[Federal Register Volume 87, Number 220 (Wednesday, November 16, 2022)]
[Notices]
[Pages 68747-68752]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2022-24877]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 72-1014, 72-51, 50-247 and 50-286; NRC-2022-0152]
Holtec Decommissioning International, LLC, Indian Point Energy
Center, Independent Spent Fuel Storage Installation
AGENCY: Nuclear Regulatory Commission.
ACTION: Exemption; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing an
exemption in response to a request submitted by Holtec Decommissioning
International, LLC (HDI), on behalf of Holtec Indian Point 2, LLC and
Holtec Indian Point 3, LLC on March 24, 2022. This exemption would, if
granted, permit HDI to load up to three MPC-32Ms, using Amendment No.
15 of the Holtec International Certificate of Compliance (CoC) No. 1014
for the HI-STORM 100 storage system, with either up to 32 fuel
assemblies each containing either a Californium-252 (Cf-252) or an
Antimony-Beryllium (Sb-Be) neutron source assemblies (NSA) with
sufficient cooling time, or a combination of up to five Plutonium-
Beryllium (Pu-Be) NSAs and up to all of the remaining basket locations
with fuel assemblies each containing either a Cf-252 or an Sb-Be NSA
with sufficient cooling time. Further, it would permit HDI to load the
fuel assemblies containing either Cf-252 or Sb-Be NSAs in any location
in the basket and the fuel assemblies containing Pu-Be NSAs such that
one is located in the center of the basket and no more than one NSA is
located in each of the four basket quadrants.
DATES: The exemption was issued on November 7, 2022.
ADDRESSES: Please refer to Docket ID NRC-2022-0152 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly available information related to this document
using any of the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2022-0152. Address
questions about Docket IDs to Stacy Schumann; telephone: 301-415-0624;
email: [email protected]. For technical questions, contact the
individual listed in the For Further Information Contact section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or
by email to [email protected]. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
NRC's PDR: You may examine and purchase copies of public
documents, by appointment, at the NRC's PDR, Room P1 B35, One White
Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. To make
an appointment to visit the PDR, please send an email to
[email protected]
[[Page 68748]]
or call 1-800-397-4209 or 301-415-4737, between 8:00 a.m. and 4:00 p.m.
Eastern Time (ET), Monday through Friday, except Federal holidays.
FOR FURTHER INFORMATION CONTACT: Chris Allen, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone: 301-415-6877; email:
[email protected].
SUPPLEMENTARY INFORMATION:
I. Background
Holtec Decommissioning International, LLC (HDI), holds a general
license for the Indian Point Energy Center Independent Spent Fuel
Storage Installation (ISFSI) under provisions in part 72 of title 10 of
the Code of Federal Regulations (10 CFR), ``Licensing Requirements for
the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive
Waste, and Reactor-Related Greater Than Class C Waste.'' Under 10 CFR
72.212(a)(2), (b)(3), (b)(5)(i), (b)(11) and 72.214, a general licensee
may store spent fuel in a cask, so long as it is one of the approved
casks listed in 10 CFR 72.214 and the general licensee conforms to the
terms, conditions, and specifications of the relevant certificate of
compliance (CoC) or amended CoC. HDI has stated that it plans to use
the HI-STORM 100 dry storage system, CoC No. 1014, Amendment No. 15 in
an upcoming spent fuel loading campaign.
II. Request/Action
By letter dated March 24, 2022, as supplemented on June 17, 2022,
HDI, on behalf of Holtec Indian Point 2, LLC and Holtec Indian Point 3,
LLC, requested an exemption under 10 CFR 72.7. HDI further clarified
its request during a Microsoft Teams call on September 20, 2022. HDI
specifically requested an exemption from the requirements of 10 CFR
72.212(b)(3), and the portion of 10 CFR 72.212(b)(11) that states
``[t]he licensee shall comply with the terms, conditions, and
specifications of the certificate of compliance (CoC).'' The exemption
request would permit, if granted, HDI to load up to three MPC-32Ms,
using Amendment No. 15 of the Holtec International Certificate of
Compliance (CoC) No. 1014 for the HI-STORM 100 storage system, with
either up to 32 fuel assemblies each containing either a Californium-
252 (Cf-252) or an Antimony-Beryllium (Sb-Be) NSA with sufficient
cooling time, or a combination of up to five fuel assemblies each
containing a Plutonium-Beryllium (Pu-Be) NSA and up to all of the
remaining basket locations with fuel assemblies each containing either
a Cf-252 or an Sb-Be NSA with sufficient cooling time. Further, as
discussed later, it would permit HDI to load the fuel assemblies
containing either Cf-252 and Sb-Be NSAs in any location in the basket
and the fuel assemblies containing Pu-Be NSAs such that one is located
in the center of the basket and no more than one is located in each of
the four basket quadrants. Additionally, although HDI's analysis
included information about polonium beryllium (Po-Be) NSAs, based on
its September 20, 2022, Microsoft Teams call, HDI indicated that they
only wanted to load Cf-252 and Sb-Be NSAs.
Although HDI only requested exemptions from 10 CFR 72.212(b)(3) and
(b)(11), to carry out this action, the NRC would also need to grant
exemptions from 72.212(a)(2), (b)(5)(i), and 72.214. Consequently, in
evaluating the request, the NRC also considered, pursuant to its
authority in 10 CFR 72.7, exempting HDI from similar requirements in 10
CFR 72.212(a)(2), 10 CFR 72.212(b)(5)(i); and 10 CFR 72.214, ``List of
Approved Spent Fuel Storage Casks.'' For clarity, when this Federal
Register notice refers to HDI's requested exemption, it means both the
two provisions from which HDI requested exemption and the additional
provisions from which the NRC staff is considering exempting HDI on its
own initiative.
III. Discussion
Pursuant to 10 CFR 72.7, the Commission may, upon application by
any interested person or upon its own initiative, grant such exemptions
from the requirements of the regulations of 10 CFR part 72 as it
determines are authorized by law and will not endanger life or property
or the common defense and security, and are otherwise in the public
interest.
The NRC staff prepared a safety evaluation report to document its
safety evaluation of the requested exemption. As summarized in this
document, the NRC's safety review concluded that the requested
exemption meets the requirements for issuance in 10 CFR 72.7.
A. The Exemption Is Authorized by Law
The Commission has the legal authority to issue exemptions from the
requirements of 10 CFR part 72 as provided in 10 CFR 72.7. Issuance of
this exemption is consistent with the Atomic Energy Act of 1954, as
amended, and is not otherwise inconsistent with NRC's regulations or
other applicable laws. Therefore, issuance of the exemption is
authorized by law.
B. Will Not Endanger Life or Property or the Common Defense and
Security
The staff reviewed HDI's exemption request and concludes, as
discussed further, that the proposed exemption from certain
requirements of 10 CFR part 72 will not cause the HI-STORM 100 storage
cask to encounter conditions beyond those for which it has already been
evaluated and demonstrated to meet the applicable safety requirements
in 10 CFR part 72. The staff followed the guidance in NUREG-2215,
``Standard Review Plan for Spent Fuel Dry Storage Systems and
Facilities,'' April 2020, to complete its safety evaluation.
Safety Review of the Requested Exemption
HDI submitted an exemption request to deviate from the requirement
in CoC No. 1014, Appendix D, table 2.1-1, section V, ``MPC MODEL: MPC-
32M,'' Item C of Amendment No. 15 for CoC No. 1014 only permits general
licensees to load a single NSA per cask. Further, per Final Safety
Analysis Report (FSAR) table 2.II.1.1, Rev. 22, the single NSA must be
located in a cell in the inner part of the basket (i.e., fuel storage
location 13, 14, 19, or 20). The staff reviewed the exemption request
and concluded that the proposed exemption from certain requirements of
10 CFR part 72 will not cause the HI-STORM 100 storage system to
encounter conditions beyond those for which it has been evaluated and
demonstrated to meet the applicable safety requirements in 10 CFR part
72.
The staff determined that the presence of additional NSAs or the
presence of those NSAs in different locations throughout the basket
will not cause the bounding canister weight previously evaluated in
approving Amendment No. 15 to be exceeded, making a structural
evaluation unnecessary. Further, the staff determined that the decay
heat contribution from activated metal associated with the NSAs at
issue in the specified locations is negligible compared to the decay
heat from the fuel assembly.
Consequently, the staff determined that a thermal evaluation is
unwarranted. Since the NSAs are located inside the confinement boundary
of the multi-purpose canister (MPC) and changing the number of NSAs, or
their locations, will not change that fact, a confinement evaluation is
also not necessary. In addition, increasing the neutron source terms by
adding NSAs in different locations does not increase the multiplication
factor. Therefore, criticality safety is not affected, and a
criticality evaluation is
[[Page 68749]]
unnecessary. Therefore, shielding is the only area potentially affected
by the requested exemption.
Shielding
The current CoC authorizes general licensees to load only a single
fuel assembly containing an NSA per cask, and that fuel assembly must
be loaded in a cell within the inner part of the basket (i.e., fuel
storage location 13, 14, 19, or 20) because NSAs can have a significant
neutron source term. The applicant developed a quantitative analysis
that explicitly evaluated the neutron dose rates associated with
storing more than one fuel assembly containing an NSA per cask to
support new loading requirements. In its analysis, the applicant
evaluated two possible high-level loading scenarios: a maximum of 32
fuel assemblies each containing an NSA and a maximum of five fuel
assemblies each containing a Pu-Be NSA.
For both scenarios, the applicant considered three primary NSA
types in its evaluation: Cf-252, Pu-Be, and Po-Be. During the September
20, 2020, Microsoft Teams call, HDI indicated that they only wanted to
load Cf-252 and Sb-Be NSAs. Consequently, the staff did not consider
Po-Be NSAs in its evaluation of this exemption. Cf-252 and Pu-Be NSAs
have half-lives of 2.646 years and 87.7 years, respectively. The
applicant also considered a secondary NSA type, Sb-Be, with a half-life
of 60.2 days. For Cf-252, which decays by neutron emission, the
analysis identified that the neutron source strength will reduce
gradually over time because the half-life is on the order of a few
years; neither long enough for the source strength to remain relatively
constant, nor short enough for the reduction to be quick. For Pu-Be,
which generates neutrons when the beryllium absorbs an alpha particle
emitted by the plutonium, the analysis identified that the neutron
source strength will remain essentially the same as when the NSA was
manufactured (i.e., it will not reduce significantly over time) because
the half-life for plutonium is very long. For Sb-Be, which produces
neutrons when the beryllium interacts with a high energy gamma emitted
by activated antimony (i.e., antimony that has absorbed neutrons), the
analysis identified that the neutron source strength will reduce very
quickly over time because of the short half-life of the activated
antimony.
In evaluating the scenario of loading a maximum of 32 fuel
assemblies containing NSAs, the applicant determined, using the initial
source strength and the half-life values in the previous paragraph,
that after seven half-lives the neutron source strength of a fuel
assembly containing either a Cf-252 or an Sb-Be NSA is negligibly
higher than the neutron source strength of a design basis fuel
assembly. Therefore, the applicant asserted that, after seven half-
lives, the presence of either a Cf-252 or an Sb-Be NSA within a design
basis fuel assembly will not significantly increase the dose rate from
a design basis fuel assembly. Consequently, the applicant concluded
that up to 32 fuel assemblies each containing either a Cf-252 or an Sb-
Be NSA can be loaded per basket, and that they can be loaded into any
basket location.
Staff reviewed the applicant's approach. In reviewing this
approach, staff found that the applicant could load up to 32 fuel
assemblies each containing either a Cf-252 or an Sb-Be NSA--with those
32 fuel assemblies having any combination of Cf-252 and Sb-Be NSAs--and
that the neutron source strength of each fuel assembly with either a
Cf-252 NSA or an Sb-Be NSA increased by only a small amount,
approximately 2 x 10-6 neutrons per second, after seven
half-lives relative to a design basis fuel assembly. Because this
increase is so small, after seven half-lives, the dose rate of a
canister containing 32 fuel assemblies with either Cf-252 or Sb-Be NSAs
that have undergone seven half-lives of decay will be very similar to
the dose rate of a container containing 32 design basis fuel
assemblies. More specifically, accounting for statistical
uncertainties, dose rates would potentially increase a millirem/hr or
less, if at all, under both normal and accident conditions. The NRC
staff considers dose rate increases of this magnitude to be negligible
relative to the dose rates from design basis fuel assemblies.
Therefore, the staff determined that the analysis demonstrated that
dose rates under both normal and accident conditions would increase
negligibly by the addition of 32 fuel assemblies containing either Cf-
252 or Sb-Be NSAs after seven half-lives of decay time. Further,
because a canister loaded with 32 fuel assemblies each containing
either a Cf-252 or Sb-Be NSA would have an NSA loaded in every fuel
loading location and because the effect on dose would be negligible,
the NRC staff concludes that loading fuel assemblies containing either
a Cf-252 or an Sb-Be NSA in any location in the basket would have a
negligible effect on dose.
In evaluating loading a maximum of five fuel assemblies each
containing a Pu-Be NSA the applicant performed dose rate calculations
assuming each NSA had the design basis fuel assembly neutron source
term in HI-STORM 100 FSAR table 5.2.15 rather than the actual source
strength of an NSA. The applicant evaluated dose rates using the
general-purpose, continuous-energy, generalized-geometry, time-
dependent Monte Carlo N-Particle (MCNP) code. The applicant used MCNP5
version 1.41 to model the MPC-32M, with up to five NSAs per basket, in
both the HI-TRAC Version MS and the HI-STORM 100S Version E overpack.
The MCNP model located one NSA in the center of the MPC-32M (i.e., cell
locations 13, 14, 19 and 20 of appendix D, figure 2.1-1). In addition,
the model located the remaining four NSAs on the basket periphery with
one NSA in each basket quadrant.
The applicant calculated the maximum dose rate from the NSAs in the
fuel assembly and not the maximum total dose rate from the fuel
assembly and the NSA. The applicant asserted that this approach would
result in conservative dose rates because the maximum dose rate due to
the design basis fuel assembly may be in a different location (e.g.,
the midplane of the overpack radial surface) from the maximum dose rate
due to the NSAs. The applicant calculated dose rates at the same
surface and one-meter locations for design basis fuel under normal
conditions as reported in HI-STORM 100 FSAR tables 5.II.1.1 and
5.II.1.3. Additionally, the applicant evaluated the dose rate at 100
meters for design basis fuel in the HI-TRAC under accident conditions
at the same locations as reported in HI-STORM 100 FSAR table 5.II.1.4.
The analysis determined the maximum dose rate increase under normal
conditions due to adding four fuel assemblies each containing a Pu-Be
NSA, in addition to the fuel assembly containing an NSA authorized by
CoC No. 1014, at the following locations: the overpack surface, one
meter from the overpack surface, the HI-TRAC surface, and one meter
from the HI-TRAC surface. The analysis calculated the following dose
rate increases at these locations: 3.44 millirem per hour (mrem/hr),
0.78 mrem/hr, 1099.92 mrem/hr and 122.69 mrem/hr respectively. Finally,
the analysis determined the maximum dose rate increase under accident
conditions due to adding four NSAs, in addition to the NSA authorized
by CoC No. 1014, at 100 meters from the HI-TRAC is 0.27 mrem/hr.
In conducting its evaluation, the applicant assumed the Pu-Be NSA
source strength equaled the design basis fuel assembly source strength
of 1.4 x 10\9\ neutrons per second. The staff
[[Page 68750]]
determined that this approach is conservative because the initial
source term of a Pu-Be NSA is approximately 1.5 x 10\6\ neutrons per
second which is less than the value HDI used. Because the MCNP code is
a standard tool in the nuclear industry for performing Monte Carlo
criticality safety and radiation shielding calculations, the staff
found MCNP an acceptable code for this application. Because the
exemption request is limited to fuel stored in an MPC-32M, which can
only be stored in the HI-STORM 100S Version E overpack, and because the
HI-TRAC MS can only be used with the HI-STORM 100S Version E overpack,
staff found it acceptable to limit the MCNP analyses to the HI-TRAC MS
and the HI-STORM 100S Version E overpack. In addition, the applicant
calculated the dose rates related to this exemption at the same
locations at which it calculated the dose rates for HI-STORM Amendment
No. 15. In issuing Amendment No. 15, staff determined the dose rates at
these locations satisfied as low as is reasonably achievable (ALARA)
principles, where relevant, and demonstrated compliance with 10 CFR
72.104 and 10 CFR 72.106, as well as 10 CFR part 20, as documented in
Section 6 of the SER staff prepared to support issuance of Amendment
No. 15. Nothing about this exemption would affect, or in any way make
inapplicable, the staff's previous finding that calculating the dose
rate at those locations is acceptable. Therefore, staff finds these
locations are appropriate for calculating dose rates associated with
this exemption.
Further, the staff reviewed the applicant's approach of only
calculating the maximum dose rate caused by the NSAs in the fuel
assemblies and not the overall maximum dose rate. The total dose rate
from two different sources (i.e., the design basis fuel assembly and
the NSA) is simply the sum of the individual dose rates. Consequently,
by taking the dose rate caused by design basis fuel assemblies in the
canister, which are found in FSAR tables 5.II.1.1, 5.II.1.3 and
5.II.1.4 and adding them to the dose rate caused by the NSAs within
fuel assemblies, the staff was able to evaluate the overall maximum
dose rate as part of its review. Therefore, the staff also found
acceptable the applicant's approach of only calculating the maximum
dose rate due to fuel assemblies containing NSAs.
When the staff approved the MPC-32M, the HI-TRAC MS and the HI-
STORM 100S Version E overpack, the staff identified two accident
conditions that increased the dose at the controlled area boundary: (1)
the draining of the neutron shield water jacket for the transfer cask
and (2) a non-mechanistic tipover of the overpack which exposes the
bottom of the cask. As discussed in the SER approving the HI-STORM 100S
Version E overpack, staff found it very unlikely that the Version E
overpack would tip over. Nothing about this exemption would affect that
conclusion. Therefore, the staff found the applicant's approach of
modeling the HI-TRAC with the assumed loss of the neutron absorber as
the bounding accident acceptable for this evaluation.
NRC staff concluded that the increased dose rates under normal
conditions from the presence of up to five fuel assemblies containing
Pu-Be NSAs are acceptable for the HI-STORM overpack because the dose
rate increase is less than a mrem/hr for all locations except at the
midplane of the radial surface on the overpack surface where it
increased by less than four mrem/hr. Relative to the dose rates from
loading the canister as already-approved, staff considers dose rate
increases of this magnitude negligible. Additionally, the dose rate
increases at a distance of one meter are even less than the dose rate
increases at the surface. Thus, relative to the dose rates from loading
the canister as already approved, the staff also considers these dose
rate increases to be negligible. Further, the HI-TRAC MS dose rates
increased by less than ten percent compared to the dose rates in HI-
STORM 100 FSAR table 5.II.1.3 at all locations both on the HI-TRAC MS
surface and one meter from the HI-TRAC MS surface except at the HI-TRAC
MS radial surface midplane where the dose rate increased by 28 percent
(i.e., 1099.92 mrem/hr). Staff considers the dose rate increase at the
HI-TRAC MS radial surface midplane a very localized effect due to the
reduced neutron shielding capability of the HI-TRAC MS compared to the
HI-STORM 100S Version E overpack. The staff considers the HI-TRAC MS
dose rate increases, including the increase at the radial surface
midplane, acceptable for the following reasons. First, radiological
workers would only be exposed to these increased dose rates for
relatively short periods of time. Second, members of the public will be
exposed to even lower dose rates since 10 CFR 72.106(b) requires a
minimum distance of 100 meters between spent fuel and members of the
public and dose rates decrease as distance increases. NRC staff also
determined that an increase in the HI-TRAC dose rates of less than ten
percent compared to the dose rates in HI-STORM 100 FSAR table 5.II.1.4
for the HI-TRAC MS accident condition dose rates due to the presence of
up to five fuel assemblies containing Pu-Be NSAs is acceptable because
staff confirmed through hand calculations that the dose at 100 meters
meets the 10 CFR 72.106 requirement assuming a 30-day duration.
Finally, after adding the dose rates considered when issuing CoC 1014,
Amendment No. 15 to the dose rate increases that would result from
approving this exemption, staff finds that canisters loaded in
accordance with this exemption will continue to satisfy overall dose
limits of 10 CFR 72.104 for normal conditions, 10 CFR 72.106 for
accident conditions, and the limits in 10 CFR part 20. These
conclusions only apply, however, when the fuel assemblies containing
the Pu-Be NSAs are loaded such that one is located in the center of the
basket (i.e., fuel storage location 13, 14, 19, or 20) and no more than
one is located in each of the four basket quadrants.
As referenced earlier, if granted, this exemption would permit HDI
to load a fuel canister with up to five fuel assemblies each containing
a Pu-Be NSA and up to all of the remaining basket locations with fuel
assemblies each containing either a Cf-252 or an Sb-Be NSA that has
decayed for at least seven half-lives. HDI did not provide an analysis
of this specific configuration. That said, as discussed previously,
staff has already analyzed a canister loaded with five fuel assemblies
each containing a Pu-Be NSA and a canister loaded with 32 fuel
assemblies each containing either a Cf-252 or an Sb-Be NSA that has
decayed for at least seven half-lives. Staff concluded that the neutron
source strength of a fuel assembly with either a Cf-252 NSA or an Sb-Be
NSA increased by only a small amount--approximately 2 x 10-6
neutrons per second--after seven half-lives relative to a design basis
fuel assembly. As discussed before, the staff concluded that that
source strength increase was so small that the neutron dose rate
increase, if any, associated with loading a canister with 32 fuel
assemblies each containing either a Cf-252 or an Sb-Be NSA would be
negligible. As the dose rate increase from loading a canister with 32
fuel assemblies each containing either a Cf-252 or an Sb-Be NSA would
be negligible, it follows that adding 27 fuel assemblies each
containing either a Cf-252 or an Sb-Be NSA that has undergone seven
half-lives of decay, will have a similarly negligible effect on dose
rate because the increase in neutron source strength will be even
smaller than when loading 32 such fuel assemblies. Consequently,
loading 27
[[Page 68751]]
fuel assemblies each containing either a Cf-252 or an Sb-Be NSA that
has undergone seven half-lives of decay into a canister with five fuel
assemblies each containing a Pu-Be NSA will negligibly increase the
neutron dose rate, if at all, beyond the neutron dose rate associated
with loading just five fuel assemblies each containing a Pu-Be NSA.
Therefore, the staff determined that under this loading scenario--up to
five fuel assemblies each containing a Pu-Be NSA and up to 27 fuel
assemblies, each containing a Cf-252 of Sb-Be NSA--the dose rates under
both normal and accident conditions will continue to satisfy overall
dose limits of 10 CFR 72.104 for normal conditions, 10 CFR 72.106 for
accident conditions, and the limits in 10 CFR part 20. Finally, the
staff determined that this loading scenario, along with the scenario of
loading 32 fuel assemblies each containing a Cf-252 or an Sb-Be NSA
bound all loading scenarios that this exemption, if granted, would
permit because the other loading scenarios will be a version of these
two scenarios with fewer fuel assemblies containing NSAs and,
therefore, less dose.
As a final note, the staff's analysis of a canister loaded with
five fuel assemblies each containing a Pu-Be NSA depends on HDI's dose
rate analysis. As discussed previously, that analysis was based on a
model with one NSA in the center of the MPC-32M (i.e., cell locations
13, 14, 19 and 20 of appendix D, figure 2.1-1) and the remaining four
NSAs on the basket periphery with one NSA in each basket quadrant.
Consequently, the staff's analysis of and conclusions about this
loading scenario--up to five fuel assemblies each containing a Pu-Be
NSA and up to 27 fuel assemblies, each containing a Cf-252 of Sb-Be
NSA--only apply when the fuel assemblies containing Pu-Be NSAs are
loaded with one in the center of the basket and a maximum of one in
each of the remaining quadrants.
Although the exemption request did not explicitly evaluate the
gamma dose associated with storing more than one NSA, the applicant
asserted that the additional gamma dose due to activation of the NSA
components will remain within the limits of 10 CFR 72.104 for normal
conditions and 10 CFR 72.106 for accident conditions. In evaluating
this assertion, staff reviewed HI-STORM 100 FSAR sections 5.2.7.1
submitted with Amendment No. 15 in which Holtec International stated
that the total Burnable Poison Rod Assembly (BPRA) activation source
term bounded the total NSA activation source term. In approving
Amendment No. 15, in SER section 6.2.2.3, the staff found the use of
the BPRA source term to represent all non-fuel hardware--including Pu-
Be, Cf-252, and Sb-Be NSAs--acceptable. Further, the SER approving
Amendment No. 15 determined that a canister loaded with 32 fuel
assemblies containing BPRAs would remain within the limits of 10 CFR
72.104 for normal conditions and 10 CFR 72.106 for accident conditions.
Because the staff found that the BPRA activation source term bounded
the NSA activation source term in approving Amendment No. 15, and
because this exemption does not change or affect that determination,
the staff determined, for this exemption request, that the gamma source
term associated with storing either five fuel assemblies each
containing a Pu-Be NSA and up to 27 fuel assemblies each containing
either a Cf-252 or an Sb-Be NSA or 32 fuel assemblies each containing
either a Cf-252 or an Sb-Be NSA in an MPC-32M canister is bounded by
the dose rates evaluated in Amendment No. 15. Therefore, because the
dose rates evaluated in Amendment No. 15 met the applicable regulatory
requirements, the staff finds that the dose due to activation of NSA
components will remain within the limits of 10 CFR 72.104 for normal
conditions, 10 CFR 72.106 for accident conditions, and the limits in 10
CFR part 20.
Finally, the staff reviewed the application from the perspective of
dose rates remaining ALARA. Staff determined that the proposed
exemption did not alter those aspects of the HI-STORM 100 system that
the SER issued with CoC No. 1014 Amendment No. 15 had indicated
contributed to a finding that ALARA had been satisfied (e.g., temporary
shielding equipment utilized during loading operations). In addition,
as explained in section 11.1.2 of the SER issued with Amendment No. 15
to CoC No. 1014, the staff found reasonable assurance that the design
of the HI-TRAC MS and the operational restrictions meet ALARA
objectives for direct radiation levels because the estimated
occupational exposure in FSAR table 10.II.3 was below the 10 CFR
20.1202(a) dose limit for an individual. For this exemption request,
staff increased the estimated occupational exposure in FSAR table
10.II.3.1 by 3.3 percent, which was the greatest increase for locations
where most operations occurred. The revised estimated occupational
exposure remained below the 10 CFR 20.1201(a) dose limit. Therefore,
consistent with these previous evaluations, the staff finds that for a
canister loaded as permitted by this exemption, the occupational doses
would remain ALARA despite the overall increase in dose.
Review of Common Defense and Security
HDI's exemption request is not related to any aspect of the
physical security or defense of the Indian Point Energy Center ISFSI.
In addition, the number of NSAs stored within a multipurpose canister
does not affect the Indian Point Energy Center ISFSI security plans.
Therefore, granting the exemption would not result in any potential
impacts to common defense and security.
As discussed earlier, the staff has evaluated the effects this
exemption would have, if granted, on shielding for the configurations
that exist during the different stages of storage operations including
under both normal and accident conditions. This evaluation includes
dose rate results which lead the staff to conclude that the HI-STORM
100 system will meet the limits in 10 CFR part 20, the 10 CFR 72.104
and 72.106 radiation protection requirements, and that ALARA principles
for occupational exposure are adequately considered and incorporated
into the HI-STORM 100 system design and operations after implementing
the exemption. The staff reached this finding based on a review that
considered the regulations, appropriate regulatory guides, applicable
codes and standards, accepted engineering practices, and the statements
and representations in the application. Based on this evaluation, the
staff concludes that granting this exemption will not endanger life,
property or the common defense and security.
D. Otherwise in the Public Interest
During a June 17, 2022, Microsoft Teams call with the NRC, the
applicant indicated that granting the requested exemption would result
in shorter operation of the spent fuel pool cleaning system. Shorter
operation of the cleaning system would generate less waste of which the
licensee would ultimately need to dispose. The staff reviewed the
information provided by HDI, and based upon the earlier stated
information, concludes that granting the requested exemption would be
in the public interest because it would result in the generation of
less low-level waste.
E. Environmental Considerations
The NRC staff also considered whether there would be any
significant environmental impacts associated with the exemption. For
this proposed action,
[[Page 68752]]
the NRC staff performed an environmental assessment pursuant to 10 CFR
51.30. The environmental assessment concluded that the proposed action
would not significantly impact the quality of the human environment.
The NRC staff concluded that the proposed action would not result in
any changes in the types or quantities of effluents that may be
released offsite, and there is no significant increase in occupational
or public radiation exposure because of the proposed action. The
environmental assessment and the finding of no significant impact was
published on October 31, 2022 (87 FR 65613).
IV. Conclusion
Based on the statements and representations provided by HDI in its
exemption request, the staff concludes that the proposed action is
authorized by law and will not endanger life, property, or the common
defense and security, and is otherwise in the public interest. As a
result, the NRC staff concludes the requested exemption meets the
requirements in 10 CFR 72.7. Therefore, the NRC staff hereby grants
HDI, an exemption from 10 CFR 72.212(a)(2), (b)(3), (b)(5)(i), (b)(11),
and 72.214, pursuant to 10 CFR 72.7, permitting HDI to load up to three
MPC-32Ms, using Amendment No. 15 for CoC No. 1014, with either up to 32
fuel assemblies each containing either a Cf-252 or an Sb-Be NSA with
sufficient cooling time, or a combination of up to five fuel assemblies
each containing a Pu-Be NSA and up to all of the remaining basket
locations with fuel assemblies each containing either a Cf-252 or an
Sb-Be NSA with sufficient cooling time. Further, it permits HDI to load
the fuel assemblies containing either Cf-252 or Sb-Be NSAs in any
location in the basket and the fuel assemblies containing Pu-Be NSAs
such that one is located in the center of the basket (i.e., fuel
storage locations 13, 14, 19, or 20) and no more than one is located in
each of the four basket quadrants.
The exemption is effective upon issuance.
V. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
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Document description ADAMS accession No.
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Issuance of Certificate of ML21118A862 (package).
Compliance No. 1014, Amendment
No. 15 for the HI-STORM 100
Multipurpose Canister Storage
System, dated May 13, 2021.
Indian Point Energy Center-- ML22083A191.
Request for Exemption from an
Allowable Contents Requirement
Contained in the Certificate of
Compliance No. 1014 for the HI-
STORM 100S Version E Cask, dated
March 24, 2022.
Indian Point Exemption ML22172A174
Environmental Assessment
Conversation Record (6-16-22),
date of contact June 16, 2022.
Neutron Source Assembly Loading ML22264A045.
Clarification Call, date of
contact September 20, 2022.
Safety Evaluation Report, dated ML22217A017.
November 7, 2022.
HI-2002444, Revision 22, Holtec ML21221A329.
International Final Safety
Analysis Report for the HI-STORM
100 Cask System, dated July 1,
2021.
------------------------------------------------------------------------
Dated: November 9, 2022.
For the Nuclear Regulatory Commission.
Yoira K. Diaz-Sanabria,
Chief, Storage and Transportation Licensing Branch, Division of Fuel
Management, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 2022-24877 Filed 11-15-22; 8:45 am]
BILLING CODE 7590-01-P