the strength or condition of the
concrete. The pad is specifically designed to support the weight of the
casks without any cracking or sinking of the pad. The bottom plate of
the cask is stainless steel and will not rust.
Comment I.5: One commenter questioned the basis and validity of
simulating the heat effect of adjacent casks radiating heat back to an
interior cask and if an analysis of the real situation had been
conducted.
Response: The impact of radiation heat transfer from neighboring
casks was calculated in the HI-STORM thermal evaluations. The method
used by the applicant was to assume that all of the radiated heat is
reflected back to the cask. This modeling assumption is equivalent to
assuming that the cask was totally encircled by other casks. In
reality, less heat will be radiated back to the cask; therefore, the
calculations bounded the effects of neighboring casks. The impact of
neighboring casks was shown to be minimal. The NRC staff does not
require validation of the analytic method with actual experimental
data.
Comment I.6: One commenter asked why the analysis assumed that the
soil below the overpack was at a constant temperature because the casks
could cause hot spots.
Response: The analyses did model the hot spots below the overpack.
The computer simulation of the overpack modeled the concrete pad that
the overpack is placed on and the temperature of the soil below the
concrete pad. The soil is one of several paths for heat to leave the
cask. The most significant path for heat dissipation is through the air
passage between the MPC and the overpack. The applicant used the
highest annual average soil temperature found in the USA. The purpose
for using the highest average temperature for the soil and air
[[Page 25256]]
in the thermal analyses is to demonstrate fuel retrievability and that
the cladding is protected during storage against degradation that leads
to gross ruptures (10 CFR Part 72.122). One acceptable method for
demonstrating that the cladding will not undergo gross rupture is to
place a limit on the allowable cladding temperature such that
reasonable assurance exists that the cladding will not significantly
degrade. A report by the Pacific Northwest National Laboratory, PNL-
6189, dated May 1987, provides one acceptable approach for establishing
a temperature limit. The PNL method is conservative when compared to
the maximum allowable degradation permitted in Part 72 of the
regulations. This method, in conjunction with the maximum annual
average temperature, solar heating (e.g., insulation), analytic
assumptions, etc., provide reasonable assurance that the requirements
of Part 72 will be met.
Comment I.7: One commenter asked why an exception was allowed for
exceeding the short term temperature limit for the fire accident
scenario and stated that an exception should not be allowed.
Response: The American Concrete Institute (ACI) establishes
temperature criteria for concrete. One, but not the only, acceptable
demonstration that the concrete overpack will maintain its intended
function is to meet the temperature criteria in ACI 349. However, as
stated in the NRC staff's Standard Review Plan (NUREG-1536), ``a small
amount of exterior concrete spalling may result from a fire, the
application of fire suppression water, rain on heated surfaces or other
high-temperature condition. The damage from these events is readily
detectable, and appropriate recovery or corrective measures may be
presumed. Therefore, the loss of such a small amount of shielding
material is not expected to cause a storage system to exceed the
regulatory requirements in 10 CFR 72.106 and, therefore, need not be
estimated or evaluated in the SAR. The NRC accepts that concrete
temperatures may exceed the temperature criteria of ACI 349 for
accidents if the temperatures result from a fire.'' The Holtec analysis
demonstrated that the amount of concrete that exceeds the ACI
temperature limit is very limited and would not pose a significant
safety hazard.
Comment I.8: One commenter asked for the basis of using an average
temperature of the gas in the gap and plenum of the limiting rod and
questioned the validity of the assumption.
Response: The purpose for evaluating the average of the gas
temperature in the fuel rod is to calculate the pressure within the
fuel rod. The computer code used in the analysis calculates the
temperature profile of the fuel rod, but does not calculate the
corresponding pressure for that rod. To calculate the pressure, the
average temperature of the gas is calculated and from the ideal gas
law, the corresponding pressure is established.
Comment I.9: One commenter asked what is in the water used for
forced water circulation under wet transfer of the fuel from the spent
fuel pool to the location for vacuum drying and if the water could
chemically affect other materials in the cavity. The commenter asked
how fast the water flows, if steam could be formed, and if the water
could physically affect other materials in the cavity, movement of
rods, flaking of paint, etc.
Response: The licensee can either use demineralized water or water
from the spent fuel pool. Neither demineralized nor spent fuel pool
water would adversely interact with the system. The flow rate of the
water is based on the heat output of the fuel assemblies and is a site-
specific issue.
Comment I.10: One commenter asked what the water chiller is used
for and what material is used as the chilling medium.
Response: The water chiller is used as the heat sink for cooling
the helium inside the MPC to below 200 deg.F. The type of water chiller
used is a site-specific issue and not part of this rulemaking activity.
Comment I.11: One commenter asked for specific criteria that
defines clearance around the cask for cooling purposes instead of
stating a reasonable amount. The commenter also asked how close other
heat sources may be located and what is considered to be a significant
heat source.
Response: The actions identified by the commenter are only valid
when a breakdown occurs in the helium coolers (LCO 3.1.3). Section
B3.1.3 of the technical specification bases states that ``if the
TRANSFR CASK is located in a relatively open area such as a typical
refuel floor, no additional actions are necessary.'' However, a
licensee may elect to perform the cooling with the cask located in a
pit or vault. This is a site-specific activity. The bases identify
three acceptable options for ensuring adequate heat transfer for the
TRANSFER CASK. The user may develop other alternatives on a site-
specific basis, considering actual fuel loading and decay heat
generation within the cask. One of the options is to fill the annulus
between the MPC and the TRANSFER CASK with water. The second option is
to remove the TRANSFER CASK from the pit or vault and place it in an
open area such as the refueling floor with a reasonable amount of
clearance around the cask and not near a significant source of heat.
The third option is to supply nominally 1000 SCFM of ambient air to the
space inside the confined space (e.g., pit or vault). With respect to
defining an acceptable distance, the licensee could use the analyzed
event of 15 feet center-to-center storage spacing that corresponds to a
four foot clearance. Smaller clearances would also be acceptable, given
the heat load rating of the cask and ambient conditions. With regard to
defining a significant heat source, this is a site specific
consideration. For example, if the plant is using a bank of radiant
heaters near the cask, then an evaluation needs to be performed to
ensure that those heaters pose no adverse impact on the cask. These
options are only guidelines to an LCO that a user would have to
consider.
Comment I.12: One commenter asked how cool air was provided to the
space inside the vault at the bottom of the overpack. The commenter
stated that this needs to be planned out ahead of time for ALARA
considerations and equipment availability.
Response: This is a site-specific issue and not part of this
rulemaking activity. The NRC staff agrees with the comment that the
user needs to plan this activity considering ALARA and equipment
availability.
Comment I.13: One commenter stated that fuel should be adequately
cooled before it goes into the transfer cask.
Response: The NRC agrees with the comment that adequate planning is
needed when performing cask cooldown and reflooding. The purpose of the
analyses performed in the SAR is to maintain the integrity of the fuel.
The requirements on the burnup and minimum cooling time serve that
purpose.
Comment I.14: One commenter asked if the temperature of the helium
accurately reflects the internal temperature of the MPC and stated that
this should be tested.
Response: The exit temperature of the helium reflects the
conditions of the fuel rods. After the helium temperature is reduced
below 200 deg.F, the bulk of the fuel will be at low temperatures,
minimizing the potential for excessive steaming. Reflooding of a
canister has been demonstrated without pre-cooling the helium. No
additional tests are needed.
[[Page 25257]]
Comment I.15: One commenter objected to the addition of ethylene
glycol solution to the demineralized water in the water jacket to
prevent freezing and asked where this had been tested. The commenter
also asked why the antifreeze was used, how the solution would mix, how
the NRC knows it will work, what types of effects it could have on the
inside of the water jacket and the channel walls, if it would add
weight, and how it is added to the water if the jacket is welded shut.
Response: Ethylene Glycol is the chemical name for ordinary
antifreeze. Adding antifreeze to the water jacket, located on the
outside of the HI-TRAC transfer cask, is an option if the user elects
to move a loaded MPC in cold weather, down to 0 deg.F. The use of
antifreeze in the water jacket does not add appreciable weight to the
HI-TRAC cask. Although the water jacket is a welded system, openings
are designed to add and remove water from the water jacket. Antifreeze
has been used in many applications to keep water from freezing. The NRC
staff believes that the industry has ample experience with antifreeze
that additional testing and validity is not necessary. Mixing of the
antifreeze is a site-specific issue that will ensure that the proper
amount of antifreeze is added to prevent the water from freezing at
temperatures down to 0 deg.F.
Comment I.16: One commenter recommended the addition of a note to
Tables 4.4.20 and 4.4.21 of the SAR to provide clarification for the
heat loads.
Response: The NRC disagrees with the comment. SAR Tables 4.4.20 and
4.4.21 refer to loading the MPC with uniformly aged fuel assemblies
emitting heat at the design basis maximum rate. Section 4.4.2
identifies these assemblies as the limiting design basis fuel
assemblies.
Comment I.17: One commenter stated that Holtec's use of a two-by-
four block array to be equivalent to an infinite array assumed a
center-to-center distance between casks of 18.6 feet. The commenter
stated that this equivalency determination between an infinite array
and a two-by-four array is invalid where the differences in cask
spacing do not meet the 18.6-feet center-to-center assumption
underlying the analysis. The commenter noted that the PSF facility
design uses a 15-foot center-to-center distance. The commenter stated
that any CoC issued for this cask system must address this shortcoming.
Response: The NRC disagrees with the comment. First, the PFS
facility is outside the scope of this rulemaking activity. Second, as
identified in Table 1.4.1 of the SAR, the analysis of a 2 by N array
was performed using a pitch of 13.5 feet, not 18.6 feet. The 18.6-feet
pitch is used for a square array. When calculating an equivalent
hydraulic diameter for the square array and taking into account that
the center-to-center spacing of neighboring casks between the pads is
38 feet, as described in Figure 1.4.1 of the SAR, the hydraulic
diameters for the two cases (square array versus 2 by N array) is the
same.
Comment I.18: One commenter stated that thermal interaction of
casks through radiative heat transfer should be considered. The
commenter also stated that the assumption that individual casks will
not interfere with cooling air supply of each other may not be correct.
Response: The NRC agrees that neighboring casks can have an
influence on each other. However, these influences have a second order
impact on the results. The analyses performed by Holtec did credit
radiation heat transfer between the neighboring casks. A bounding
calculation was performed with an ambient temperature of 125 deg.F.
That calculation accounted for heat reflected by the hot concrete pad
and heat generated by neighboring casks. Although not required by the
NRC staff's review of the SAR submittal, Holtec, in response to other
inquiries, performed a sensitivity study to quantify the impact of
neighboring casks and the impact of the sun heating the concrete pad.
The impact of increasing the spacing between casks by a factor of
five in the radial direction resulted in a decrease in the peak cask
surface temperature of 16 oF for an ambient temperature of 100 deg.F
and 17 deg.F for an ambient temperature of 125 deg.F. The impact on
peak clad temperature resulted in a decrease of 6 deg.F for an ambient
temperature of 100 deg.F and a decrease of 8 deg.F for an ambient
temperature of 125 deg.F. Because the peak clad temperature is on the
order of 760+ deg.F, the impact of neighboring casks is minimal.
Comment I.19: One commenter stated that the temperature of the
reflecting boundary should be taken as the temperature of the cask in
interaction with the other casks and not the temperature of an isolated
cask.
Response: The NRC agrees that one method for calculating the impact
of neighboring casks is to model the neighboring casks in the array.
Another acceptable method, that was used by the applicant, is to model
the limiting (highest temperature) cask and assume that all the
radiation it emits is reflected back. This analysis bounds the amount
of radiation that neighboring casks can impose on the center cask. This
bounding analysis is acceptable. As noted in the response to comment
I.18, above, the impact of neighboring casks is minimal, given the
significant margins between the allowable temperatures and the bounding
calculated temperatures.
Comment I.20: One commenter stated that the Holtec model does not
appear to take into account that the heating of the concrete pad is
likely to diminish the ``chimney effect'' of the intake and outlet
vents. The commenter stated that if Holtec had taken this effect into
account, the calculated temperature would be higher in Revision 9 of
the SAR.
Response: In a response to other inquires, Holtec performed
calculations to quantify the effect of concrete pad heating on the cask
performance. For the bounding 125 deg.F ambient temperature event,
neglecting the heat reflected by the pad resulted in a reduction of
cask surface temperature of 10 deg.F and a reduction in peak clad
temperature of 6 deg.F. These temperature differences illustrate that
the concrete pad has negligible impact on the cask.
Comment I.21: One commenter stated that ambient temperature should
be defined due to the importance of the term. The commenter noted that
ambient temperature is an important assumption in the thermal
calculations and an important design element in the CoC. The commenter
stated that the gross oversimplification of the concept of ambient
temperature renders the Holtec thermal analysis completely useless. The
commenter noted that Holtec assumes that the ambient temperature at the
intake and outlet vents is the same; however, the temperature at ground
level will be significantly higher than it will be some distance above
due to the ground absorbing solar energy. The commenter stated that a
desert may have a surface temperature of 180 deg.F, much higher than
the 80 deg.F assumed by Holtec as an intake temperature. This would
reduce the effective buoyancy and air velocity through the cooling
ducts and result in a higher fuel cladding temperature.
Response: The NRC disagrees with the comment. The thermal response
of a cask is very slow. This is due to the large mass of the system. An
analogy can be reached by observing the buildings constructed in the
desert. Massive concrete is used to maintain the indoor temperatures at
reasonable conditions where air conditioners do not exist. The
temperature in those regions fluctuates over each day. For these
structures, an estimate of the average conditions can be assessed by
assuming a bounding average daily temperature. Holtec used such a
method. In addition, the method
[[Page 25258]]
assumed the maximum solar heating specified in 10 CFR Part 71 averaged
over a 24-hour period. Holtec used bounding assumptions approved in the
NRC staff's SER.
Comment I.22: One commenter stated that NRC should have reviewed
the inputs and outputs of the FLUENT calculation. The commenter also
stated that the NRC should have conducted an independent analysis and
validation of the thermal model employed by Holtec. The commenter
stated that the HI-STAR analysis cannot be extrapolated to the HI-STORM
cask because the casks are constructed of different materials, with
different methods of heat dispersion. The commenter stated that the NRC
performed a superficial review and had abdicated its role as
independent regulator and should not issue a CoC for the HI-STORM 100
cask system because there is no lawful basis.
Response: The NRC disagrees with the comment. The NRC staff's
review of the HI-STORM system was not superficial. This is clearly
demonstrated by the NRC staff's requests for additional information,
the applicant's many revisions to the SAR to address NRC staff
concerns, and commitments made by the applicant as outlined in Appendix
12B of the SAR. The NRC staff does not perform independent confirmatory
calculations for every analysis submitted in an application, nor does
the NRC staff routinely review the inputs and outputs of the computer
calculations without cause. Independent analyses that duplicate the
extensive computer calculations performed by an applicant may, at
times, be performed for various reasons. Some reasons include, but are
not limited to, concern that a major error exists in the calculations;
allegations that calculations were improperly performed; use of new
modeling techniques not previously reviewed by the NRC staff; crediting
heat transfer mechanisms not previously reviewed by the NRC staff;
concern that the margin in a complex analysis is small; and concern
that little conservatism exists in the modeling approach.
For the HI-STORM application, the NRC staff reviewed the basic
assumptions used in the calculations, as identified in the SAR and in
the NRC's requests for additional information. A detailed review of
every number is not warranted. As for performing independent analysis
and validation, the NRC staff was able to reach its safety findings
without the need for such calculations. The need for these calculations
is case specific, as addressed above. For HI-STORM, the applicant used
computer codes that are employed by the NRC and have been found
acceptable. The applicant demonstrated its knowledge of the code by
benchmarking its methodology with a full-scale spent fuel cask
instrumented with thermocouples, validating its thermal model and
providing reasonable assurance that its analysts have good working
knowledge of the code to perform the required calculations. The NRC
staff's review of the applicant's methods and assumptions indicate
ample margin and conservatism in the analyses.
The HI-STORM application review process was conducted under NRC
policy and guidance, and as required by the regulations in 10 CFR Part
72. Regarding the reference to the HI-STAR analysis in Section 4.5.4 of
the preliminary SER, the NRC staff intended to indicate that it was
aware that Holtec's use of the FLUENT code had been previously found
acceptable for the HI-STAR application. This reference was not intended
to imply that the NRC staff relied on the HI-STAR calculations or the
prior evaluation in its evaluation of the HI-STORM cask. Section 4.5.4
of the SER has been modified to clarify the description of the NRC
staff's review. Also, to better illustrate the NRC staff's review of
the applicant's submittal, Section 4 of the SER was supplemented with
additional information.
Comment I.23: One comment indicated that the SER states that the
ambient temperature under normal conditions must be less than 80 deg.F.
In addition, the commenter believed Holtec assumed that the ambient
temperature at the inlet and outlet vents is the same and did not
consider warming of the air by heat generated by neighboring casks and
the concrete pad. The commenter stated that calculations indicated that
a desert may have a surface temperature of 180 deg.F and that the
temperature 0.5 m above the ground would be 130 deg.F.
Response: The NRC disagrees with the comment. The SER does not
state that the ambient temperature under normal conditions must be less
than 80 deg.F. The applicant evaluated the cask conditions with an
annual average ambient temperature of 80 deg.F. The use of an annual
average ambient temperature is used in conjunction with the method
described in a Pacific Northwest National Laboratory report PNL-6189.
The method provides one acceptable means for obtaining reasonable
assurance that the requirements in 10 CFR Part 72 will be met. These
requirements include protecting the cladding from degradation that
leads to gross ruptures and designing the storage system to allow ready
retrieval of the spent fuel or high-level radioactive waste. With
respect to the 180 deg.F surface temperature in the desert, the SAR
assumptions used in the 125 deg.F ambient temperature calculation
credits solar heating (also referred to as solar insolation) and heat
generated by the casks. Holtec calculated a concrete pad surface
temperature of 206 deg.F (surrounding the concrete overpack), an
ambient temperature just above the inlet vent of the overpack of
136 deg.F, and a concrete temperature at the outlet vent of the
overpack of 182 deg.F. The NRC staff finds that the Holtec calculation
adequately models the thermal responses of the cask and its
environment.
J. Technical Specifications
Comment J.1: One commenter asked for clarification on the
conditions for use and the TSs, and if they could be changed without an
amendment.
Response: The conditions for cask use are specified in the CoC, and
includes Appendix A (TSs) and Appendix B (Approved Contents and Design
Features). These conditions cannot be changed without an amendment to
the certificate.
Comment J.2: One commenter stated that the Use and Application
section of the TSs is confusing and allows too much flexibility for
completion times and frequencies, and that the TSs should be simple to
understand and done on time.
Response: The NRC disagrees with the comment. The Section 1.0,
``Use and Application'' of the HI-STORM 100 TSs are modeled on the
Improved Standard Technical Specifications (ISTS) for power reactors.
The ISTS were developed as the result of extensive technical meetings
and discussions between the NRC staff and the nuclear power industry in
the early 1990s in an effort to improve clarity and consistency of the
power reactor TSs and to make them easier for operators to use. The
most likely users of the HI-STORM 100 Cask System TSs are power reactor
licensees familiar with the format of the ISTS. The NRC staff believes
that the format of the HI-STORM 100 TSs will make them easier for
operators to use and will help to achieve consistency between power
reactor and spent fuel dry cask storage TSs. The NRC staff disagrees
that there is too much flexibility for completion times and frequency.
The NRC staff believes that the specific wording of the TSs clearly
specifies the allowable time to complete a required action and the
frequency of any surveillance requirements.
[[Page 25259]]
Comment J.3: One commenter objected to the use of the term
``TRANSPORT'' in TS 3.2.2 and indicated that movement to the pad should
be used because this CoC is for storage only.
Response: The NRC disagrees with the comment. The term TRANSPORT
OPERATIONS is specifically defined in Section 1.1 of the Technical
Specifications and includes all activities involved in moving a loaded
overpack or transfer cask to and from the ISFSI pad. Further
clarification of the term is not warranted.
Comment J.4: One commenter stated that ``Each'' should be in large
letters in LCO 3.2.2. The commenter also asked why all the removable
contamination is not removed instead of setting a limit.
Response: The NRC disagrees with the comment. The capitalization of
``each'' is consistent with the format of the TSs. As discussed in the
TS Bases, Section B.3.2.2, the contamination limits for the transfer
cask are established from guidance in NRC IE Circular 87-01. The limits
are based on minimum level of activity that can be routinely detected
under a surface contamination control program using direct survey
methods. These limits are consistent with levels that prevent the
spread of contamination to clean areas and are significantly less than
the levels associated with significant occupational exposure.
Comment J.5: One commenter stated that the dry run should be
conducted in sequence and not an alternate step sequence as permitted
by TS 5.2.
Response: The NRC disagrees with the comment. The dry runs are
performed in discrete functional areas to demonstrate the ability to
perform certain activities as anticipated. The order of performance of
the functional areas, for the purpose of a dry run, is not directly
pertinent to a demonstration of a user's capability. The operating
procedures and technical specifications already control, as necessary,
functional areas that must be performed sequentially for safe storage.
The NRC staff considers it important to allow the cask user the
necessary flexibility to allocate the appropriate resources and
oversight to the performance of dry runs thatmay involve performing and
concentrating on certain activities that would be out of sequence with
a cask loading.
Comment J.6: One commenter asked why no lifting height limit was
established for the vertical orientation of the transfer cask in TS 5.5
and stated that there should be a limit established.
Response: In the SAR, the design basis drop event analysis is based
on the horizontal lifting height of 42 inches. Therefore, TS 5.5 only
specifies the lifting height of the horizontal lifting limit. TS 5.5.c
permits vertical lifting of loaded transfer cask to any height
necessary to perform cask handling operations, including the MPC
transfer. However, the lifts must be made with structures and
components designed to prevent a drop and in accordance with the
criteria specified in CoC, Appendix B, Section 3.5 and SAR Section
2.3.3.1. Therefore, a vertical lift height limit was not established.
Comment J.7: One commenter asked if the diamond-shaped water rod
mentioned in note 10 of TS Table
2.1-3 had been completely analyzed.
Response: The shape (geometry) of water rods that are part of the
fuel assembly, is considered in the evaluation.
Comment J.8: One commenter recommended deleting the words ``For
OVERPACKS with installed temperature monitoring equipment'' at the
beginning of the second option under SR 3.1.2.1 because users should
have the option of using temporary equipment.
Response: The NRC agrees with the comment. Temperature monitoring,
a surveillance option permitted in SR 3.1.2.1, could be conducted with
either temporary or permanently installed equipment. The term
``installed'' could be interpreted as a requirement that the
temperature monitoring equipment be permanently fixed. Therefore, the
beginning of the second option under SR 3.1.2.1 has been reworded as
follows: ``For OVERPACKS with temperature monitoring equipment'' (i.e.,
the word ``installed'' has been deleted).
Comment J.9: One commenter recommended several miscellaneous
editorial changes to the appendices to the CoC.
Response: The NRC agrees with the comment. The appendices to the
CoC have been revised to correct typographical errors and incorporate
minor editorial changes.
Comment J.10: One commenter recommended that Items 5.2.f and 5.2.j
in Section 5 of the TSs be revised to insert the phrase ``(for which a
mock-up may be used)'' at the end of the items for consistency with SAR
Section 12.2.2.
Response: The NRC agrees with the comment. Items 5.2.f and 5.2.j of
the TSs have been revised to indicate that a mock-up may be used for
those specific dry-run evolutions.
Comment J.11: One commenter recommended that item 5.5.c in Section
5 of the TSs be revised to replace the words ``and MPC'' with ``or
OVERPACK'' because some utilities plan to implement an MPC transfer
scheme that requires temporary lifting of the loaded OVERPACK above its
lift height limit.
Response: The NRC disagrees with the comment. There are no
evaluations, equipment design criteria, or other information in the SAR
that support lifting a loaded overpack above its lift height limit.
Comment J.12: One commenter recommended revising the definitions of
DAMAGED FUEL ASSEMBLY and PLANAR-AVERAGE INITIAL ENRICHMENT in Section
1 of Appendix B to the CoC to reflect the evolution of these terms and
for consistency with those in the HI-STAR 100 CoC.
Response: The NRC agrees with this comment. The CoC, Appendix B,
Section 1 has been revised to reflect the new definitions.
Comment J.13: One commenter recommended revising CoC, Appendix B,
Section 3.4.6.c to replace the specified yield strength with the
equivalent ASTM Grade specification.
Response: The NRC agrees with the comment in part. Storage pad
design is a site-specific issue that needs to be addressed in the cask
user's 10 CFR 72.212 evaluation. CoC, Appendix B, Section 3.4.6.c lists
the design parameters for the storage pads. It is not a list of
components for fabrication. By using the specific ASTM Grade
specification as recommended by the commenter, namely, ASTM A615, Grade
60, the designer of the pad will not have the flexibility to choose
other reinforcing steels that could also be used (e.g. ASTM A616 or
A617, Grade 60, etc.). To allow flexibility for the design and still
ensure adequate reinforcement in the pad CoC, Appendix B, Section
3.4.6.c has been changed to state that reinforcement shall be 60 ksi
yield strength ASTM material.
Comment J.14: One commenter recommended eliminating the requirement
for impact limiters at the cask transfer facility contained in CoC,
Appendix B, Item 3.5.2.2.
Response: The NRC staff assumes that the commenter's reference to
Section 3.5.2.2 is a typographical error because the requirement for an
impact limiter is in CoC, Appendix B, Item 3.5.2.1. The NRC agrees in
part with the comment. The specific requirement for an impact limiter
has been eliminated from CoC, Appendix B, Section 3.5.2.1.4. The NRC
determined that this requirement is too restrictive because other
methods may be available to prevent a canister breach in the event of a
canister drop during transfer operations. Instead, Item 3.5.2.1.4 has
been revised to require that
[[Page 25260]]
the CTF be designed, constructed, and evaluated to ensure that if the
MPC is dropped during an inter-cask transfer operation, its confinement
boundary would not be breached.
However, the NRC disagrees with the underlying reason for the
comment which is: Because a single failure proof crane (or equivalent)
is required in the CTF, the design features to mitigate the consequence
of a drop should not be necessary. The NRC staff acknowledges that the
use of a single-failure proof crane precludes the possibility of a
heavy load drop event. The requirement for a mitigating feature in the
CTF design is a defense-in-depth measure that is consistent with the
overall philosophy and approach of NUREG-0612. This philosophy
encompasses an intent to prevent as well as to mitigate the
consequences of postulated accidental load drops. The NRC staff notes
that, even with a single-failure proof crane, NUREG-0612 still imposes
a requirement for a safe load travel path ``to minimize the potential
for heavy loads, if dropped, to impact irradiated fuel in the reactor
vessel and in the spent fuel pool, or to impact safe shutdown
equipment.'' The NRC staff views the mitigating feature in the CTF as a
defense-in-depth measure equivalent to the safe load path. Its function
is to protect the MPC confinement boundary and the integrity of the
spent fuel in the MPC in case of a postulated drop.
Comment J.15: One commenter recommended that CoC, Appendix B, Item
3.5.2.1.4 be clarified to indicate that the acceptance criterion for
the impact limiter also applies to the use of mobile cranes.
Response: The NRC agrees with the comment. As discussed in the
response to J.14, CoC, Appendix B, Item 3.5.2.1.4 has been revised to
require that the CTF be designed and evaluated to ensure that if the
MPC is dropped during an inter-cask transfer operation, its confinement
boundary would not be breached. Section 3.5.2.1.4 has also been revised
to specify that this requirement and acceptance criterion apply to both
stationary and mobile cranes.
Comment J.16: One commenter recommended that CoC, Appendix B, Item
3.5.2.1.4 be revised to clarify the scope of drops that require
evaluation in designing the impact limiter.
Response: The NRC agrees with the comment. CoC, Appendix B, Item
3.5.2.1.4 has been revised to clarify that the potential drops that
require evaluation are those that may occur during inter-cask transfer
operations.
Comment J.17: One commenter recommended that the TSs be removed
from Appendix 12.A of the SAR because they are included in the
appendices to the CoC.
Response: The NRC agrees with the comment. The TSs have been
removed from the SAR.
K. Miscellaneous
Comment K.1: One commenter expressed approval that movement could
be conducted at 0 deg.F and above.
Response: No response is necessary.
Comment K.2: One commenter stated that the HI-TRAC transfer cask
must be as safe as the HI-STORM overpack if it is to be used outside
the reactor security fence.
Response: The NRC staff reviewed the HI-TRAC transfer cask and
determined that, like the HI-STORM overpack, it will perform its
intended safety functions under the design basis normal, off-normal,
and accident events. It should be noted that the CoC authorizes use of
the HI-TRAC only within the owner-controlled areas of a licensed power
reactor.
Comment K.3: One commenter asked if the inner canister could be
dropped through, if water could spill out of the overpack, and if the
water helped to disperse the fuel particles.
Response: During a canister transfer operation, the transfer cask
is placed on top of the storage overpack. The canister is then lowered
through the bottom of the transfer cask into the overpack. It is
unlikely that a canister drop would occur during this operation because
the canister must be lifted with equipment (i.e., a single failure
proof crane or equivalent) that are designed to prevent a drop. In
addition, the overpack contains only traces of water that is part of
the concrete material and the canister is dry during cask transfer
operations.
Comment K.4: One commenter questioned the assumption that the HI-
TRAC remains static because there are a number of man-made or natural
causes that could put it in motion, drop, tipover, roll, etc.
Response: The HI-TRAC is required to be independently secured on
top of the overpack during the transfer of the MPC.
Comment K.5: One commenter asked when the measuring equipment (for
checking tolerances) is calibrated.
Response: The timing of calibration at the fabricator's facility is
beyond the scope of this rule. However, the implemented QA program at
the fabricator's facility provides reasonable assurance that the
measuring equipment for checking tolerances of fabrication will be
appropriately calibrated.
Comment K.6: One commenter asked if the restraint of 11 inches in
vertical height for overpack handling would actually preclude a corner
drop situation. The commenter asked how a corner drop could be
initiated, such as a defective trunnion or lifting lug, etc.
Response: The 11-inch restriction on lifting height for the
overpack was calculated to ensure that deceleration loading to the
loaded MPC would not exceed the design criteria for the confinement
boundary of the basket. A tipover of the overpack cannot occur if the
baseplate is limited to 11 inches above a receiving surface.
Comment K.7: One commenter asked what happens to the inside of the
cask during a horizontal drop of 50 inches.
Response: The effect of a 50-inch horizontal drop of a cask was not
evaluated because the horizontal lifting height limit for the transfer
cask is 42 inches. The 50-inch carry height specified in the SER was a
typographical error and has been corrected to 42 inches. There is no
effect on the confinement function of the MPC as a result of a
horizontal drop of 42 inches. The structural evaluation shows that all
stresses are within allowable values and that the confinement boundary
integrity of the MPC is not impaired.
Comment K.8: One commenter requested that the SER define what is
meant by cladding oxide thickness on page 4-1 of the SER.
Response: The NRC disagrees with the comment. Cladding oxide
thickness is a measure of corrosion at the clad surface. As water
interacts with the zirconium clad, the zirconium can interact with the
oxygen molecules to create zirconium oxide (ZrO2 ). The
terminology is commonly used in the spent fuel storage arena and a
definition in the SER is not necessary.
Comment K.9: One commenter asked why the internal rod pressure is
assumed to remain the same. The commenter asked how the gas behaves in
a dry cask and if it can leak from pinhole leaks and hairline cracks
over the storage period. The commenter further asked how the lower
pressure in the rods affects the analysis and heat transfer.
Response: The internal rod pressure is derived from the initial gas
inserted during fabrication plus the fission product gases that develop
during power production within the reactor core. In a closed system
(e.g., the fuel pin), the pressure is a function of the gases in the
fuel rod and the average temperature of the gas. As the decay heat
decreases with time, so does the temperature and the pressure.
[[Page 25261]]
Therefore, the rod temperature does not remain the same. This is
similar to inflating a balloon with hot air and placing the balloon in
the refrigerator. As the gases cool, the pressure decreases, as is
implied by the smaller diameter of the balloon. The lower pressure
reduces the stress on the cladding and permits a higher allowable
temperature limit. If the rod experiences a pinhole leak or a hairline
crack, the gases inside the rod will mix with the helium gas in the
cask and reduce the internal pressure within the rod.
Reduction of the internal fuel rod pressure results in added
assurance that the cladding will remain stable because the internal
pressure will have equilibrated with that of the cask. The gases from
the fuel pin mix with the gases in the cask and decreases the thermal
conductivity of the helium, while at the same time increasing the
density of the gas. The analyses for accident conditions incorporate
the impact of reduced conductivity of the helium gases. This impact is
reduced when crediting cooling that results from natural circulation of
the gases inside the cask. The use of a maximum allowable temperature
limit provides assurance that the fuel pins will remain intact
throughout the storage period. For conservatism, the applicant assumed
that 1 percent of the cladding experiences a leak under normal
conditions, a 10-percent leak under off-normal conditions, and a 100-
percent leak under accident conditions.
Comment K.10: One commenter asked what cask design was tested at
INEEL (page 4-3 of the SER).
Response: Several full scale cask designs were tested at the Idaho
National Engineering and Environmental Laboratory. The cask used by
Holtec to validate the FLUENT computer code was the TN-24P. The heat
output of the cask was 23 kW. The NRC staff found the FLUENT computer
code acceptable for calculating the thermal response of a spent fuel
cask.
Comment K.11: One commenter expressed concern over water and debris
going into cracks on the pad and then freezing and thawing causing
concrete upheaval and subsequent cask tipover.
Response: Issues related to cask storage pad will be addressed in
the cask user's evaluation under 10 CFR 72.212 and is beyond the scope
of this rule.
Comment K.12: One commenter asked how moisture and pollution in the
air could affect the casks and pad over time and if the pad would ever
need to be replaced.
Response: The cask can withstand the ambient environmental
conditions over its 20-year license period with no significant
degradation. The adequacy of the pad must be addressed by the cask
users in their 10 CFR 72.212 evaluation and is beyond the scope of this
rule. Cask users are responsible for inspecting and maintaining the
pad. With appropriate maintenance, air pollution or moisture would not
cause significant degradation to the pad.
Comment K.13: One commenter asked if both the helium and fission
gases created the pressure inside the rods and for an explanation of
the fission gases. The commenter also asked why only 30 percent of the
fission product gas was assumed to be released instead of 100 percent
because over time 100 percent would likely leak out.
Response: Fission gases are byproducts of uranium splitting in a
reactor. These include gases such as hydrogen, krypton, and iodine. The
gases are contained inside the fuel rod. Data have shown that a
conservative estimate of 30 percent of the gases generated inside the
fuel pellet can escape to the gap that exists between the fuel pellet
and the cladding. This is a conservatively large number used for
calculating dosage. Experimental data has shown this number to be
significantly less. The rest of the gases are trapped inside the fuel
pellet. Therefore, assuming that 100 percent of the gases are released
from the fuel pellet is not realistic. Helium gas is added to the MPC
to keep the environment inside the cask inert so it does not promote
corrosion and to help cool the fuel by transferring heat from the fuel
rods to the wall of the cask. The impact of helium gas on the pressure
within a fuel rod is not as significant as the temperature of the gas
within the fuel rod.
Comment K.14: One commenter asked what is in the water of the water
jacket and if the water could affect the carbon steel channels or get
into the pool through a weld crack or leak and affect the pool. The
commenter also asked how hot the water and the lead get, and if the
water could cause pressure buildup in the channels.
Response: The water used in the water jacket is demineralized water
as is used in the loading pool, but without boron addition because the
boron is unnecessary for loading/unloading operations. Carbon steel
corrodes very slowly in demineralized water; thus, its effect may be
ignored for the durations experienced in loading operations. If the
cask is to be loaded in cold weather, antifreeze may be added to the
jacket water. Antifreeze contains an inhibitor to prevent corrosion.
There would be no significant effect if the jacket water or water with
antifreeze leaked into the pool. With regard to the water and lead
temperatures and pressure buildup in the water jackets, see the
response to comment I.2.
Comment K.15: One commenter asked if there was a recent study on
cladding degradation from creep cavitation.
Response: Studies on cladding degradation were performed several
years ago. These studies led to the development of analytic methods to
calculate the maximum allowable peak clad temperature limits. A report
developed by the Pacific Northwest National Laboratory (PNL) in May
1987, PNL-6189, ``Recommended Temperature Limits for Dry Storage of
Spent Light Water Reactor Zircaloy-Clad Fuel Rods in Inert Gas''
provides an acceptable method for assessing cladding temperature
limits.
Comment K.16: One commenter stated that the 100-ton transfer cask
should not be included in the certification because it is site-specific
and not made the same as the 125-ton cask.
Response: NRC disagrees with the comment. The 100-ton and 125-ton
transfer cask designs have been evaluated and found to meet the
regulatory requirements of 10 CFR Part 72. The 100-ton transfer cask
design is not considered site-specific and is approved under this rule
for use by any general licensee as part of the HI-STROM 100 system as
described in the SAR. Section 2.0.3 of the SAR provides guidance
regarding site-specific ALARA objectives that should be considered by
each user when using either transfer cask design.
Comment K.17: One commenter asked what does reasonable assurance
mean in Section 5.1.2 of the SER regarding acceptability of the
shielding design criteria.
Response: The finding in Section 5.1.2 is intended to mean that the
NRC staff believes that the dose rate criteria presented in the SAR are
acceptable values and that a cask system operating at these values can
meet the applicable radiological requirements of 10 CFR Parts 20 and
72. The SAR subsequently demonstrates that the dose rates calculated
for the HI-STORM system meet the regulatory requirements.
Comment K.18: One commenter asked if the MOX (mixed oxide) fuel was
covered by the sabotage report. The commenter asked if MOX fuel had
been tested and verified to be safe for this design. The commenter
further questioned how the NRC could include MOX fuel in the SER
evaluation and stated that storage of MOX fuel should
[[Page 25262]]
not be allowed by the certification. The commenter also asked how we
know that storage of MOX fuel will work as expected because it has not
yet been tested in Canada.
Response: The sabotage report is beyond the scope of this rule.
However, the design and physical characteristics of a MOX fuel assembly
are very similar to those of a uranium fuel assembly. The primary
difference is the fuel pellet constituents and its effects on the
radiological source term. Testing of MOX fuel is also beyond the scope
of this rule.
The HI-STORM design was evaluated for storage of the MOX fuel
assemblies listed in the Appendix B to the CoC using computer codes and
models. In lieu of testing, the NRC finds analytic conclusions that are
based on sound engineering methods and practices to be acceptable.
Testing is only required if the analytic methods have not been
validated or assured to be appropriate and/or conservative. The NRC
staff reviewed the applicant's analyses and found them acceptable. The
basis of the safety review and findings are identified in the SER and
the CoC.
Comment K.19: One commenter asked if all the analysis was based on
the 100-ton transfer cask or did HI-STORM 100 refer to something else.
Response: The shielding analysis presented in the SAR evaluated
both the 100-ton and 125-ton transfer cask designs as part of the HI-
STORM 100 cask system.
Comment K.20: One commenter asked how the NRC could base its
evaluation on historical statements when reference documents indicate
Inconel impurity may be higher than 1000 ppm. The commenter further
asked what the historical statements were and how we know if the
statements are valid.
Response: The applicant's analysis of cobalt impurities are
discussed in Section 5.2.1 and 5.2.3 of the SER. The applicant showed
that the cobalt impurity value of 1000 ppm assumed in the shielding
analyses was appropriate based on industry data and analyses of post-
irradiation cooling of older fuel types that may have had higher cobalt
impurities for the HI-STORM 100 cask system. As discussed in Section
5.2.1 of the SAR, historical statements included industry data gathered
by the applicant from utilities and vendors.
Cobalt impurities were not necessarily controlled for older fuel
designs. However, the applicant showed that the post-irradiation
cooling time that is inherent to these older fuel types significantly
reduces the HI-STORM 100 dose rates. Therefore, the effects of higher
impurities are mitigated. Based on historical knowledge of recent
cobalt reduction programs, the decay effects on older fuel, and its own
independent evaluations, NRC has reasonable assurance that the
historical statements referenced in the application are used
appropriately for the HI-STORM 100. Furthermore, each cask user will
operate the HI-STORM 100 under a 10 CFR Part 20 radiological protection
program and will be required to verify dose rates that are specified in
the TSs. This defense-in-depth approach will mitigate potential
hardware activation anomalies and ensure compliance with radiological
requirements.
Comment K.21: One commenter asked if the steel transport overpacks
could be reused, how contaminated the overpacks would be after use, the
number of times an overpack could be reused, and if they would be
checked after each use.
Response: This comment that concerns the HI-STAR steel transport
overpack, is beyond the scope of this rule on the Holtec HI-STORM 100
cask system.
Comment K.22: One commenter was pleased that the NRC had evaluated
uneven flooding.
Response: No response is necessary.
Comment K.23: One commenter asked about the chance of one of the
screens being damaged or loosened in unloading and the debris floating
out with the cooling water into the pool.
Response: The damaged fuel container that is placed in the MPC is
stainless steel and is designed to retain damaged fuel and debris in a
safe configuration under all normal, off-normal, and accident
conditions. The damaged fuel container also provides a means to safely
handle the damaged fuel and debris during loading and unloading. It is
not considered credible that the screens will fall off or fail. However
if a screen failed, there would be no release of radioactive material
during storage since the MPC is sealed. Consideration of loose debris
during unloading is addressed in SAR Section 8.3 which outlines the MPC
unloading operations in a spent fuel pool and specifically considers
loose debris in the MPC. Additionally, the spent fuel pool filtration
system would capture any debris that remained in the pool.
Comment K.24: One commenter asked why the volume of water removed
from the cask is recorded and why this is not done for other cask
designs.
Response: The purpose of recording the volume of water removed from
the canister is to identify the open volume in the canister. This open
volume is used to calculate the amount of helium to be added to the
cask following vacuum drying. The procedure and equation used for this
procedure is discussed on page 8.1-21 in the HI-STORM SAR. The comment
concerning other cask designs is beyond the scope of this rule.
Comment K.25: One commenter stated that a detailed procedure on
mitigating the possibility of fuel crud particulates dispersal should
be included in the documents and that the procedure should not be site-
specific.
Response: NRC disagrees with the comment. The generic unloading
procedures for the HI-STORM 100 system are designed to mitigate crud
dispersal. However, each cask user will need to develop detailed
unloading procedures that incorporate the ALARA objectives of its site-
specific radiation protection program. NRC expects the cask user to
consider the specific characteristics of its fuel, including crud
phenomena, when developing these procedures.
Comment K.26: One commenter asked how the utilities are required to
document that they will not lift the overpack any higher than 11 inches
and that the receiving surface hardness does not exceed that analyzed
in the SAR. The commenter stated that the criteria should be clarified
and which surface should be indicated.
Response: The receiving surface is the top of the storage pad as
clearly stated in Sections 3.4.2 and 11.2.3.2 of the SER and described
in Section 3.4.10 of the SAR. Users of the HI-STORM 100 system are
required to meet Appendices A and B of the CoC that list the design
parameters for surface hardness and the restriction for lifting height.
Furthermore, the cask users are required to develop detailed written
operating procedures. The restriction on lifting height must be
incorporated into the operating procedures subject to NRC inspection.
Comment K.27: One commenter stated that Condition 8 should remain
in the CoC.
Response: The NRC agrees with the comment. Condition 8 has not been
removed from the CoC. Under Condition 8, Certificate holders who wish
to make changes to the CoC, including Appendices A and B, must submit
an application for amendment of the Certificate.
Comment K.28: One commenter asked how upending/downending of the
transfer cask affected the water in the neutron shield, how the
licensee knows the shield is full, what happens to the contents of the
cask when the position changes, what are the stresses and pressures,
and if the debris in damaged fuel containers goes through the screen.
[[Page 25263]]
Response: The structural, shielding, and confinement functions of
the transfer cask are not affected during movement of the cask. The
neutron shield will normally be filled through the drain valve at the
bottom of the water jacket and is considered full as water exits the
vent port at the top of the water jacket. The vent plug is then
installed to retain the water in the jacket. During the upending and
downending of the transfer cask, water remains within the neutron
shield and fuel debris remains within the confinement boundary of the
MPC. The structural evaluation in the SAR showed all the stresses and
pressures to remain within allowable values.
Comment K.29: One commenter stated that exceptions to the codes
should not be allowed and that the NRC should demand full code
requirements.
Response: The NRC disagrees with this comment. Exceptions
(alternatives) to the ASME Code specifications may be granted by the
NRC staff on a case-by-case basis. During the NRC staff review of a
proposed alternative, the applicant must demonstrate that the proposed
alternative to the Code satisfies one of the following criteria: (1)
The alternative provides an acceptable level of quality and safety, or,
(2) compliance with a specific Code requirement would result in
hardship or unusual difficulty without a compensating increase in the
level of quality or safety.
Comment K.30: One commenter stated that videos should not be used
as a permanent record. The commenter stated that black and white photos
and negatives should be used and that the negatives should be kept in
museum qualified storage. The commenter asked what method is best to
document weld integrity and how the records are stored. The NRC should
have specific criteria for record keeping requirements.
Response: The NRC disagrees with the comment. The NRC's regulations
do not explicitly require specific criteria for record keeping to
document weld integrity by the applicant. A permanent record of
completed welds will be made using video, photographic, or other means
that can provide a retrievable record of weld integrity. As per
accepted industry practice, the record is typically in color format, in
order to capture the red dye typically used for PT examinations. The
general licensee's QA program will specify the types of records and how
the records are to be stored.
Comment K.31: One commenter stated that even if the overpack
baseplates, shell, pedestal shell, and radial plates have large margins
of safety in the design, they should still be examined to code.
Response: Holtec has committed to inspect the welds of the overpack
baseplate to the shell, pedestal shell, and radial plates under ASME
Code Section V, Article 9. Weld inspection acceptance criteria meet the
requirements in ASME Section III, Subsection NF-5360.
Comment K.32: One commenter asked why a mobile lifting device is
used and why it is not required to meet the requirements of NUREG-0612,
Section 5.1.6(2) for new cranes. If a new crane is necessary to meet
the requirements, the utilities should get one and not be allowed to
lower requirements.
Response: A mobile lifting device is an alternative option to a
stationary lifting device that may be used in a CTF. The decision to
use either a mobile or stationary lifting device would be made by the
cask users and would be based on their plant's site-specific needs.
NUREG-0612, Section 5.1.6(2) specifies that new cranes should be
designed to meet NUREG-0554, ``Single-Failure-Proof Cranes for Nuclear
Power Plants.'' These requirements are not applicable to mobile lifting
devices which are not single-failure-proof; therefore, mobile lifting
devices are exempted from this particular requirement in NUREG-0612. To
ensure that the mobile lifting device has the equivalent level of
safety as a single-failure-proof crane, additional conditions in CoC,
Appendix B, Sections 3.5.2.2.1, 3.5.2.2.2, and 3.5.2.2.4 were imposed.
Comment K.33: One commenter stated that a discussion on the cask
transfer facility should be included in the SER, and that the public
should not have to read the SAR to understand the generic design. The
commenter requested that this part of the cask transfer facility be
resubmitted with a complete clear design with specific criteria.
Response: The NRC disagrees with the comment. SER Section 1.1
discusses the CTF in a level of detail appropriate for an SER. The
detailed design and operating criteria for the CTF are given in SAR
Section 2.3.3.1. This satisfies 10 CFR 72.24, which requires that the
SAR contain information on structures, systems, and components
important to safety in sufficient detail for the NRC staff to make its
regulatory finding. Repeating this information in the SER is not
necessary. The NRC disagrees that cask transfer facility should be
resubmitted with a complete clear design with specific criteria. The
specific criteria for the CTF are already given in CoC, Appendix B,
Section 3.4, and SAR Section 2.3.3.1. As discussed in the response to
E.30, NRC found it unnecessary to approve a specific CTF design.
Comment K.34: One commenter recommended that Section 3.5.7 of the
SER be revised to reflect that transport of the HI-TRAC transfer cask
in the vertical orientation is permitted. The comment also recommended
that ``50 inches'' be changed to ``42 inches'' to be consistent with TS
Table 5-1.
Response: The NRC agrees with the comment. The SER has been
modified to reflect that transport of the HI-TRAC transfer cask in the
vertical orientation is permitted. The horizontal lifting height per TS
Table 5-1 will be corrected to 42 inches to correct the typographical
error.
Comment K.35: One commenter recommended that Section 9.1.2.2.b of
the SER be revised to delete ``(either to the fuel pool or the site
licensee's off-gas system)'' because users may or may not have these
systems at their plants.
Response: The NRC agrees with the comment. It is up to the cask
users to develop the specific procedures for venting the MPC and to
determine the appropriate location under their plant's waste gas
handling system design and radiation protection program. Section
9.1.2.2.b of the SER has been modified as recommended.
Summary of Final Revisions
As a result of the NRC staff's response to public comments, or to
rectify issues identified during the comment period, TSs 5.2.f and
5.2.j have been modified (see comment J.10). The NRC staff has also
updated the CoC, including Appendix B, and has removed the bases
section from the TSs attached to the CoC to ensure consistency with
NRC's format and content. The NRC staff has also modified its SER. In
addition, the NRC staff has modified the rule language by changing the
word ``Certification'' to ``Certificate'' to clarify that it is
actually the Certificate that expires.
Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' approved by the Commission on June 30, 1997,
and published in the Federal Register on September 3, 1997 (62 FR
46517), this rule is classified as compatibility Category ``NRC.''
Compatibility is not required for Category ``NRC'' regulations. The NRC
program elements in this category are those that relate directly to
areas of regulation reserved to the NRC by the Atomic Energy Act of
1954, as amended (AEA), or the provisions of Title 10 of the Code of
Federal Regulations. Although an Agreement State may not adopt program
[[Page 25264]]
elements reserved to NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with the
particular State's administrative procedure laws, but does not confer
regulatory authority on the State.
Finding of No Significant Environmental Impact: Availability
Under the National Environmental Policy Act of 1969, as amended,
and the Commission's regulations in Subpart A of 10 CFR Part 51, the
NRC has determined that this rule is not a major Federal action
significantly affecting the quality of the human environment and
therefore, an environmental impact statement is not required. This
final rule adds an additional cask to the list of approved spent fuel
storage casks that power reactor licensees can use to store spent fuel
at reactor sites without additional site-specific approvals from the
Commission. The environmental assessment and finding of no significant
impact on which this determination is based are available for
inspection at the NRC Public Document Room, 2120 L Street NW. (Lower
Level), Washington, DC. Single copies of the environmental assessment
and finding of no significant impact are available from Merri Horn,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555, telephone (301) 415-8126,
e-mail mlh1@nrc.gov.
Paperwork Reduction Act Statement
This final rule does not contain a new or amended information
collection requirement subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the
Office of Management and Budget, approval number 3150-0132.
Public Protection Notification
If a means used to impose an information collection does not
display a currently valid OMB control number, the NRC may not conduct
or sponsor, and a person is not required to respond to, the information
collection.
Voluntary Consensus Standards
The National Technology Transfer Act of 1995 (Pub. L. 104-113)
requires that Federal agencies use technical standards that are
developed or adopted by voluntary consensus standards bodies unless the
use of such a standard is inconsistent with applicable law or otherwise
impractical. In this final rule, the NRC is adding the Holtec
International HI-STORM 100 cask system to the list of NRC-approved cask
systems for spent fuel storage in 10 CFR 72.214. This action does not
constitute the establishment of a standard that establishes generally-
applicable requirements.
Regulatory Analysis
On July 18, 1990 (55 FR 29181), the Commission issued an amendment
to 10 CFR Part 72. The amendment provided for the storage of spent
nuclear fuel in cask systems with designs approved by the NRC under a
general license. Any nuclear power reactor licensee can use cask
systems with designs approved by the NRC to store spent nuclear fuel if
it notifies the NRC in advance, the spent fuel is stored under the
conditions specified in the cask's CoC, and the conditions of the
general license are met. In that rule, four spent fuel storage casks
were approved for use at reactor sites and were listed in 10 CFR
72.214. That rule envisioned that storage casks certified in the future
could be routinely added to the listing in 10 CFR 72.214 through the
rulemaking process. Procedures and criteria for obtaining NRC approval
of new spent fuel storage cask designs were provided in 10 CFR Part 72,
Subpart L.
The alternative to this action is to withhold approval of this new
design and issue a site-specific license to each utility that proposes
to use the casks. This alternative would cost both the NRC and
utilities more time and money for each site-specific license.
Conducting site-specific reviews would ignore the procedures and
criteria currently in place for the addition of new cask designs that
can be used under a general license, and would be in conflict with NWPA
direction to the Commission to approve technologies for the use of
spent fuel storage at the sites of civilian nuclear power reactors
without, to the maximum extent practicable, the need for additional
site reviews. This alternative also would tend to exclude new vendors
from the business market without cause and would arbitrarily limit the
choice of cask designs available to power reactor licensees. This final
rule will eliminate the above problems and is consistent with previous
Commission actions. Further, the rule will have no adverse effect on
public health and safety.
The benefit of this rule to nuclear power reactor licensees is to
make available a greater choice of spent fuel storage cask designs that
can be used under a general license. The new cask vendors with casks to
be listed in 10 CFR 72.214 benefit by having to obtain NRC certificates
only once for a design that can then be used by more than one power
reactor licensee. The NRC also benefits because it will need to certify
a cask design only once for use by multiple licensees. Casks approved
through rulemaking are to be suitable for use under a range of
environmental conditions sufficiently broad to encompass multiple
nuclear power plants in the United States without the need for further
site-specific approval by NRC. Vendors with cask designs already listed
may be adversely impacted because power reactor licensees may choose a
newly listed design over an existing one. However, the NRC is required
by its regulations and NWPA direction to certify and list approved
casks. This rule has no significant identifiable impact or benefit on
other Government agencies.
Based on the above discussion of the benefits and impacts of the
alternatives, the NRC concludes that the requirements of the final rule
are commensurate with the Commission's responsibilities for public
health and safety and the common defense and security. No other
available alternative is believed to be as satisfactory, and thus, this
action is recommended.
Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996, the NRC has determined that this action is not a
major rule and has verified this determination with the Office of
Information and Regulatory Affairs, Office of Management and Budget.
Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this rule will not, if
promulgated, have a significant economic impact on a substantial number
of small entities. This rule affects only the licensing and operation
of nuclear power plants, independent spent fuel storage facilities, and
Holtec International. The companies that own these plants do not fall
within the scope of the definition of ``small entities'' set forth in
the Regulatory Flexibility Act or the Small Business Size Standards set
out in regulations issued by the Small Business Administration at 13
CFR Part 121.
Backfit Analysis
The NRC has determined that the backfit rule (10 CFR 50.109 or 10
CFR 72.62) does not apply to this rule because this amendment does not
involve any provisions that would impose backfits as defined in the
backfit
[[Page 25265]]
rule. Therefore, a backfit analysis is not required.
List of Subjects in 10 CFR Part 72
Administrative practice and procedure, Hazardous waste, Nuclear
materials, Occupational safety and health, Penalities, Radiation
protection, Reporting and recordkeeping requirements, Security
measures, Spent fuel, Whistleblowing.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 553; the NRC is adopting the
following amendments to 10 CFR part 72.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE
1. The authority citation for Part 72 continues to read as follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat.
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 10d-
48b, sec. 7902, 10b Stat. 31b3 (42 U.S.C. 5851); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135,
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153,
10155, 10157, 10161, 10168).
Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),
10168(c),(d)). Section 72.46 also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2244, (42 U.S.C. 10101,
10137(a), 10161(h)). Subparts K and L are also issued under sec.
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252
(42 U.S.C. 10198).
2. In Section 72.214, Certificate of Compliance 1014 is added to
read as follows:
Sec. 72.214 List of approved spent fuel storage casks.
* * * * *
Certificate Number: 1014.
SAR Submitted by: Holtec International.
SAR Title: Final Safety Analysis Report for the HI-STORM 100 Cask
System.
Docket Number: 72-1014.
Certificate Expiration Date: June 1, 2020.
Model Number: HI-STORM 100.
Dated at Rockville, Maryland, this 12th day of April, 2000.
For the Nuclear Regulatory Commission.
Frank J. Miraglia, Jr.,
Acting Executive Director for Operations.
[FR Doc. 00-10393 Filed 4-28-00; 8:45 am]
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