[Federal Register Volume 71, Number 237 (Monday, December 11, 2006)]
[Rules and Regulations]
[Pages 71463-71472]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-20962]
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Rules and Regulations
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Federal Register / Vol. 71, No. 237 / Monday, December 11, 2006 /
Rules and Regulations
[[Page 71463]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 72
RIN 3150-AH93
List of Approved Spent Fuel Storage Casks: NUHOMS[supreg] HD
Addition
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
-----------------------------------------------------------------------
SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to add the NUHOMS[supreg] HD cask system to the list of
approved spent fuel storage casks. This final rule allows the holders
of power reactor operating licenses to store spent fuel in this
approved cask system under a general license.
DATES: Effective Date: The final rule is effective on January 10, 2007.
ADDRESSES: Publicly available documents related to this rulemaking may
be viewed electronically on the public computers located at the NRC's
Public Document Room (PDR), Room O1F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland. The PDR reproduction contractor
will copy documents for a fee. Selected documents can be viewed and
downloaded electronically via the NRC's rulemaking Web site at http://ruleforum.llnl.gov.
Publicly available documents created or received at the NRC are
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/NRC/reading-rm/adams.html. From this site, the public can
gain entry into the NRC's Agencywide Document Access and Management
System (ADAMS), which provides text and image files of NRC's public
documents. If you do not have access to ADAMS or if there are any
problems in accessing the documents located in ADAMS, contact the NRC
PDR Reference staff at (800) 397-4209, (301) 415-4737, or by e-mail to
[email protected].
FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Federal
and State Materials and Environmental Management Programs, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
6219, e-mail [email protected].
SUPPLEMENTARY INFORMATION:
Background
Section 218(a) of the Nuclear Waste Policy Act of 1982, as amended
(NWPA), requires that ``[t]he Secretary [of the Department of Energy
(DOE)] shall establish a demonstration program, in cooperation with the
private sector, for the dry storage of spent nuclear fuel at civilian
nuclear power reactor 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, that ``[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 by
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 spent fuel
storage cask designs.
Discussion
On May 5, 2004, and as supplemented on July 6, August 16, October
11, October 28, November 19, 2004; February 18, March 7, April 14, May
20, May 24, August 16, 2005; and January 24, February 15, and September
19, 2006, the certificate holder, Transnuclear, Inc. (TN), submitted an
application to the NRC to add the NUHOMS[supreg] HD cask system to the
list of NRC-approved casks for spent fuel storage in 10 CFR 72.214. The
NUHOMS[supreg] HD System provides for the horizontal storage of high
burnup spent pressurized water reactor fuel assemblies in a Dry
Shielded Canister (DSC) that is placed in a horizontal storage module
(HSM) utilizing an OS-187H transfer cask (TC). The system is an
improved version of the Standardized NUHOMS[supreg] System described in
Certificate of Compliance (CoC) No. 1004. The NUHOMS[supreg] HD System
has been optimized for high thermal loads, limited space, and radiation
shielding performance. The -32PTH DSC included in this system is
similar to the -24PTH DSC submitted for licensing as Amendment No. 8 to
the Standardized NUHOMS[supreg] System. The -32PTH DSC will be
transferred during loading operations using the OS-187H TC. The OS-187H
TC is very similar to the OS-197 and OS-197 TCs described in the final
safety analysis report for the Standardized NUHOMS[supreg] System. The
-32PTH DSC will be stored in an HSM, designated the HSM-H. The HSM-H is
virtually identical to the HSM-H submitted for licensing as Amendment
No. 8 to the Standardized NUHOMS[supreg] System. The NRC staff
performed a detailed safety evaluation of the proposed CoC request and
found that an acceptable safety margin is maintained. In addition, the
NRC staff has determined that there continues to be reasonable
assurance that public health and safety and the environment will be
adequately protected.
The NRC published a direct final rule (71 FR 25740; May 2, 2006)
and the companion proposed rule (71 FR 25782) in the Federal Register
to add the NUHOMS[supreg] HD cask system to the listing in 10 CFR
72.214. The comment period ended on July 17, 2006. Six comment letters
were received on the proposed rule. The comments were considered to be
significant and adverse and warranted withdrawal of the direct final
rule. A notice of withdrawal was published in the Federal Register on
July 13, 2006; 71 FR 39520.
Based on NRC review and analysis of public comments, the staff has
modified, as appropriate, Technical Specifications (TS) and the
Approved Contents and Design Features, for the NUHOMS[supreg] HD
system. The staff has also modified its preliminary Safety Evaluation
Report (SER). In particular, regarding the potential for the dry
[[Page 71464]]
shielded canister to corrode in a coastal marine environment, TN
committed to specifying a weathering steel for Independent Spent Fuel
Storage Installations (ISFSIs) located near a coastal marine
environment. The staff made corresponding changes to the SER and added
a requirement to TS 4.4.1 to capture this commitment for the HSM-H.
The proposed TS and SER have been revised in response to Comment 2.
Specifically, based on questions from the staff regarding this issue,
TN committed in a letter dated September 19, 2006, to add the following
to Section 3.4.1.4 of the Safety Analysis Report (SAR) for the NUHOMS
HD design: ``If an independent spent fuel storage installation site is
located in a coastal salt water marine atmosphere, then any load-
bearing carbon steel DSC support structure rail components of any
associated HSM-H shall be procured with a minimum 0.20 percent copper
content for corrosion resistance.'' This commitment has also been
captured in NUHOMS[supreg] HD TS 4.4.1 for the HSM-H, and the staff
made corresponding changes to SER Section 3.2.1 to document its
evaluation.
The NRC finds that the TN NUHOMS[supreg] HD cask system, as
designed and when fabricated and used in accordance with the conditions
specified in its CoC, meets the requirements of 10 CFR Part 72. Thus,
use of the TN NUHOMS[supreg] HD cask system, as approved by the NRC,
will provide adequate protection of public health and safety and the
environment. With this final rule, the NRC is approving the use of the
TN NUHOMS[supreg] HD 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 January 10, 2007. Single copies of
the CoC and SER are available for public inspection and/or copying for
a fee at the NRC Public Document Room, O-1F21, 11555 Rockville Pike,
Rockville, MD.
Discussion of Amendments by Section
Section 72.214 List of Approved Spent Fuel Storage Casks
CoC No. 1030 is added to the list of approved spent fuel storage
casks.
Summary of Public Comments on the Proposed Rule
The NRC received six comment letters on the proposed rule. The
commenters included representatives from industry and members of the
public. Copies of the public comments are available for review in the
NRC's Public Document Room, O-1F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland.
Comments on the Transnuclear, Inc., NUHOMS[supreg] HD Cask System
Several of the commenters provided specific comments on the NRC
staff's preliminary SER and the TS. To the extent possible, the
comments on a particular subject are grouped together. The listing of
the Transnuclear, Inc., NUHOMS[supreg] HD 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:
Comment 1: Three commenters raised issues with using Boral[supreg]
for criticality control. One commenter pointed to documented widespread
evidence of Boral degradation; e.g., in Spain, Boral was banned from
all casks after evidence of Boral's swelling and hydrogen generation
was found in laboratory testing, and in the U.S., Boral has exhibited
swelling, blistering, and instances of major hydrogen gas generation in
dry cask fuel storage applications. Two commenters noted that NRC
issued Generic Safety Issue No. 196 to study the Boral degradation
problem. Other remarks concerning Boral are noted as follows: (1) The
problem has been occurring for 20 to 30 years; (2) Boral problems occur
on a random basis, and it is impossible to predict the product's
performance because of uncertainty in the level of porosity in the
aluminum boron carbide core of the cladded product; (3) Boral was the
material choice in past years mainly because there were no economical
alternatives; (4) The use of Boral was understandable 10 or even 5
years ago because fully dense metallic neutron absorbers were not
commercially available then, but now aluminum alloy-based neutron
absorbers with high boron content are produced by several suppliers;
(5) Boral is used today only because of its cost savings to the cask
supplier, and it is not worth putting the health and safety of workers
who load the cask at risk; (6) From a metallurgical point of view, the
most consistent performance will be demonstrated from an aluminum boron
carbide neutron absorbing product which exhibits 100 percent of
theoretical density, and only a fully dense neutron absorber will
completely eliminate the potential of swelling and hydrogen gas
generation phenomenon.
Response: The NRC is aware that canisters containing
BORALTM may generate hydrogen while the canister is
submerged in the spent fuel pool during short-term loading operations.
This was observed at the Columbia Generating Station in 2002.
BORALTM will react with the spent fuel pool water during
loading operations and generate hydrogen. The magnitude of the hydrogen
generation could depend on many factors, such as pool water chemistry,
batch-to-batch variations, time-at-temperature, etc. The hydrogen
generation does not decrease the efficacy of the material as a neutron
absorber. As is the case with most casks licensed by the NRC, the SAR
for the NUHOMS[supreg] HD describes hydrogen generation mitigating
procedures. Vendors of casks certified by NRC have recommended that the
utilities monitor for hydrogen gas during loading operations and state
that a purge be used when hydrogen gas concentration exceeds 2.4
percent prior to or during root-pass welding of the lid.
The NRC is aware that BORALTM can swell or blister under
high temperatures and hydrostatic pressures as was observed in Spain.
In October 2003, the NRC received a letter from the Empresa Nacional de
Residuos Radiactivos, S. A. (ENRESA) concerning this matter in the
Spanish cask. However, it is our understanding that the Equipos
Nucleares, S.A (ENSA) test conditions, under which blistering was
observed, were conducted at high heat-up rates and high hydrostatic
pressures, well beyond those for operating conditions for the dry cask
storage systems in the U.S. It is also our understanding that the high
heat-up rates and hydrostatic pressures did not permit the liquid to
drain prior to expanding, thereby leading to blistering. This was due
to low porosity of the BORALTM matrix structure which does
not facilitate water egress under the conditions mentioned above. The
letter from ENRESA concerning this matter in the Spanish cask and the
BORALTM blistering never stated that BORALTM has
been banned from use in Spain. It should be noted that no U.S. vendors
or utilities have reported any BORALTM blistering during
loading operations or manufacturer acceptance testing of a cask.
The staff in the Spent Fuel Storage and Transportation Division
have shared data and reports with the staff in the office of Nuclear
Regulatory Research concerning GSI-196, BORALTM degradation.
All data, reports, and letters (domestic and foreign) provided to
ascertain criticality implications of BORALTM degradation in
the context of dry cask storage of spent fuel have shown that the
efficacy
[[Page 71465]]
was not reduced in BORALTM used in dry cask storage systems.
Blistering or swelling in BORALTM has been reported to
occur under wet storage conditions in the spent fuel pools at both
domestic and foreign reactors. For example, in September 2003, FPL
Energy Seabrook, LLC, reported bulging of the BORALTM coupon
used to monitor the performance of the spent fuel pool racks. The
bulging of this coupon was due to blistering. FPL's examination and
analysis of the coupon indicated no loss in the B-10 areal density.
Neutron attenuation and radiography measurements have been
conducted on the BORALTM test coupons--both seal-welded and
vented--subjected to multiple wetting/drying cycles and varying heat-up
rates to simulate wet storage and typical cask loading conditions. In
the many test reports reviewed by the NRC staff, blistering usually
occurred in the low-porosity (low B4C content) coupons. The data
reported that the boron-10 areal density in the blistered specimens
remained unaffected. Thus, neutron attenuation efficacy was not
affected in the BORALTM. It should be noted that the
Seabrook licensee, who reported blistering in the BORALTM
coupons after about 7 years of wet storage in the spent fuel pool,
reportedly demonstrated that BORALTM suffered no loss of
effectiveness as a neutron absorber.
The NRC is aware that other neutron absorber materials are now
available to the cask vendors; however, the NRC does not recommend any
brand of material to the vendors. To date, tests have shown that the
BORALTM material still performs its intended function with
or without the blisters being present.
The NRC staff does not dispute the advantages of the near-
theoretical-density neutron absorber materials, which have become
available in recent years. However, blistering has not been shown to
affect dose to workers involved in the cask loading process.
Additionally, if hydrogen gas is detected during the loading
operations, the vendors and licensees can use mitigating procedures to
vent and purge the cask. This procedure is recommended prior to
welding; thus, worker safety can be ensured.
The NRC staff does agree that this problem of blistering and
hydrogen generation has not been reported in the absorber materials
that have a 100-percent dense matrix. However, the NRC has reviewed
evaluations by the Energy Power Research Institute (EPRI) and cask
vendors, and for the most part, the boron areal density (10B/
cm2) in the blistered specimens remained unaffected. Thus,
neutron attenuation was not affected, and there was no impact on
BORAL's effectiveness as a neutron absorber.
Comment 2: One commenter stated that the structural steel frame
used to support the DSC poses a serious risk to public health and
safety. The commenter made the following points: (1) From contact with
the air and humidity in the environment, these structurals can corrode
from the inside as well as from the outside. Particularly at coastal
sites, anything that can corrode, will corrode. Even stainless steel
develops stress corrosion cracks. (2) The upright tubes make up the
only support structure for the fuel-filled canister. They cannot be
inspected from the outside of the NUHOMS because they cannot be seen.
All primary supports must be inspected periodically, and it is a fatal
flaw to have a fuel storage canister perched about 6 feet in the air on
top of a steel frame which cannot be inspected at all. It is a
dangerous sort of design for unrestricted use around our country,
including the plants in salt air environments.
Response: Regarding Part (1), above, it is widely recognized that
corrosion is a significant concern in coastal marine environments due
to the wind borne salts deposited upon structures. Based on questions
from the staff regarding this issue, TN committed in a September 19,
2006, letter to add the following to Section 3.4.1.4 of the SAR for the
NUHOMS[supreg] HD design: ``If an independent spent fuel storage
installation site is located in a coastal salt water marine atmosphere,
then any load-bearing carbon steel DSC support structure rail
components of any associated HSM-H shall be procured with a minimum
0.20 percent copper content for corrosion resistance.'' This commitment
has also been captured in NUHOMS[supreg] HD TS 4.4.1 for the HSM-H.
Consequently, the TN design incorporates a requirement to use
atmospheric corrosion resisting steels (a.k.a., weathering steels) when
the spent fuel storage site is near a coastal marine environment.
A significant body of technical literature exists, which provides
corrosion rate data for a variety of steel alloys exposed to the
elements at coastal sites. From this data, TN recognized that
weathering steels provide ample corrosion resistance in a coastal
marine atmosphere. This corrosion resistance would assure that the
accumulated corrosion loss over a 20-year license period would be
immaterial to the structural integrity of the support steel inside the
HSM-H.
It should be noted that the data used to determine the required
corrosion allowance are for samples fully exposed to the elements. It
is known that samples that are fully shielded from the sun and rain
show a significantly lower corrosion rate than fully exposed samples.
The structural steel of the HSM-H is entirely enclosed inside a
ventilated concrete structure that totally shields the steel from
sunlight and precipitation. TN chose to employ the higher corrosion
rate data for fully exposed samples as the basis for their corrosion
allowance. This provides an added degree of conservatism to their
design.
In addition to the use of corrosion-resisting steels, TN has
specified the application of a corrosion resistant coating over the
support steel. The coating may be one of several systems. One system
consists of an inorganic zinc primer with an epoxy overcoat. This is an
industry-recognized, high performance, and long-lived industrial
coating system that is designed to withstand very severe environments.
Although the coating is specified, it is not credited in the corrosion
rate calculations that are part of the structural steel design margins.
The staff finds that the use of corrosion-resisting steel with a
calculated corrosion rate derived from a more severe exposure
environment is appropriate. Additionally, the staff finds that the use
of a coating system, and the fact that the steel is enclosed in a dry,
interior-like environment, provide additional protection against
corrosion. Thus, the staff finds that this TN design provides
reasonable assurance that the system will not experience any
significant corrosion during the 20-year license period at a coastal
spent fuel storage site.
Regarding Part (2), the commenter is correct that the canister, in
some models of the HSM, is supported in the vertical direction by a
series of columns or legs, six in total, that are made of structural
steel tubing. These columns are part of a three-dimensional welded and
bolted frame anchored vertically and horizontally to the reinforced
concrete storage module. The three pairs of columns that are each less
than 3.5 feet long support a cross beam which then provides support at
three locations for each of the two support rails. The framing design
concept is similar to that used in structural steel framing of multi-
story buildings, tankage support systems, and other applications where
a three-dimensional framing concept is appropriate. In this case, since
the frame
[[Page 71466]]
is provided with lateral supports at the location of each column to the
reinforced concrete horizontal storage module, the frame is considered
to be a braced-frame and, therefore, has limited lateral deflection
that can occur at the top of the frame. The design concept is not
considered to be unique, out-of-the-ordinary, or a dangerous design
configuration for this intended use. The design conditions that
represent the environment in which the frame must function have been
incorporated into the design criteria. In other models of the HSM, the
support rails are supported directly on the reinforced concrete storage
module by embedded anchors. The NUHOMS[supreg] HD support rails are
supported and anchored in this manner.
The commenter used the term ``primary support'' and indicated that
all primary supports must be inspected periodically. While the
NUHOMS[supreg] HD can be used at a nuclear power plant, the
certification of the dry spent fuel storage system is carried out under
10 CFR part 72 and not 10 CFR part 50. Consequently, the assertion made
by the commenter that ``all primary supports'' must be inspected
periodically may be in reference to a requirement in 10 CFR 50.55a(f),
for inservice testing requirements for nuclear power reactor facilities
for various classes of components. These 10 CFR part 50 requirements do
not apply to the passive systems that are under the jurisdiction of 10
CFR part 72. The design criteria used for the design of the
NUHOMS[supreg] HD system, to support the canisters in the horizontal
storage module, are sufficiently robust so that periodic inservice
inspections of these structural components are not deemed to be
necessary. It is correct that there is a requirement that is identified
in 10 CFR 72.122(f) related to testing and maintenance of systems and
components that are important to safety. Such systems and components
are to be designed to permit inspection. The NUHOMS[supreg] HD rail
support system could be visually inspected by remote operations using
fiber optics into the HSM-H via the vent system, or the HSM-H can be
opened, the canister extracted into the transfer cask, and the rail
supports inspected, after appropriate radiation surveys and procedures
are met. The environmental concern in Part (2) of the comment is
addressed in Part (1) response.
Comment 3: A commenter raised the following concern with respect to
flooding: Section 4.6.3 of the Generic Technical Specification states
that flood ``levels up to 50 feet and water velocity of 5 fps'' are
allowed. The commenter was concerned about the flooding condition in
which the floodwater rises to fill the inlet ducts in NUHOMS[supreg]
(all of the air inlet ducts in the NUHOMS[supreg] module lie at the
ground level). He questioned that if the floodwater rises high enough
to block off the air flow through the inlet ducts, the DSC would not
cool and concluded that without the ventilation airflow, the DSC would
overheat and may even explode from pressure buildup. It seemed to the
commenter that TN considered only the case of deep submergence flood in
the safety evaluation, which is not a risky condition because the DSC
is cooled by the flood water. The commenter further stated that low
flood level is a risky condition since the DSC is several feet above
the ground, and a flood of any height that remains below the DSC will
choke off the ventilation air and cause the DSC to overheat. The
commenter was surprised that NRC would issue ``general certification''
to a ventilated cask like this one to be used in flood plains,
considering that there are many ``nukes'' on river basins that are in
the potential flood zone. The commenter further stated that the
condition of partial height flood should be given full technical
consideration.
Response: Regarding low level floods in the situation when the
bottom vents are blocked, evaporative cooling will cool the upper
volume of the HSM and the DSC as demonstrated below. A thermal analysis
of a typical HSM and DSC with a fuel heat load of 24kW in accident
conditions demonstrates that the DSC support steel maximum temperature
is 615 [deg]F, and the DSC shell maximum temperature is 642 [deg]F.
These component temperatures would provide evaporation of the water in
the bottom of the HSM. The evaporated water would cool the DSC and the
upper volume of the HSM. The staff notes that the NUHOMS[supreg] HD
technical specification maximum heat load is 34.8 kW. Even at the
higher heat loads, staff believes that evaporative cooling will prevent
the DSC from overheating. In addition, the flood water will help cool
the submerged portion of the HSM cavity. Therefore, the staff concludes
that the DSC will not overheat, and the resulting DSC internal
pressures will not exceed the design pressure.
Comment 4: One commenter believed that TS 4.6.3 was unclear in the
statement that NRC has allowed ``seismic loads of up to 0.3 g
horizontal and up to 0.2 g vertical'' on the system. The commenter
asked for the location in the storage facility to which the g-loads
correspond, either at the C.G. of the storage system or at the pad
surface on the module's centerline, and also asked if the g-load limits
include the effect of soil-structure interaction alluded to in
Paragraph 4.2.2. Another commenter assumed that the 0.3 g horizontal
and 0.2 g vertical seismic events (per page 4-7 of Design Features in
the Certificate) are free-field accelerations at the site and stated
that they will get amplified at the pad due to soil-structure
interaction. The on-the-pad accelerations will be further magnified at
the rails due to the flexibility of the DSC support structure. Combined
with the rattling impulse from the fuel, the commenter believed that a
canister may roll off the rails.
Response: The permissible seismic loads of 0.3 g horizontal and 0.2
g vertical noted by the first commenter are the maximum values at the
top of the HSM-H or the top of the supporting basemat or pad the
NUHOMS[supreg] HD system is allowed to be subjected to. The design of
the HSM-H and the NUHOMS[supreg] HD system is based on the amplified
response spectra value of 0.37 g in the orthogonal horizontal direction
and 0.20 g in the vertical direction on the 0.3 g and 0.2 g values
respectively. The 0.30 g horizontal and 0.20 g vertical values also
reflect the resulting maximum permitted accelerations at the top of the
basemat or pad after a soil-structure interaction analysis has been
performed, if necessary, by the cask system user for the specific site
using the site-specific free field g-values. The fact left unstated is
that where a soil-structure interaction analysis must be performed by
the user, the resulting amplified response value at the center of
gravity of the loaded HSM-H must not exceed 0.37 g in the horizontal
direction and 0.20 g in the vertical direction. Based on the proposed
rule, if either of these values were exceeded, the NUHOMS[supreg] HD
system could not be used.
The interpretation of the second commenter is not what is reflected
in the TS as discussed above. The TS g-values are not generally
consistent with the free-field acceleration values at most sites.
The design conditions have included analyses of the canister in
place on the rail support system under the design lateral loads from
the seismic events, and there is no canister roll off from the rail
support system.
Comment 5: One commenter found that the DSC support structure is
not restrained against all four walls of the concrete module. A 45-ton
container resting unsecured on the rails that are not braced against
the four walls is a physically unstable arrangement. The commenter
asked if this configuration had been analyzed to ensure that failure
[[Page 71467]]
from resonance would not occur during earthquakes. The commenter stated
that he could not find any evidence of such an evaluation in the TSAR
or the NRC's SER.
Response: It is unclear to the NRC staff what the source and basis
are for these comments. The comments do not relate to the
NUHOMS[supreg] HD system. There is no document identified as the TSAR
(Topical Safety Analysis Report) associated with this docket
application (72-1030). This terminology was associated with
applications submitted in the late 1980s and early 1990s (e.g., TN-24
and TN-32 cask systems). The commenter's description of the DSC support
structure does not match that of the NUHOMS[supreg] HD system. For the
NUHOMS[supreg] HD system, the DSC support structure consists of a pair
of structural steel rails of 12-inch deep wide-flange sections that are
anchored to the reinforced concrete horizontal storage module at the
bottom flanges and connected by two struts and are, therefore,
considered braced. This configuration is provided in the SAR for the
NUHOMS[supreg] HD system. The seismic analysis determined that
amplified accelerations are based on the frequency analysis, so that
any issue of resonance has been incorporated into the analysis and then
into the design of the individual members.
Comment 6: One commenter believed that being able to remove the
container at the end of 20 years of licensed life should be an
important safety consideration. The commenter inquired and found that
no plant that has loaded a NUHOMS[supreg] in the country has ever
attempted to remove the container after a few years of storage. The
commenter wanted to know what would happen if the aging of the rails
and container's surfaces due to years of weathering were to cause the
canister to bind to the rails.
Response: The canister itself is constructed of stainless steel.
The top of the support beam has a stainless steel cover plate welded
along its entire length. This stainless steel plate forms the surface
upon which the canister rests and also serves as a sliding surface for
canister installation or removal operations. This plate may be
lubricated if desired.
Long-term experiments, where stainless steel samples were exposed
to the weather at coastal marine sites, have demonstrated that
stainless steel is highly resistant to atmospheric corrosion under
those conditions. In the case of the TN NUHOMS[supreg] HD design, the
canister and related support rails are shielded from direct exposure to
the weather (being enclosed in a ventilated enclosure). This sheltering
from the direct weather would result in little, if any, corrosion
compared to the already insignificant amounts that could occur if these
components were fully exposed to the weather. Absent corrosion, there
is no likelihood that the canister would bind to the support rails.
Because of this, and the fact that a lubricant (grease) could be
applied to the rails, if desired, the staff believes it to be highly
unlikely that any difficulty would arise during a removal operation,
even after an extended period of time.
Comment 7: A commenter asked what would happen if uneven settlement
of the pad from the heavy weight of the module were to cause the
canister to bind to the rails.
Response: Uneven settlement of the pad, commonly referred to as
differential settlement, is not expected to occur. If it were to occur,
it is highly unlikely that it would result in any differential movement
between the two supporting rails for the canister that would cause the
canister to bind to the rails. The reinforced concrete pad and the
reinforced concrete horizontal storage module represent a very stiff
structural combination, so that relative movement between the support
rails cannot be logically projected based on the structural response
from any differential settlement across the supporting base pad.
Further, the adequacy of the pad to support the horizontal storage
module, without detrimental settlements, is required under the
requirements of 10 CFR 72.212. The adequacy must be maintained under
static and dynamic loads of the storage cask system, considering
potential amplification of earthquakes through soil-structure
interaction, soil liquefaction, and other soil instabilities due to
vibratory ground motion, if these conditions exist at a site. Binding
of the canister to the support rails from settlement or differential
movement is not expected under any design condition.
Comment 8: A commenter asked what would happen if the 60 kips of
permissible extraction force to remove the container are not
sufficient. The commenter stated that this scenario is ignored in the
Technical Specification of TN's TSAR.
Response: See also response to Comment 5 regarding a document
misidentified as TN's TSAR. If settlement or differential settlement of
a limited magnitude were to develop over the years, the transport
trailer is equipped with hydraulic jacks or positioners and an
alignment system, identified as the skid positioning system that is
normally used for the alignment of the transfer cask. This same system
can be used to accommodate effects resulting from limited settlement or
differential settlement between the basemat or storage pad and the
approach slab. If a situation were to develop where the support skid
positioning system could not accommodate the magnitude of the movement,
the approach slab can be modified or other measures taken.
Comment 9: A commenter stated that the NUHOMS[supreg] HSM is much
heavier and bigger than the previous models, noting that each loaded
module weighs over 200 tons and questioned whether the ground
underneath the NUHOMS[supreg] housing would settle over the years under
the weight of the modules. The commenter also cited NRC's SER on page
3-7: ``It is assumed that an axial load of 80 kips is required for
insertion and 60 kips for extraction,'' and stated that this seems
backwards. More force will be needed to extract the canister than to
insert it (when the rail is new and greased). The commenter questioned
how a safety concern would be addressed if because of settlement and
weather effects, 60 kips is not enough to pull the canister out, and
how the NUHOMS[supreg] would be emptied of fuel if the canister binded
to the rails. The commenter believed that this would be a huge concern
to people living near the NUHOMS[supreg] sites. He further stated that
the minimum the NRC should do is to require that a demo of canister
extractions at a couple of sites loaded with NUHOMS for 10 years (or
more) be done to prove that the horizontally loaded canister can be
successfully extracted.
Response: With regard to the commenter's concern about the weight
of NUHOMS[supreg] HSM, the 80-kip insertion load, and the 60-kip
extraction load, it is noted that as stated in the SER on page 3-7,
these are the design load conditions under normal operation loading
conditions. In the off-normal operation loading condition, the
extraction force can be allowed to reach 80 kips under that design
condition. The dry cask storage system has been evaluated against the
regulatory requirements for retrievability of the spent fuel, and a
demonstration of canister extraction from the horizontal storage module
is not deemed necessary at some time after 10 years of storage. The
extraction system has been determined to be capable of functioning
during the term of the certificate.
Comment 10: A commenter stated that he could not find any
evaluation of safety for the following scenarios when the DSC is being
inserted into the HSM:
Scenario 1: The transfer cask skid has been unfastened from the
trailer and the
[[Page 71468]]
transfer cask lid has been removed making the DSC axially unrestrained,
but before the skid has been fastened to the HSM and the hydraulic ram
has been engaged to the DSC grapple ring. An earthquake during this
period, depending on its magnitude, has the potential to cause
uncontrolled DSC movement and cause a significant radiation exposure
event to the workers that could be potentially deadly to the workers.
Scenario 2: The DSC has been installed in the HSM, but the HSM lid
(a heavy circular lid that also restrains the DSC in the axial
direction) is not yet in place. An earthquake during this period could
cause a major radiation exposure event that could be potentially deadly
to the workers.
Response: Scenario 1: For the described scenario, the position of
the transfer cask for the NUHOMS[supreg] HD system, before the lid is
removed, is on the transfer trailer, with the cask within several feet
of the open HSM-H cavity, after the centerlines of the HSM-H and the
cask have been verified to be approximately coincident. The lid of the
cask is then removed. The transfer trailer is then backed to within a
few inches of the face of the HSM-H, the trailer brakes are set, and
the tractor is disconnected from the trailer and moved away. The
transfer trailer vertical jacks are positioned to locate the vertical
position of the cask in its approximate insertion orientation. The skid
tie-down bracket fasteners are removed, and the position of the cask is
corrected, as needed for alignment, using the hydraulic skid
positioning system. Then, the optical survey equipment and reference
marks are used for adjusting the final alignment. The skid positioning
system is then used for that final alignment, and the canister is
inserted into the HSM-H access opening docking collar. The transfer
cask is then secured to the HSM-H using the cask restraints.
A large seismic event, during the period of time from when the
transfer cask lid is removed and is several feet from the HSM-H, and
before the transfer cask is anchored to the HSM-H with a sufficiently
large horizontal axial component, could overcome the frictional
resistance that keeps the canister inside the transfer cask. This would
not, however, be an uncontrolled DSC movement, since the DSC inside the
transfer cask has only an approximately \1/4\-inch radial gap, which
controls the movement to essentially longitudinal/axial movement with
the maximum lateral position of the DSC changing by approximately \1/
64\-inch for each inch of longitudinal/axial movement. The
longitudinal/axial movement is limited by the distance of several feet
between the transfer cask opening and the face of the HSM-H. A
longitudinal/axial movement of 3 to 5 feet of the DSC from the transfer
cask opening would not constitute an uncontrolled DSC movement, since
that longitudinal/axial movement is limited by the face of the HSM-H
module.
The possibility of the hypothesized scenario is considered to be
much less than what is considered significant for design accident
conditions arising from handling and storage of spent nuclear fuel. The
seismic event, to produce the hypothesized movement, must have a large
enough component of acceleration in the longitudinal/axial direction of
the positioned transfer cask that can be at any point on the compass,
and the event must occur within a time period of 2 to 4 hours. On an
annual basis, this would occur only three to five times per year for a
given facility. If such a remote accidental event were ever to occur,
plant operations personnel would respond by placing temporary shielding
with equipment over any exposed portion of the DSC.
Scenario 2: The operations' procedures identify that upon
disengagement of the transfer cask from the HSM-H, the canister's axial
seismic restraint is installed. This is a design feature that uses a
structural steel embedment in the reinforced concrete of the HSM-H as
the anchor point for the retainer device. The commenter's assumption
that the HSM-H lid or door is the axial retainer for the canister is
incorrect.
Comment 11: One commenter stated that the DSC is pushed into the
HSM module using a simple hydraulic ram that has no redundant load
handling features. A simple failure such as loss of hydraulic pressure
during the pushing operation would leave the DSC in a partially
inserted configuration. The commenter believed that a single failure
proof ram system should be required or TN should demonstrate that a ram
failure halfway through the DSC pushing process can be dealt with using
credible recovery measures. The commenter did not believe that NRC has
ever considered this issue or that TN has ever been asked to provide an
answer.
Response: The functioning of the ram operating system is not
considered to be a system that is safety related since the canister is
confined and shielded during the period of ram operations. A failure in
the location, as hypothesized by the commenter, presents an operational
problem, but no significant issues are created. The corrective action
would be to repair the operating system of the ram. NRC has considered
this scenario, and the NRC agrees with the safety classification of the
ram assembly that it is ``Not Important To Safety'' as identified in
Table 2-5 of the applicant's SAR.
Comment 12: A commenter stated that the DSC, according to NRC's
SER, can survive the drop from 80 inches height, but was concerned
about how a dropped DSC would be lifted from the pad. The DSC seems to
have no lifting or handling attachments except for the grapple, which
is useable only to engage the ram for a horizontal push.
Response: The commenter is correct in that there are no lifting or
handling attachments other than the grapple ring for a loaded canister.
The DSC is placed into the transfer cask within the fuel pool and then
is loaded with spent fuel. Then, after removal from the fuel pool and
preparation for transfer, the closed cask is moved on the transfer
trailer in a horizontal orientation to a location outside the fuel
handling building. The transfer trailer and cask with the DSC closed
inside are moved to the pad area. The DSC is not lifted out of the
transfer cask, but is pushed out of the cylindrical transfer cask
directly into the HSM-H in a horizontal position, with the transfer
cask coupled to the HSM-H, creating a connecting tunnel space
completely enclosing the DSC. This operating procedure makes the
possibility of a dropped DSC on the pad extremely unlikely and an
accident that is beyond the design basis accident. If a beyond design
accident condition were to arise where a loaded and unshielded DSC had
to be lifted, the first step would be to provide temporary shielding
and probably execute a remote lift in the horizontal position with a
device brought in for special use. Such special procedures can be
developed for an accident condition response. It should be noted that
the 80-inch side drop is for the DSC inside the transfer cask.
Comment 13: A commenter stated that NRC should require a stiff
foundation underneath the NUHOMS[supreg] to support the weight of the
NUHOMS[supreg]. At present, the commenter sees nothing in the proposed
certificate that requires a strong support foundation to be built. He
believes this to be a serious oversight.
Response: The weight of the NUHOMS HD[supreg] system, as installed
in-place, including the HSM-H, the DSC, and the spent fuel, is to be
supported by the ISFSI basemat or pad. That structure is identified in
accordance with 10 CFR 72.3 as ``Not Important to Safety.'' The basemat
or pad is designed, constructed,
[[Page 71469]]
maintained, and tested as a commercial grade item designed to be in
compliance with 10 CFR 72.212(b)(2). This regulation requires that the
user of the NUHOMS[supreg] HD cask system must evaluate and establish
that the following criteria are met:
(1) The cask storage pads and areas have been designed to
adequately support the static and dynamic loads of the storage casks,
considering potential amplification of earthquakes through soil-
structure interaction and soil liquefaction potential or other soil
instability due to vibratory ground motion.
(2) For the HSM-H loaded with a filled -32 PTH DSC, the weight is
approximately 207.5 tons that is distributed over the pad area, which,
as a minimum, is approximately 200 square feet.
(3) The static load bearing pressure on the supporting soil
material would normally be approximately 2075 pounds per square foot, a
common value used for residential and commercial building foundations
on fine-grained soils.
(4) The loading on the foundation is not considered to be
structurally significant or unusually high.
Comment 14: A commenter expressed the following concerns pertaining
to storing fuel horizontally in a hot state:
(1) After searching the public filings by TN on this docket and
Docket No. 72-1004, the commenter could not find a single evaluation of
the consequences of storing fuel horizontally over long periods of
time. In discussions between Westinghouse and a utility, the conclusion
that they reached was that ``additional analyses and evaluation will be
needed to determine whether it is permissible to store Westinghouse's
fuel horizontally.''
(2) A lot of fuel is already in NUHOMS[supreg] at many sites. What
is happening to all of the fuel stored outside of the fuel supplier's
(Westinghouse's) specifications is unknown because the condition cannot
be examined.
Response: In response to (1), after searching the TN filings, one
document was found in which Westinghouse stated that ``* * * additional
analyses and evaluation may be needed * * *.'' The NRC staff
independently performed a generic analysis of spent fuel stored
horizontally under the design service condition and for the service
life of the NUHOMS[supreg] storage system. This analysis looked at the
structural capability of the spent fuel materials to perform in the
horizontal position without degrading spent fuel performance.
There are two sources of stress in the fuel cladding, when in the
horizontal orientation, that could result in creep. These are internal
pressurization of the fuel rod and gravity. Two possible sources of
deformation of the cladding, bending and creep, are possible under the
horizontal position. The bending stress and the hoop stress are both
considerably less than the yield stress under internal pressure and a
horizontal position. The bending deflection, at the center of the span
between the grid spacers, due to the downward gravitational load of the
fuel, is approximately 3 millimeters. No changes occur in the stresses
or radial growth as a result of storage in the horizontal position. The
creep deformation is self limiting under both stresses due to the
decreasing temperature of the fuel with time. If the initial maximum
temperature is kept below 400 [deg]C, as recommended by Interim Staff
Guidance (ISG)-11, then the creep deformation under the maximum
allowable pressurization is less than 1 percent over a 20-year storage
period. No cladding failure is expected at this strain level. The
additional downward load, due to the gravitational force from the
unsupported, approximately 300 grams of fuel between the grid spacer
supports, increases the longitudinal stress by no more than 1 percent
of the material strength and results in a minuscule increase of the
hoop stress. Therefore, no more additional creep is expected in the
horizontal orientation than in the vertical orientation.
In response to (2), the cask vendors specify the range of
parameters for the fuel to be stored in the CoC. The worst case fuel is
analyzed as in paragraph (1), above. The fuel is evaluated when it is
removed from the reactor to determine if it falls in the specified
envelope. If it is in this envelope, no adverse fuel performance is
expected.
Comment 15: A commenter stated that, in the future, the fuel that
will be stored will have burned longer in the reactor. The commenter
believed that the NRC should perform a careful safety evaluation before
permitting even more fuel, particularly well burned fuel, to be stored
horizontally. The commenter cited NRC's SER on page 4-6 that reads:
``The NUHOMS HD DSC only undergoes a one-time temperature drop during
backfilling of the DSC with helium gas. Because this is a one-time
event, the DSC does not undergo any thermal cycling.'' The commenter
stated that the SER evidently assumes that the fuel will never be
unloaded, unpackaged, and reloaded after it has been vacuum dried and
backfilled. If that is the underlying basis of the SER, the commenter
believes that the certificate should be restricted to only once-through
loading such that there is no likelihood of thermal cycling of the
fuel.
Response: The staff has performed a safety evaluation and analyzed
the effects of these parameters on the storage of fuel as provided in
the guidance contained in ISG-11, Rev. 3. Higher burnup fuels will have
the following characteristics:
(1) A higher cladding stress caused by a higher internal pressure
due to an increased fission gas release from the pellets;
(2) A higher hydrogen content in the cladding resulting in a
decrease in mechanical properties; and
(3) A higher heat generation rate.
As long as the fuel burnup is below the approved in-reactor burnup
limit (currently 62.5 GWd/MTU) and is maintained in a nonoxidizing
atmosphere below 400 [deg]C, there are no active degradation mechanisms
that would cause cladding breaches to occur under normal storage
conditions. In addition, the structural review must include mechanical
properties of the cladding at the limit of the approved burnup to
determine the behavior of the fuel under off-normal and accident
conditions.
The staff has evaluated the issue of thermal cycling on the
behavior of irradiated fuel. Two issues of concern were thermal shock
during reflood, if wet unloading occurs, and hydride reorientation.
Reflood analysis is required in every SAR to evaluate the ability of
the cladding to tolerate the thermal shock to the cladding due to the
rapid submergence of the hot fuel in the cool pool water. For the
NUHOMS[supreg] HD unloading operation, the maximum fuel cladding
temperature during cask reflood is calculated to be significantly less
than the vacuum drying condition because of the presence of water
vapor. Consequently, during cask reflood, a lower temperature rise is
expected when compared with that for the cask vacuum drying operations.
Hydride reorientation, which might degrade the mechanical
properties of the cladding, occurs when hydrogen goes into solution and
is subsequently precipitated under stress during cooling. A number of
studies indicate that thermal cycling may contribute to the phenomena
of reorientation. To limit the occurrence of hydride reorientation in
the cladding during storage, drying, etc., ISG-11, Rev. 3, limits the
number of thermal cycles that the fuel can experience to 10 or less.
Thermal cycling is only a concern if thermal
[[Page 71470]]
cycling takes place early in the storage period when the fuel is
relatively hot.
Under normal storage conditions, there are no mechanisms to degrade
the fuel to the point where a loaded cask would have to be opened
prematurely. At later times in the storage period, when unloading and
repackaging are expected to occur, the temperatures will be at a lower
maximum temperature due to the reduced decay heat, and as a result,
less hydrogen (the solubility decreases exponentially with temperature)
will be able to go into solution during these operations. In addition,
the maximum stress in the rods will be less than at the initial vacuum
drying, due to the lower temperature during unloading and repackaging.
As a result, hydride reorientation, and consequently thermal cycling,
is not of concern during unloading later in the storage period.
Comment 16: A commenter stated that ``NRC's SER says that--The
application performed dynamic impact analysis using LS-DYNA 3D on a
cask-pad-soil finite element model * * *.'' The commenter believed that
this was not true and noted that the FSAR shows that the applicant used
a cookbook approach, developed by EPRI in the time when LS-DYNA was not
widely used, which is considered to be unconservative by most experts.
The commenter further stated that, according to the experts he
consulted, a true LS-DYNA analysis would have shown much greater g-
loads under an 80-inch drop. Therefore, the SAR analysis on which the
NRC has relied is inadequate and unconservative.
Response: The analytical method used by the applicant referred to
by the commenter was performed as described in the NRC's SER using
NUREG/CR-6608, dated February 1998, using LS-DYNA 3D. This is a
commercial finite element dynamic analysis software package capable of
three-dimensional representations. The DYNA 3D software package used in
the development of the analysis procedure described in NUREG/CR-6608 by
Lawrence Livermore National Laboratory is the comparable software
package that has been used in the national laboratories. The analytical
approach used in NUREG/CR-6608 is considered by NRC as an acceptable
method of evaluation for low-velocity impacts such as a dropped cask.
It is recognized that, in this approach, the transfer cask internals
that include the canister, the fuel basket, and the spent fuel are
modeled only by their mass and their mass distribution.
Comment 17: A commenter believed that the tornado missile analysis
in Chapter 11 of the NUHOMS[supreg] FSAR does not consider the damaging
scenario of missile impact. The commenter stated that the analysis
assumes impact over the concrete walls. The most dangerous impact would
occur if the missile were to hit the fasteners that keep the door of
the HSM in place. If the fastener fails from the missile impact, then
the door will come loose and the canister will be uncovered, exposing
people nearby to radiation. The commenter did not see any evaluation of
this scenario in TNs FSAR or NRC's SER.
Response: The scenario proposed by the commenter, while not
specifically identified, is encompassed by and bounded by the scenarios
specifically discussed in the referenced documents. First, it is
necessary to have an accurate understanding of the physical
configuration of the door of the HSM-H and the opening for the door on
the front wall of the HSM-H base assembly. The door thickness is a
total of 2.53 feet made up of 0.65 feet of steel, and the remainder is
made of concrete. Approximately 97 percent of the total thickness of
the door is inside the plane of the outside face of the HSM-H, filling
the recessed hole. The door is supported within the hole on two radial
bearing pads that support the door on the 1.875-foot thickness of
concrete of the 2.53-foot door thickness. The door is not supported in
the vertical direction by the fasteners that the commenter addressed.
The failure of one of those fasteners, as a result of a local missile
impact, would not dislodge the door from the HSM-H base unit, and the
door's radiation shielding capability would remain. Since the relevant
missiles used to evaluate local missile damage effects all have
physical dimensions and resulting damage zone dimensions much less than
the spacing of the subject fasteners, multiple fastener loss is not
likely. The fasteners' minimum spacing is approximately 5 feet, whereas
the missiles considered relevant have maximum dimensions of
approximately 1.5 feet. Even with multiple fastener failures, the thick
door assembly will most likely remain in the deeply recessed opening
after a local missile strike on the door's steel exterior, since the
door assembly would have to move axially outward nearly 2 feet in order
for the HSM-H to be rendered to a condition with an open door.
Comment 18: A commenter expressed concern with the way the canister
is stored. The commenter stated that it seems that the canister is
lying on a couple of rails, and it is held in place by gravity and
nothing else (no straps, no frame, no structurals to restrain it).
Response: The commenter is correct that the canister is supported
by two structural support rails. These are configured to create a
cradle for the canister. The two rails of the cradle are each oriented
at 30 degrees off the vertical centerline through the DSC, as it is in
the stored horizontal position. With the 60-degree angle between the
rail supports, a simple calculation demonstrates that a side load,
through the center of gravity of the DSC, would have to exceed
approximately 0.55 grams to disturb the at-rest position of the stored
cask. This value, for lateral load, exceeds the control limits that are
placed on this system, regarding the sites where the system could be
used. That results in a design transverse load of 0.41 grams on the
DSC. In the longitudinal direction, the DSC is restrained from movement
on the rail support system by the axial retainer system that restrains
DSC movement, with respect to the HSM-H.
Comment 19: A commenter understood that the fuel is stored in the
canister in a non-fixed manner and that during an earthquake, the fuel
would move in the canister. The commenter inferred from reading the SAR
that most of the canister's weight is in the fuel. He stated that if
most of the weight is free to move about in the canister, then there is
a risk of the canister rolling over and falling down during an
earthquake.
Response: The maximum values for comparing weight distribution for
a loaded DSC are that 46.6 percent of the total weight of a loaded DSC
is the weight of the spent fuel and the other 53.4 percent is the
weight of the canister, the internal basket, and other hardware of the
cask. The internal fuel basket is a cellular structure that provides a
storage position 8.7 inches by 8.7 inches in cross-section for each of
the 32 spent fuel assemblies that are stored. The orthogonal grid of
the assemblage of these 32 cells is circumscribed by a circle created
by metallic basket rails that transition from the grid configuration to
a circle concentric with the inside surface of the canister. The radial
space from the fuel basket and basket rails to the inside face of the
canister is one-eighth of an inch. This configuration does not allow
gross freedom of movement of the stored fuel, but only provides
sufficient space to allow for loading and unloading of the spent fuel
and for the thermal growth that is expected. Consequently, there is
minimal lateral displacement of the spent fuel that can occur inside
the canister.
Comment 20: One commenter stated that he did not find a time
history analysis in Appendix 3.9.9.10.2 of the
[[Page 71471]]
SAR to determine if canister bouncing or rolling might occur. He also
stated that it did not appear that the effect of soil-structure
interaction was mentioned.
Response: As described in Section 3.9.9.10.2 of Appendix 3.9.9 of
the SAR, the seismic design basis for the HSM-H and the stored spent
fuel in the canister is based on the maximum peak accelerations at the
top of the basemat, or pad structure, not exceeding 0.3 grams in the
horizontal direction or 0.20 grams in the vertical direction. For the
sites where soil-structure interaction analysis is considered
important, the user of the NUHOMS[reg] HD system will have
to determine that these values are not exceeded. Additionally, as
indicated in the TS, Section 4.0, Design Features, amplified seismic
response spectra from such an analysis would be produced. The HSM-H
system, with the stored canister, is based on a limit of 0.37 grams in
both transverse and longitudinal directions and 0.20 grams in the
vertical direction, at the center of gravity of the HSM-H, with respect
to the amplified response spectra. Within these limits of
accelerations, there will be no uncontrolled motion of the canister
that would result in a safety issue.
Summary of Final Revisions
The proposed TS and SER have been revised in response to Comment 2
to capture and document TN's commitment to add the following to Section
3.4.1.4 of the SAR for the NUHOMS[reg] HD design: ``If an
independent spent fuel storage installation site is located in a
coastal salt water marine atmosphere, then any load-bearing carbon
steel DSC support structure rail components of any associated HSM-H
shall be procured with a minimum 0.20 percent copper content for
corrosion resistance.''
Voluntary Consensus Standards
The National Technology Transfer and Advancement 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
NUHOMS[reg] HD 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.
Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' approved by the Commission on June 30, 1997,
and published in the Federal Register on September 3, 1997 (62 FR
46517), this rule is classified as Compatibility Category ``NRC.''
Compatibility is not required for Category ``NRC'' regulations. The NRC
program elements in this category are those that relate directly to
areas of regulation reserved to the NRC by the Atomic Energy Act of
1954, as amended (AEA), or the provisions of Title 10 of the Code of
Federal Regulations. Although an Agreement State may not adopt program
elements reserved to NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with the
particular State's administrative procedure laws but does not confer
regulatory authority on the State.
Finding of No Significant Environmental Impact: Availability
Under the National Environmental Policy Act of 1969, as amended,
and the NRC regulations in Subpart A of 10 CFR Part 51, the NRC has
determined that this rule, if adopted, would not be 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 EA and finding of no significant impact on which this
determination is based are available for inspection at the NRC Public
Document Room, 11555 Rockville Pike, Rockville, MD. Single copies of
the EA and finding of no significant impact are available from Jayne M.
McCausland, Office of Federal and State Materials and Environmental
Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (301) 415-6219, e-mail [email protected].
Paperwork Reduction Act Statement
This final rule does not contain a new or amended information
collection requirement subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the
Office of Management and Budget, Approval Number 3150-0132.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
Regulatory Analysis
On July 18, 1990 (55 FR 29181), the Commission issued an amendment
to 10 CFR Part 72. The amendment provided for the storage of spent
nuclear fuel in cask systems with designs approved by the NRC under a
general license. Any nuclear power reactor licensee can use cask
systems with designs approved by the NRC to store spent nuclear fuel if
it notifies the NRC in advance, the spent fuel is stored under the
conditions specified in the cask's CoC, and the conditions of the
general license are met. In that rule, four spent fuel storage casks
were approved for use at reactor sites and were listed in 10 CFR
72.214. That rule envisioned that storage casks certified in the future
could be routinely added to the listing in 10 CFR 72.214 through the
rulemaking process. Procedures and criteria for obtaining NRC approval
of new spent fuel storage cask designs were provided in 10 CFR Part 72,
Subpart L.
The alternative to this action is to withhold approval of this new
design and issue a site-specific license to each utility that proposes
to use the casks. This alternative would cost both the NRC and
utilities more time and money for each site-specific license.
Conducting site-specific reviews would ignore the procedures and
criteria currently in place for the addition of new cask designs that
can be used under a general license, and would be in conflict with NWPA
direction to the Commission to approve technologies for the use of
spent fuel storage at the sites of civilian nuclear power reactors
without, to the maximum extent practicable, the need for additional
site reviews. This alternative also would tend to exclude new vendors
from the business market without cause and would arbitrarily limit the
choice of cask designs available to power reactor licensees. This final
rulemaking will eliminate the above problems and is consistent with
previous Commission actions. Further, the rule will have no adverse
effect on public health and safety.
The benefit of this rule to nuclear power reactor licensees is to
make available a greater choice of spent fuel storage cask designs that
can be used under a general license. The new cask
[[Page 71472]]
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.
Regulatory Flexibility Certification
Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the
NRC certifies that this rule will not, if issued, have a significant
economic impact on a substantial number of small entities. This final
rule affects only the licensing and operation of nuclear power plants,
independent spent fuel storage facilities, and TN. 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 final rule because this amendment
does not involve any provisions that would impose backfits as defined.
Therefore, a backfit analysis is not required.
Congressional Review Act
Under the Congressional Review 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.
List of Subjects in 10 CFR Part 72
Administrative practice and procedure, Criminal penalties, Manpower
training programs, Nuclear materials, Occupational safety and health,
Penalties, Radiation protection, Reporting and recordkeeping
requirements, Security measures, Spent fuel, Whistleblowing.
0
For the reasons set out in the preamble and under the authority of the
Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of
1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting the
following amendments to 10 CFR part 72.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE
0
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. 102-
486, sec. 7902, 106 Stat. 3123 (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); sec. 1704, 112 Stat. 2750 (44 U.S.C.
3504 note); sec. 651(e), Pub. L. 109-58, 119 Stat. 806-10 (42 U.S.C.
2014, 2021, 2021b, 2111).
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, 2224 (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).
0
2. In Sec. 72.214, Certificate of Compliance 1030 is added to read as
follows:
Sec. 72.214 List of approved spent fuel storage casks.
* * * * *
Certificate Number: 1030.
Initial Certificate Effective Date: January 10, 2007.
SAR Submitted by: Transnuclear, Inc.
SAR Title: Final Safety Analysis Report for the
NUHOMS[reg] HD Horizontal Modular Storage System Irradiated
Nuclear Fuel.
Docket Number: 72-1030.
Certificate Expiration Date: January 11, 2027.
Model Number: NUHOMS[reg] HD-32PTH.
Dated at Rockville, Maryland, this 22nd day of November, 2006.
For the Nuclear Regulatory Commission.
William F. Kane,
Acting Executive Director for Operations.
[FR Doc. E6-20962 Filed 12-8-06; 8:45 am]
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