[Federal Register Volume 69, Number 4 (Wednesday, January 7, 2004)]
[Rules and Regulations]
[Pages 849-858]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-313]
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Rules and Regulations
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Federal Register / Vol. 69 , No. 4 / Wednesday, January 7, 2004 /
Rules and Regulations
[[Page 849]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 72
RIN 3150-AH36
List of Approved Spent Fuel Storage Casks: Standardized
NUHOMS[reg]-24P, -52B, -61BT, -24PHB, and -32PT Revision
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
-----------------------------------------------------------------------
SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to revise the Transnuclear, Inc. (TN) Standardized
NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask system listing
within the ``List of Approved Spent Fuel Storage Casks'' to include
Amendment No. 5 to Certificate of Compliance (CoC) Number 1004.
Amendment No. 5 will add another dry shielded canister (DSC),
designated NUHOMS[reg]-32PT DSC, to the authorized contents
of the Standardized NUHOMS[reg]-24P, -52B, -61BT, and -24PHB
cask system. This canister is designed to accommodate 32 pressurized
water reactor (PWR) assemblies with or without Burnable Poison Rod
Assemblies. It is designed for use with the existing
NUHOMS[reg] Horizontal Storage Module and
NUHOMS[reg] Transfer Cask under a general license.
EFFECTIVE DATE: This final rule is effective on January 7, 2004.
FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Nuclear
Material Safety and Safeguards, 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, 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. The NRC subsequently issued a final rule on December 22, 1994
(59 FR 65920), that approved the Standardized NUHOMS[reg]-
24P and -52B cask design and added it to the list of NRC-approved cask
designs in Sec. 72.214 as Certificate of Compliance Number (CoC No.)
1004. Amendments No. 3 and 6 added the -61BT DSC and the -24PHB DSC,
respectively, to the system.
Discussion
On June 29, 2001, the certificate holder (TN) submitted an
application to the NRC to amend CoC No. 1004 to add another dry
shielded canister, designated NUHOMS[reg]-32PT DSC, to the
authorized contents of the Standardized NUHOMS[reg]-24P, -
52B, -61BT, and -24PHB cask system. This canister is designed to
accommodate 32 PWR assemblies with or without Burnable Poison Rod
Assemblies. It is designed for use with the existing
NUHOMS[reg] Horizontal Storage Module and
NUHOMS[reg] Transfer Cask. No other changes to the
Standardized NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask
system were requested in this application. The NRC staff performed a
detailed safety evaluation of the proposed CoC amendment request and
found that an acceptable safety margin is maintained. In addition, the
NRC staff has determined that there is still reasonable assurance that
public health and safety and the environment will be adequately
protected.
This rule revises the Standardized NUHOMS[reg]-24P, -
52B, -61BT, and -24PHB cask system listing in Sec. 72.214 by adding
Amendment No. 5 to CoC No. 1004. The particular Technical
Specifications (TS) which are changed are identified in the NRC staff's
Safety Evaluation Report (SER) for Amendment No. 5.
The NRC published a direct final rule (68 FR 49683; August 19,
2003) and the companion proposed rule (68 FR 49726) in the Federal
Register to revise the TN Standardized NUHOMS[reg]-24P, -
52B, -61BT, and -24PHB cask system listing in 10 CFR 72.214 to include
Amendment 5 to the CoC. The comment period ended on September 18, 2003.
One comment letter was 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 October 30, 2003; 68 FR 61734.
The NRC finds that the amended TN Standardized
NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask system, as
designed and when fabricated and used in accordance with the conditions
specified in its CoC, meets the requirements of Part 72. Thus, use of
the amended TN Standardized NUHOMS[reg]-24P, -52B, -61BT,
and -24PHB 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 Standardized
NUHOMS[reg]-24P, -52B, -61BT, -24PHB, and -32PT 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 7, 2004. Single copies of the CoC and SER are available for
public inspection and/or copying for a fee at the NRC Public Document
Room, 11555 Rockville Pike, Rockville, MD. Copies of the public
comments are
[[Page 850]]
available for review in the NRC Public Document Room, 11555 Rockville
Pike, Rockville, MD.
Summary of Public Comments on the Proposed Rule
The NRC received one comment letter on the proposed rule. A copy of
the comment letter is available for review in the NRC Public Document
Room. The NRC's responses to the issues raised by the commenter follow.
As stated in the proposed rule (68 FR 49726; August 19, 2003), the NRC
considered this amendment to be a noncontroversial and routine action.
Therefore, the NRC published a direct final rule (68 FR 49683; August
19, 2003) concurrent with the proposed rule (68 FR 49683; August 19,
2003). The NRC indicated that if it received a ``significant adverse
comment'' on the proposed rule, the NRC would publish a document
withdrawing the direct final rule and subsequently publish a final rule
that addressed comments made on the proposed rule. The NRC believes
some of the issues raised by the commenter were ``significant adverse
comments.'' Therefore, the NRC published a notice withdrawing the
direct final rule (68 FR 61734; October 30, 2003). This subsequent
final rule addresses the issues raised by the commenter that were
within the scope of the proposed rule.
Comments on Amendment 5 to the TN Standardized NUHOMS[reg]-
24P, -52B, -61BT, -24PHB, and -32PT Cask System
The commenter provided specific comments on the Technical
Specifications, the SER, and the Final Safety Analysis Report (FSAR).
None of these documents were changed as a result of public comments. A
review of the comments and the NRC's responses follows:
Comment 1: The commenter stated that TS 1.1.1 set the limits of
0.17g vertical and 0.25g horizontal on seismic accelerations and
identified these limits as site-specific parameters. The commenter also
stated that the SER was equally ambiguous in paragraph 3.1.2.1.7. The
commenter recommended that the TS be corrected to state unequivocally
that 0.25g and 0.17g are, respectively, the maximum permitted values of
the peak horizontal and vertical accelerations at the NUHOMS/
Independent Fuel Storage Installation (ISFSI) pad interface.
To support this recommendation, the commenter referred to an
inspection of the FSAR which revealed that 0.25g and 0.17g are applied
as peak horizontal and vertical ground accelerations on the NUHOMS
system. The commenter stated that it is common knowledge in
geomechanics that the free field accelerations at the site can be
magnified considerably on the pad due to soil-structure interaction
effects. The commenter added that TN's analysis of NUHOMS assumes that
0.25g and 0.17g horizontal and vertical accelerations are applied on
the horizontal storage module (HSM) basemat; thus, these are the
limiting values of on-the-pad accelerations, not ``site parameters'' as
noted in the TS.
Response: Page A-1 of the Technical Specifications states the
following. ``* * * site specific parameters and analyses, identified in
the SER, that will need verification by the system user, are, as a
minimum, as follows: * * *''. Item 3, in that listing, states: ``The
horizontal and vertical seismic acceleration levels of 0.25g and 0.17g,
respectively.''
The commenter indicates that the SER is ambiguous in addressing
when the site-specific seismic parameters are to be taken as design
values. In quoting Section 3.1.2.1.7 of the SER, the commenter did not
include the second sentence of the SER paragraph. That second sentence
of the paragraph states that: ``The location of these accelerations is
taken at the top of the concrete pad/basemat of the HSM.'' What the
actual values are is a function of the site which includes the ground
accelerations and soil structure interaction effects.
No additional clarification is necessary in the Technical
Specifications.
Comment 2: The commenter quoted a portion of Sec. 72.130 which
mandates that the ISFSI must be designed for decommissioning,
particularly it must be designed ``to facilitate the removal of
radioactive wastes * * *''.
The commenter stated that, based on the information presented in
the FSARs and NRC's SER, one cannot conclude with reasonable confidence
that the loaded -32PT dry shielded canisters will be able to be removed
by the hydraulic ram after the NUHOMS modules have been on the storage
pad for their licensed life (20 years).
To support this view, the commenter presented two main technical
reasons for pessimism with regard to the removal of the loaded DSCs
after 20 years of storage; namely, potential for long-term settlement
of the pad and weathering (corrosion) of the DSC/rail interface under
extended exposure (20 years) to the elements.
With respect to long-term settlement, the commenter noted that TS
1.2.9 stipulates that the transfer ``cask must be aligned with respect
to the horizontal storage module (HSM) so that the longitudinal
centerline of the DSC in the transfer cask is within +/- \1/8\ inch of
its true position when the cask is docked with the HSM front access
opening.'' Further, this requirement, imposed to enable the DSC to be
moved horizontally, is tedious but doable during initial loading.
However, calculations performed for typical storage pads loaded with
heavy casks show that the long-term differential settlement from soil
creep can be several inches over 20 years. The commenter stated that
NUHOMS's FSAR makes no special demands on the soil strength to limit
long-term settlement of the pad. The commenter further stated that
there are no specific strength limits applied on the NUHOMS pad either
which, along with the absence of a mandated hard subgrade, would likely
lead to several inches of differential settlement of the pad over 20
years of storage, and the user's ability to maintain the alignment
specified in TS 1.2.9 will be lost. The commenter claimed that the DSC
will be in an irremovable state, in direct violation of Sec. 72.130.
Response: As stated in Section 1.3.1.2 of the FSAR, ``The HSMs are
constructed on a load bearing foundation which consists of a reinforced
concrete basemat on compacted engineered fill.'' The general licensee
is responsible for the design and construction of the HSM load bearing
foundations. If a properly designed and constructed foundation system
is completed for the basemat, several inches of hypothesized
differential settlement should not develop. If differential settlement
of a limited magnitude were to develop, the transport trailer is
equipped with hydraulic jacks/positioners and an alignment system
identified as the support 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 differential settlement
between the basemat and the approach slab. If a situation were to
develop where the support skid positioning system could not accommodate
the differential settlement, the approach slab can be modified or other
measures can be taken. See the following response on corrosion and
environment.
Comment 3: The commenter stated that, under the general CoC
authority, the NUHOMS system can be installed at any site in the U.S.,
including coastal sites and marine environments. The potential for
surface corrosion, including pitting the DSC and HSM rail surfaces
under the ambient
[[Page 851]]
environmental conditions and its effect on the removability of the DSC,
has not been considered in NUHOMS's August 2000 FSAR for the
Standardized NUHOMS System or NRC's SER. This is in violation of Sec.
72.236(m).
Response: The potential for surface corrosion (i.e., pitting
corrosion) under the ambient environmental condition and its effect on
the retrievability of the DSC has been considered by the selection of
corrosion resistant materials. The DSC shell structure is fabricated
from ASME SA 240, Type 304 stainless steel. Type 304 stainless steel
has excellent corrosion resistance in a wide range of atmospheric
environments and many corrosive media. The corrosion resistance is
provided by the 18 percent minimum chromium content. The material used
as the sliding surface of the DSC is a high-hardness stainless steel
plate (Nitronic 60). The Nitronic 60 has similar corrosion resistance
as Type 304 stainless steel. This plate is mounted on the HSM rails as
shown in Drawing No. NUH-03-6016-SAR contained in FSAR, Appendix E. The
surface of the Nitronic 60 is lubricated to minimize friction.
Additionally, both the DSC and the DSC support structure are housed
inside of the HSM reinforced concrete structure which protects it from
direct exposure to the weather. Therefore, staff concludes that none of
the DSC and HSM rail materials are expected to degrade or react with
each other. Further, staff concludes that the NUHOMS design considers
the effects of environmental conditions and retrievability and meets
the requirements of 10 CFR 72.236(m).
Comment 4: The commenter claimed that the maximum allowable
hydraulic push and pull forces specified in the FSAR are not equal. The
commenter stated that the push force is 80 kilopounds (kips); the
permitted pull force is only 60 kips. The commenter further stated that
it is during the removal of the DSC, when the DSC must be dragged over
the corroded HSM rails, that the risk of failure to remove the canister
lies. Yet, the allowable pull for the DSC extraction condition is 25
percent less than the available push force during initial insertion.
Further, the coefficient of friction during DSC push assumed in the
FSAR to be 0.2 is unrealistically low for weathered sliding surfaces.
Response: The commenter is in error in stating that the maximum
allowed extraction force for the removal of the DSC from the HSM is 60
kips. It is 60 kips under normal loading and 80 kips for off-normal
loadings which is equal to the off-normal insertion loading (FSAR Table
3.2-1 and SER Section 3.1.2.1.2). The permitted loads for insertion and
extraction are the same, but there is a difference in the permitted
stress allowables. As stated on page 3.1-6 of the FSAR, the hydraulic
ram used to exert the insertion or extraction force is sized assuming a
coefficient of friction of 1.0.
Comment 5: The commenter noted that, in the FSAR, there was no
stress analysis of the DSC bottom cover plate that is being pulled by
the hydraulic ram against friction, in conjunction with the internal
pressure present in the canister. The commenter stated that internal
pressure and the hydraulic ram pull force act in concert to maximize
the stress level in the cover plate and its junction with the DSC
shell. The commenter believed that neglect of analysis of this scenario
leaves the structural adequacy of the bottom outer lid open to
question.
Response: Table 8.2-24 of Revision 5 of the FSAR shows that an
analysis of the DSC was done for accident unloading conditions that
assumed the full force of the ram (80 kips) and an internal pressure of
60 psi. The analysis showed that this situation was bounded by the 75g
side drop load at Service Level D. Tables M.2-15 and M.3.7-10 show the
same situation for the NUHOMS[reg]-32PT system with the new
internal design pressure of 105 psi. Sections 3.1.2.2 and 3.3.2 of the
SER address these tables.
Comment 6: The commenter discussed the process of inserting a DSC
in the HSM and noted that this requires careful alignment of large
fabricated components in open air and that the time duration for such
activities can be long. The commenter stated that the NRC imposes
seismic requirements on canister transfer outside of Part 50 structures
even in vertical operations (see NAC-UMS or HI-STORM FSAR, for
example). Yet, for the more tedious horizontal insertion process in
NUHOMS, there is no treatment of a concurrent seismic event or even
tornado-borne missiles during DSC transfer operations. The commenter
stated that this violates a provision in Sec. 72.122(b)(2)(1) which
requires that structures, systems, and components must be able to
withstand the effects of natural phenomena such as earthquakes.
Response: The FSAR amendment in Section M.3.7.3.6 states that the
effects of a seismic event occurring when a loaded DSC is resting
inside the transfer cask (TC) have been analyzed. Reference is made to
the fact that the conditions for the 32PT are bounded by the conditions
used for the 24P analyses described in the original FSAR. The
referenced section, Section 8.2.3.2(D), indicates that all conditions
existing during loading or transport operations are enveloped by two
loading cases that are described in the FSAR, one of which envelops and
applies to this condition. TN has performed a stability analysis that
shows there is a safety factor of at least 2.0 against overturning the
cask/trailer assembly during a seismic event in this bounding case.
During the cask transfer operation, the cask/trailer unit is attached
to the HSM by the cask restraint devices that are anchored into the
front of the HSM and are attached to the trunnions of the TC as shown
in FSAR Figure 4.2-13. These restraints are designed for accident
conditions and envelop seismic loads. The TC and the HSM are designed
for tornado missiles as described in Section 3.2.1 of the FSAR,
Revision 5. The NUHOMS system is designed to withstand seismic
conditions as well as those produced by tornado-borne missiles.
Comment 7: The commenter stated that the 32PT DSC is the heaviest
canister proposed for use thus far in the HSM. The commenter noted that
NUHOMS's FSAR asserts that the DSC support structure is braced,
presumably to incorporate seismic resistance. A review of the sketches
provided in the FSAR showed no bracing. The commenter provided marked
up pages from NUHOMS's FSAR for the Standardized NUHOMS System to
indicate the missing braces. The commenter stated that, without the
braces, the DSC support structure in the HSM is weak against axial or
lateral overturning moments, especially the increased g-loads that will
accompany the heavier 32PT DSC.
Response: The commenter is correct in stating that the 32PT DSC is
the heaviest canister to date proposed for use in the NUHOMS Storage
System. As stated by Transnuclear, Inc., on page 1.1-2 of the proposed
FSAR revision for Amendment 5, the HSM has been qualified for a DSC
weight of 102,000 pounds that envelops the 101,380 pounds for the 32PT
in the storage configuration. As stated on page M.1-1 of Amendment 5,
there is no change to the HSM required for the 32PT component for the
NUHOMS system.
As shown in the FSAR, Revision 5, the DSC is supported on two rails
that are supported by a structural steel frame in the cavity of the
HSM. The frame structure is anchored to the reinforced concrete floor
slab, the side walls, and the front wall. Figures 4.2-6 and 4.2-7
illustrate the longitudinal and transverse sections of the HSM with the
DSC support structure inside. Figures 4.2-8 and 4.2-9 provide
additional
[[Page 852]]
details of the DSC support structure. These drawings show that the
structural steel frame is a braced frame in both the transverse and
longitudinal directions. A braced frame does not have to be
additionally braced with diagonal bracing. Each planar frame or bent of
the three dimensional structural frame is braced or restrained from
transverse lateral movement, in the plane of the frame or bent, at the
top by a structural steel channel section that acts as a strut or tie
to the reinforced concrete wall of the HSM. In the longitudinal
direction, the entire three-dimensional structural frame is braced
through the rail extension plate and base plate that are anchored to
reinforced concrete of the throat of the opening of the HSM. Figure
8.1-20 of the FSAR, Revision 5, presents the DSC structural support
analytical model showing that this three dimensional (space) frame is
considered to be a braced frame. It should be noted that there is
another NUHOMS storage system, the Advanced NUHOMS Storage System, that
has different features and was developed for higher seismic application
areas.
The DSC support structure inside the HSM is adequate for the
specified input values to show conformance with Sec. 72.236.
Comment 8: The commenter stated that the consideration of the
tornado-borne missile in the FSAR for the Standardized NUHOMS System is
oblivious to the real vulnerability of the HSM. The commenter further
stated that the entire 3-foot thick top roof is held by a mere 4
anchors about 1\1/2\ inches in diameter, and the concrete-filled front
door (over 7,000 pounds in weight) is not even held by bolts (rather by
3 straps). The commenter asserted that the FSAR for the Standardized
NUHOMS System provides no analysis of the integrity of these weak
locations in the HSM under natural environmental phenomena loads.
Response: Although the roof is held to the base by eight 1\1/4\-
inch steel bolts and the roof attachment angle assembly which would
resist a significant lateral force, these are not the design features
provided to resist roof lateral loads and other accident loads. There
is a 4-inch key or ledge of concrete which sits in the base that is
designed to resist lateral loads of the roof. Downward vertical loads
are resisted by shear and bending of the roof with the downward loads
carried out at the periphery in bearing to the base unit walls. The key
detail can be seen in drawing NUH-03-6015, Rev. 5, Sheet 1 of 2.
Contrary to the assertion of the commenter, the HSM door is held on
by bolts, not straps. Analyses of the HSM and the HSM door are
presented in FSAR Sections 8.2.2 and 8.2.3 for tornado and seismic
conditions. These analyses show that the entire HSM has been qualified
for its design basis tornado and wind loads.
The HSM structure is adequately designed to resist the tornado and
seismic loading conditions as required by Sec. 72.236.
Comment 9: The commenter stated that how the structural features
will resist a larger impact such as a plane should be a matter of
concern to the agency in the after-9/11 world.
Response: The Commission believes that the best approach to dealing
with threats from aircraft is through strengthening airport and airline
security measures. Consequently, we continue to work closely with the
appropriate Federal agencies to enhance aviation security and thereby
the security of nuclear power plants and other NRC-licensed facilities.
Shortly after the September 11, 2001, attacks, the NRC, working with
representatives of the Federal Aviation Administration (FAA) and
Department of Defense (DOD), determined that a Notice To Airman
(NOTAM), issued by the FAA, was the appropriate vehicle to protect the
airspace above sensitive sites. This NOTAM strongly urged pilots to not
circle or loiter over the following sites: Nuclear/Electrical power
plants, power distribution stations, dams, reservoirs, refineries, or
military installations, or expect to be interviewed by law enforcement
personnel. Further, the NRC issued orders imposing additional physical
protection measures for independent spent fuel storage installations
using dry storage.
The NRC is conducting a comprehensive evaluation that includes
consideration of potential consequences of terrorist attacks using
various explosives or other terrorist techniques on dry storage casks.
As part of this evaluation, the agency is looking at the structural
integrity of dry storage cask systems and will consider the need for
additional design requirements to enhance licensee security and public
safety.
Comment 10: The commenter noted that, according to the FSARs, the -
32PT DSC has purportedly been analyzed for a drop from 80 inches onto
an unyielding surface with the added assumption that the transfer cask
is rigid. This event is postulated to account for a potential drop of
the loaded DSC in the transfer cask during its handling on the basemat.
The calculations to compute the g-load, however, use an antiquated
method that was determined to be unconservative by the NRC in the mid-
1990s.
The commenter stated that, in 1997, the NRC established the
acceptable method for reliably and conservatively predicting the g-load
in a paper titled ``NRC Staff Technical Approach for Spent Fuel Storage
Cask Drop and Tipover Accident Analysis.'' The commenter believed that
the method relied on in the FSAR is unconservative and that a much
higher value than 75g's will develop if the NUHOMS[reg]-32PT
DSC undergoes a free fall of 80 inches on a rigid surface without the
benefits of an impact limiter.
Response: The commenter's reference to ``the NRC paper sets down
the acceptable method for reliably and conservatively predicting the g-
load'' has apparently been misinterpreted to mean that this is the only
acceptable method for calculating the impact loads. The referenced
paper, in its title, uses the words ``technical approach'' that is
intended to imply that the methodology therein is acceptable to the
NRC, but that does not mean that it is the only acceptable methodology
that could be utilized. Analysis of drops from heights of up to 80
inches were chosen because they were representative of the worst case
drops that might be found at an ISFSI, or along the transfer route.
There was no assumption that the impacted surface was essentially
unyielding or rigid. The methodology adopted by TN considered the
stiffness of the impacted surface. As noted on page 3-19 of the NRC
staff Safety Evaluation Report dated December 1994 for the Standardized
NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel,
the NRC staff independently completed calculations to verify that the
design deceleration values were conservative.
Comment 11: The commenter stated that TS 1.2.13 permits lift
heights of up to 80 inches in cold conditions based on nil ductility
transition (NDT) temperature considerations of the transfer cask's
materials. The commenter further stated that the underlying documents
[Safety Analysis Report (SAR) or SER] do not address the top and bottom
shield plugs that are very thick (over 6 inches) and made of a steel
that is low-temperature incompetent (A-36). The commenter believed that
at -20 F, the A-36 plugs will suffer extensive fracture under a 75-g
impact load, perhaps even pulverization.
Response: The shield plugs are fabricated from American Society for
Testing and Materials (ASTM) A36 steel, a commonly used steel for
structural applications. ASTM A36 was
[[Page 853]]
selected because of its high strength and metallurgical stability.
However, if this material should experience temperatures below -
20[deg]F, its ductility (or fracture toughness) and its ability to be
used for structural applications may be insufficient and, thereby, lead
to potential fracture of the material. To address this issue, the user
is constrained by the TS to ensure that fracture (pulverization, as
characterized by the comment) does not occur. TS 1.2.13 prescribes the
following limits: (1) No lifts or handling of the TC/DSC at any height
are permissible at DSC temperatures below -20[deg]F inside the spent
fuel pool building; (2) the maximum lift height of the TC/DSC shall be
80 inches if the basket temperature is below 0[deg]F, but higher than -
20[deg]F inside the spent fuel pool building; and (3) the maximum lift
height and handling height for all transfer operations outside of the
spent fuel pool building shall be 80 inches, and the basket temperature
may not be lower than 0[deg]F. Therefore, staff has concluded that the
ASTM A36 carbon steel has sufficient fracture toughness (material
properties) to remain functional, when operated under the limitations
set forth in the TS.
Comment 12: The commenter stated that he was greatly concerned
about the clear absence of critical structural welds in the fuel basket
in the -32PT DSC. The commenter manually circled areas in the drawing
details released to the public that show absence of welds in the fuel
basket at critical load transfer locations under a horizontal drop
condition.
Response: The commenter is correct in that welds are not shown in
the drawing that was marked up and attached to the comments. However,
this drawing is not intended to show the weld location and types
because this information is contained in proprietary drawing NUH-32PT-
1004, Rev 0, Sheet 2 of 2. All required critical locations are welded
together. Section M.1.2.1 of Amendment 5 on page M.1-4 of the
nonproprietary version provides a verbal description of the basket
assembly. The following statement is made in that section: ``The basket
structure consists of a grid assembly of welded stainless steel plates
or tubes that make up a grid of 32 fuel compartments.''
Comment 13: The commenter stated that TNW's stress analysis of the
basket appears to have a serious error, perhaps an erroneous assumption
in the finite element model. The commenter stated that critical stress
analyses figures were deleted from the nonproprietary copy and he could
not offer further help.
Response: The commenter gives no information regarding any specific
reference to the related NUHOMS documents and gives no indication as to
the origin of the stress such as thermal, seismic, or some other
loading condition with respect to the comment. It is assumed that the
commenter believes that there are no welds between the various cells of
the basket assembly and that the finite element analysis was conducted
on a model that represented a continuum or structural integrity across
the interfaces among the cells. With regard to the comment that
``critical stress analyses figures are deleted from the non-proprietary
copy,'' if the commenter is referring to Figures M.3.6-1 through M.3.6-
4, those figures in the proprietary version of Amendment 5 do not
identify stresses. Instead, these figures provide the modeling details
of the finite elements used in the analyses. The NRC staff has not
identified any significant erroneous assumptions in the finite element
models utilized.
Comment 14: The commenter quoted from NUREG-1536, Chapter 11, V.1,
that ``an event may be analyzed for regulatory purposes even though no
credible cause can be identified. Such events should be clearly
identified as nonmechanistic.''
The commenter stated that NRC's regulatory practice has been to
require a nonmechanistic tipover analysis of casks in long-term
storage. According to the NUHOMS FSAR for the NUHOMS Standardized
System, each horizontal storage module is freestanding. The height (15
feet) to width radio (9.7 feet wide) of the horizontal storage module
is comparable to vertical ventilated systems (that tend to be about 18
feet high by 11 feet diameter) where NRC has always demanded a
nonmechanistic tipover analysis. The commenter asked the question why
the special dispensation for NUHOMS, with its top heavy structure (a 3-
foot thick top roof held in place by slim anchors).
Response: The commenter states that the height to width ratio (15
feet to 9.7 feet) is comparable to vertical ventilated systems. This
does not take into account the two side shield walls attached to a
single HSM. This would make the limiting dimension 9.7 feet +4 feet =
13.7 feet. Therefore, the height to width ratio is not comparable to
vertical ventilated systems (\15/13\.7= 1.09 is considerably less than
\18/11\=1.6). The tipover analyses, however, are carried out on a
single HSM unit.
The tipover of a single HSM was considered under specific loading
conditions, namely the tornado effects as well as the seismic effects.
The discussion on these analyses is included in the FSAR, Revision 5,
in Sections 8.2.2.2.A.(i) and 8.2.3.2.B.(iii). The factors of safety
are 1.38 and 1.24, respectively, against tipover. In the case of the
tipover or liftoff of the 32PT DSC from the DSC support structure rails
inside the HSM from a seismic event, the factor of safety is 1.20 as
identified in Section M.3.7.3.1.2 of FSAR Amendment 5.
The nonmechanistic tipover analysis of a cask system is performed
to ascertain that a cask that is handled, lifted, and moved will not
suffer a loss of function under a tipover event. In other words, the
specific cause or mechanism of that event such as a failed lifting
apparatus or human error in the attachment of the lifting device is not
identified as a credible cause. In the case of the NUHOMS design
concept, the cask storage system that includes the DSC inside the HSM
is never handled, lifted, or moved. The nonmechanistic events for this
system are those considered when the DSC is in the TC as indicated in
Figure 8.2-3 of the FSAR, Revision 5.
The relevant considerations have been made for the nonmechanistic
tipover events.
Comment 15: The neutron absorber panels in 32PT DSC appear not to
be ``fixed'' as required by Sec. 72.124(b). Response: The
neutron absorber plates are fixed in place. The plates are fixed using
screws as shown on Drawing No. NUH-32PT-1003-SAR Sheet 2, Rev. 2.
Comment 16: The commenter stated that the required B-10 loading in
the neutron absorber panels is minuscule, merely 0.007 gm/sq.cm., less
than even 52BT for BWR fuel (which is 0.016 gm/sq.cm.), and a small
fraction of that used in other casks (such as NAC-STC).
Response: The B-10 neutron absorber panels are not solely relied
upon for criticality control. The minimum B-10 content of the absorber
panels, along with the poison rod assemblies (PRAs) and the borated
water, ensures that the 32PT canister will remain subcritical during
loading and unloading operations.
Comment 17: The commenter stated that the reliance for reactivity
control seems to be based on the so-called Poison Rod Assemblies
(PRAs). These PRAs, vital to criticality control, are little more than
stainless steel tubes filled with ``B4C pellets'' (see PSER,
Section 3.1.4.2). There are no requirements imposed on the size and
integrity of the welds that will join the closure plugs to these thin-
walled tubes (as little as 0.018-inch thick per Figure M.1.6-2 in the
SAR).
[[Page 854]]
Response: The NUHOMS SAR includes commitments to perform
dimensional measurements and visual examination for both the neutron
absorber plates and PRAs in Section M.9. The visual examination (per
ASME or American Welding Society (AWS)) will identify any weld
discontinuities (such as cracks, porosity, blisters, or foreign
inclusions) on the end cap of the PRA.
Comment 18: The commenter stated that the so-called nonstructural
PRA closure welds, without any regulatory requirements on their NDE,
are the sole barrier against leaching out Boron Carbide from the PRAs.
The commenter stated that a total reliance on the micro-seal welds to
hold B4C in place to preserve criticality safety appeared to
be incredulous, considering that the PRAs will be subject to thermal
stresses during fuel loading and be quite hot in long-term storage. The
commenter added that there is no requirement to purge air and moisture
from the PRA tubes before seal welding its contents. This means
entrained air and moisture will be locked in every PRA in the stored
fuel.
Response: The temperatures that the PRAs are subjected to are not
hot enough to generate a significant pressure from the relative
humidity inside of the tube. The NRC staff does not anticipate a loss
of the seal welded end cap due to internal pressure build-up. Further,
because there is no electrolyte present in the PRAs and since boron
carbide is insoluble and inert, there should be no corrosion or
chemical interaction between the stainless steel and the boron carbide
pellets. It should be noted that if there were any defective weld
discontinuities on the end cap of a PRA while the cask is inside the
pool, there would be practically no leaching of boron from the
defective weld on the closure plug. Boron carbide is virtually
insoluble in water. See ASTM Standard Specification for Nuclear-Grade
Boron Carbide Powders (C 750-03). Additionally, as stated in Section
M.1.2.2.3.1 of the SAR, the PRAs are only necessary during loading and
unloading operations. The NRC staff has concluded that the criticality
safety is not compromised during loading and unloading operations
because there is no mechanism that will cause leaching out of the boron
from the PRAs.
Comment 19: The commenter stated that the 32PT DCS is in violation
of Sec. 72.236(h) which requires that the ``spent fuel storage cask
must be compatible with wet and dry spent fuel loading and unloading
facilities.'' To support this view, the commenter stated that the
storage slots in the 32PT DSC are 8.7-inch x 8.7-inch (nominal) opening
(see PSER). The FSAR for the Standardized NUHOMS System specifies ``the
minimum open dimension or each fuel compartment is 8.60 inches x 8.60
inches.'' The commenter stated that, having worked for PWR Nuclear
Steam Safety System (NSSS) suppliers for many years, no Westinghouse or
B&W plant has fuel storage racks with 8.6-inch (min) or 8.7-inch (nom.)
opening dimension. Irradiated fuel tends to bend, bow, and twist in the
reactor; for this reason, PWR reactor suppliers require large storage
cell openings. The 32PT DSC, with 8.6-inch (min.) opening, would be an
engineered stuck fuel event.
Response: The dimensions of the fuel compartment openings are
adequate to accommodate the fuel assemblies including the Westinghouse
and Babcock & Wilcox types. There is no degradation mechanism that
would cause an assembly already in a cask to bow, except for an
accident. Therefore, if an assembly is able to be loaded into a cask,
it should be able to be unloaded.
Comment 20: In a related matter to Comment 19, above, the commenter
expressed deep reservation about the loose aluminum blocks (visible in
FSAR Amendment 5) that are assumed to be snugly fitting. The commenter
stated that the 32PT DSC will be made from a thinner shell (\1/2\-inch)
(to hold a heavier basket) than prior NUHOMS DSCs (\5/8\-inch thick
shell). This means that the shell in the 32PT DSC will ovalize more
from its dead weight and from full-length butt welds. The commenter
further stated that snugly fitted aluminum blocks may appear acceptable
on paper, but in real hardware are impossible to manufacture, and told
NRC to recall that the lack of fabricability of VSC-24 baskets
(cracking of steel plates at the toe of the bend) caused the industry
an untold amount of grief.
Response: The commenter referenced Figure M.3.7.3, but it is
assumed to have been intended to mean Figure M.3.7-3, ``0-Degree Side
Drop Stress Intensity, 32PT Basket With Aluminum Transition Rails
(Support Rails at +/-18.5-Degrees),'' in making the comment that ``the
loose aluminum blocks * * * that are assumed to be snugly fitting.''
Figure M.3.7-3 is a schematic representation of the transverse cross-
section of a DSC that illustrates the stress levels in the materials
but does not show details of the configuration. Section M.1.5 of the
FSAR contains the drawings that illustrate a configuration of the
aluminum transition rail sections with respect to the stainless steel
plates they are attached to. Drawing NUH-32PT-1006NP-SAR, Sheet 1 of 1,
illustrates that there are attachment connectors between the aluminum
transition rails, the rail plates, and the basket assembly. The
connectors are stainless steel studs welded to the outside of the
basket assembly. The studs and the basket assembly are shown on Drawing
NUH-32PT-1003NP-SAR, Sheets 1 and 2 of 2, as Detail 2. The connection
configuration also provides for differential thermal movements.
Therefore, the aluminum transition rails are not loose and do not rely
on a snug fit for their position.
The commenter indicates that because of the reduced thickness of
the cylindrical shell of the 32PT DSC and the full length butt welds,
there will be increased ovalization of the DSC shell under dead loads.
The implication of the comment is apparently that this increased
ovalization could potentially cause the assumed snugly fitting
transition rails to become even looser. The DSC was analyzed for dead
loads using the ANSYS finite element models shown in Figures 8.1-14a
and 8.1-14b in the FSAR. One loading condition considers the fuel
loaded DSC in a horizontal position with the dead loads. The fuel-
loaded portions of the basket assembly bear on transition rails that
then bear on the inner shell of the DSC. Figures M.3.6-3 and M.3.6-4
illustrate the model used with the shell and the basket for a typical
support condition of the loaded DSC. Such a model is then analyzed to
determine the primary membrane and membrane plus bending stresses as
well as for the primary plus secondary stresses. Deformed shapes are
also obtained from such analyses.
Figure M.3.6-12 illustrates the stress intensities in the DSC shell
and the aluminum transition rails under the dead load of the spent fuel
inside the basket assembly as supported in an HSM. This is considered a
normal loading condition, and the appropriate stress allowables are
17,500 psi for primary membrane stress, 26,300 psi for membrane plus
bending stresses, and 54,300 psi for primary plus secondary stresses.
This particular loading condition produces very low stress intensities
in the shell material that are 2,650 psi, 6,000 psi, and 7,000 psi,
respectively, as identified by stress type above, as shown in Table
M.3.6-2. With the worst case thermal effects that can be present under
these normal conditions, combined with the dead load, the stress for
the primary plus secondary stresses increases to 44,550 psi, still less
than the 54,300 psi allowable. Figures M.3.6-12 and M.3.6-13 illustrate
the results of the analyses.
[[Page 855]]
With these stress levels that show that the material remains in the
elastic behavior range, deformations will remain elastic. Specific
comparisons of elastic deformations between a 0.625-inch shell
thickness and a 0.500-inch shell thickness under dead load conditions
have not been made by the NRC. It is correct that there would be more
ovalization with a thinner shell; however, the incremental change has
no apparent impact on the capability of the DSC to perform its intended
storage function cradled on the pair of support rails within the HSM.
The effects of longitudinal butt welds in the cylindrical shell on the
tendency of the shell to become oval have been considered and have been
determined to be of no safety consequence.
The commenter states that snugly fitting aluminum blocks that are
the transition rails will be impossible to manufacture. This comment is
assumed to have been related to the difficulty that could arise if the
positions of the aluminum transition rails were to rely on a ``snug
fit.'' As noted above, the transition rails are positioned controlled
via studs attached to the basket assembly. The NRC has no information
that would indicate that the solid aluminum transition rails cannot be
manufactured by current machining practices to the necessary dimensions
and tolerances.
Comment 21: The commenter stated that he was surprised to learn
from the supplier's FSAR that a loaded 32PT DSC canister will have no
provision to be lifted on its own and must be lifted by the TC. The
commenter also stated that if the DSC were to be separated from the TC
under an accident event, there would be no means to lift and handle the
canister. The commenter considered the lack of ability to separately
handle a loaded canister to be a severe weakness that violates the
notion of retrievability under Sec. 72.122(l).
Response: Retrievability, with regard to certificates of compliance
for spent fuel storage casks, is addressed in Sec. 72.236(m), which
states: ``To the extent practicable in the design of the storage casks,
consideration should be given to compatibility with removal of the
stored spent fuel from the reactor site, transportation, and ultimate
disposition by the Department of Energy.'' This refers to retrieval of
the fuel assemblies from the canister. This design meets this
requirement. The canister is able to be handled and placed into the
transfer cask before loading of assemblies. The canister is then
handled as one piece with the transfer cask until it is placed within
the storage module. There are no postulated accidents when the canister
is inadvertently separated from the transfer cask.
Comment 22: The commenter referred to Section 1.2.24 of the TS
which states: ``* * * for the NUHOMS-32PT system, the fuel cladding
limits are based on Interim Staff Guidance (ISG)-11, Revision 2.'' The
commenter disagreed and quoted from page 2 of ISG-11, Rev. 2:
``Accordingly, the materials reviewer should coordinate with the
thermal reviewer to assure that the maximum calculated temperatures for
normal conditions of storage, and for short-term operations including
cask drying and backfilling, do not exceed 400[deg]C (752[deg]F).''
The commenter noted that in direct violation of the above
requirement, the Amendment 5 FSAR states in Section 4.1: ``During
short-term conditions, the fuel temperature limit is 570[deg]C.''
The commenter further stated that calculated temperature values in
Table M4.2 indicate that the ISG-11, Rev. 2, limit is exceeded by wide
margins under short-term normal conditions.
Response: The comment is based on an older version of Amendment 5
to FSAR CoC 1004 (Rev. 0, June 2001). The correct version of the SAR
corresponds to the following reference: Transnuclear West, Amendment
No. 5 to NUHOMS CoC 1004, Addition of 32PT DSC to Standardized NUHOMS
System, Rev. 4, January 2003, which complies with ISG-11, Rev. 2.
Comment 23: The commenter stated that use of durable materials that
are proven for their intended function must be a basic plank of dry
storage system design, and a mandated fact under Sec. 72.122(a), (b),
and (c). One objection raised by the commenter to the materials being
proposed for the 32PT DSC was that the shield plugs at the two ends of
the DSC are made from one of the cheapest carbon steels available (A-
36). The commenter noted that the lower plug (along with air) is
permanently sandwiched between the two stainless plates. This plug will
expand and contract under heat, as will the entrained air in the space,
constantly stressing the welds that confine the plug. Thermal
differential expansion between carbon and stainless steel will further
increase stresses in those same welds. The commenter asked why the
plugs could not be made of machined stainless steel, which would
eliminate material incompatibility, remove most entrained air, and
remove long-term concerns.
Response: The material used for the shield plug is appropriate
based on the following: First, the shield plugs are fabricated from
ASTM A-36 steel, a commonly used steel for structural applications.
Second, brittle fracture of the carbon steel is not expected because
the ductile-to-brittle transition temperature is below the expected
operating temperatures. Third, the shield plugs are also plated with
electroless nickel in response to NRC Bulletin 96-04 to ensure that a
chemical reaction does not occur. This coating is not expected to react
with the spent fuel pool water to produce unsafe levels of flammable
gas. Fourth, there are small radial clearances provided between the
carbon steel bottom shield plug and the stainless steel DSC shell.
Fifth, Table M.3.3-1, ASME Code Materials Data for SA-240 Type
Stainless Steel, and Table M.3.3-2, Materials Data for ASTM A-36 Steel,
show that the thermal coefficient of expansion is of the same order of
magnitude between 100 to 800[deg]F. Sixth, the residence time of a plug
in water is limited to cask loading operations and then vacuum dried.
Therefore, any degradation would be minimal. The NRC staff concludes
that these material properties are acceptable and appropriate for the
expected load conditions (e.g., hot or cold temperature, wet or dry
conditions) during the license period and in accordance with regulatory
requirements.
Comment 24: Related to Comment 23, above, another objection raised
by the commenter with respect to the materials being proposed for the
32PT DSC was the neutron absorber. The commenter was not able to locate
any specificity on the brands of neutron absorbers permitted by the
CoC. The commenter stated that neutron absorbers use aluminum, which is
a most reactive material, and stated that NRC has been wise in
controlling the specific make of neutron absorbers that are permitted
to be used and felt that this caution is well placed, considering the
1996 hydrogen ignition event in SNC's product. Referring to a section
in the PSER that stated that purging of the canister during lid welding
is not required, the commenter disagreed and stated that it is unsafe
to make purging elective if aluminum-based neutron absorber coated
carbon steels are present in the canister. He referred to the lesson
learned from the Columbia Generating Station experience.
The commenter recommended that the CoC specify the acceptable
neutron absorbers to ensure compliance with the above-cited regulation
and not let a CoC holder make the choice of neutron absorber
unilaterally.
Response: Technical Specification Table 1-1h imposes requirements
on
[[Page 856]]
neutron absorbers materials for the boron.
The NRC staff is aware of a slight potential for chemical or
galvanic reaction between the aluminum and stainless steel in contact
with borated water spent fuel pools. This reaction may produce small
amounts of hydrogen, during loading and unloading operations. Further,
the NRC staff is aware of hydrogen being generated from prepassivated
Boral. This reaction may also produce small amounts of hydrogen, during
loading and unloading operations. As stated in M.3.4 of the SAR, small
amounts of hydrogen could be produced during loading and unloading
operations. The applicant's analysis showed that a hydrogen
concentration of 2.39 percent can be generated. However, the NRC staff
recognizes that this amount of hydrogen is below the ignition limit of
4 percent. However, to address the potential hazards associated with
hydrogen gas, the applicant employs mitigation actions contained in the
generic procedures of SAR Sections M.8.1.3 and M.3.4. These sections
state that if hydrogen gas is detected at concentrations above 2.4
percent in air at anytime before or during welding operations, the
hydrogen gas will be removed by purging the suspect regions with an
inert gas. The NRC staff concluded during this review that the guidance
in the generic procedures is adequate to prevent formation of any
hydrogen gas that may be generated during welding operations. Hence,
the potential reaction of the aluminum with the spent fuel pool water
will be minimized and not impact the efficacy of the poison material.
Neutron absorber materials such as Metamic and BorAlyn have
undergone qualification testing. The qualification testing included an
evaluation for hydrogen generation. The qualification test program was
reviewed and approved by the NRC for these two materials.
Finally, any neutron absorbers used inside of an approved cask
design must have been shown through qualification testing to be
effective and durable during the license period. The tests and data are
usually submitted along with the license application and are subject to
review and questioning by the NRC staff. After the absorber material
has been approved at a particular level of B-10 credit by the NRC, the
SER discusses the technical basis for approval. It should be noted that
the licensee may potentially use any neutron absorber material at that
approved level of B-10 credit in its cask provided it meets the
requirements in Sec. 72.48. Therefore, there is no reason to reference
the manufacturer/brand name of the neutron absorber in the CoC.
Comment 25: Referring to paragraph M.4.6.3 of the FSAR for
Amendment 5, the commenter concluded that a fire event in the vicinity
of the HSM was ruled out. The commenter stated that this inference is
also supported by the text matter in the FSAR for the Standardized
NUHOMS[reg] System. The commenter believed that the FSAR
statements ruling out fire around the HSM are erroneous because the
hydraulic fluid in the ram and the fuel in the heavy-haul trailer are
credible sources of fire for a previously loaded HSM located in the
vicinity of the HSM being loaded.
The commenter stated that the a priori exclusion of fire analysis
at the HSM is inconsistent with NRC's previous certification reviews of
other ventilation systems and that it is also unsafe.
Response: The fire event associated with the loading operations and
storage within the HSM (including fires in the vicinity of the HSM) is
bounded by the analyzed transfer cask fire event. The transfer cask
fire analysis was based on very conservative assumptions. Other site-
specific fires have to be addressed by the system user planning to use
the NUHOMS[reg]-32PT storage cask, as part of the Sec.
72.212 evaluations.
Comment 26: The commenter referred to Section M.3.1.2.1 of the FSAR
for Amendment 5 which states that the inner bottom cover plate-to-shell
joint is subjected to volumetric and liquid penetrant examination as
required by Subsection NB of Section III of the ASME Code. The
commenter stated that examination of this weld cannot be radiographed
or ultrasonically tested by virtue of its geometry.
Response: The examination of the full penetration weld corner joint
used on the inner bottom cover plate-to-shell weld is specifically
addressed in paragraph NB-5231(c) of the ASME Boiler and Pressure
Vessel Code Section III, Subsection NB. The geometry of the weld in
question is in accordance with Figure NB-4243-1(f). As stated by TN,
the weld geometry of Figure NB-4243-1(f) is able to be successfully
examined ultrasonically in conformance with the ASME Code requirements.
Comment 27: The commenter states that Section 4.8 of the SER
accepts sudden quenching of irradiated fuel at 678[deg]F in water
during reflooding operation. The commenter stated that quenching would
cause a sudden cooling of the fuel, and the 117[deg]F temperature limit
would undoubtedly be exceeded, a restriction imposed by ISG-11, Rev. 2,
presumably to protect semibrittle irradiated fuel from thermal shock.
The commenter urged the NRC to reconsider this unnecessary regulatory
leniency.
Response: Section 4.8 of the SER states that the maximum cladding
temperature reached during vacuum drying after approximately 33 hours
is 678[deg]F (358.88[deg]C). This is below the maximum limit of
752[deg]F (400[deg]C) per ISG-11. The maximum temperature difference
for the fuel cladding during drying and backfilling operations is
100[deg]F (55.55[deg]C). This meets the thermal cycling criteria
specified by ISG-11, which states that the temperature differences
greater than 117[deg]F (65[deg]C) should not be permitted. The maximum
fuel cladding temperature during cask reflood operations will be
significantly less than the vacuum drying condition because of the
presence of water and/or steam in the DSC cavity.
Comment 28: Referencing Section 3.7 in the Amendment 5 FSAR, the
commenter stated that the consideration of flood in the FSAR is merely
to treat it as a source of hydrostatic load. The commenter believed
that a low elevation flood that submerges the bottom duct is far more
dangerous. He stated that a partially submerged HSM, heated by the DCS
through radiation and convection and chilled by the rising floodwaters,
will cause severe thermal stresses in its reinforced concrete
structure. The commenter further stated that because the HSM's walls
are both structural members and biological shield, a thru-thickness
crack from large thermal strains induced by a short-duration flash
flood will be unacceptable for public health and safety. The commenter
stated that there is no consideration of this scenario in the
supporting licensing material provided by TNW and added that it calls
for a careful analysis.
Response: As stated in the FSAR, Revision 5, Section 8.2.4,
recovery from flooding events has been addressed, and the case of
completely blocked inlet and outlet vents has been addressed in Section
M.4.6.1 of proposed Amendment 5. The blocked vent condition is assumed
to be superimposed concurrently with the extreme off-normal ambient
thermal condition of 117[deg]F with insolation. Under these
conservative design conditions, there is a 40-hour period at minimum,
that must elapse before there are thermal conditions arising that would
approach design limits. The Technical Specifications in Attachment A of
the CoC on page A-57 address the fact that there is daily (every 24
hours) visual surveillance required of the exterior of the vents as
well as a close-up inspection performed to see that
[[Page 857]]
there are no vent blockages. If blockage is found, action must be taken
to clear the vent(s) within the 40-hour time period because, as shown
in Figure 8.2-16, the concrete temperature limit of 350[deg]F will be
reached in the concrete roof structure of the HSM.
Additionally, in the situation when only the bottom vent is
blocked, the water would begin to evaporate from the heat load. This
would provide evaporative cooling to the DSC and the upper volume of
the HSM. Such a situation would be bounded by the analysis of blocked
circulation vents with ambient temperatures at their extremes (-
40[deg]F and 117[deg]F) as noted above. In these situations, the
maximum temperature gradients experienced by the HSM are 102[deg]F and
99[deg]F, respectively, as shown in Table 8.1-17 of the FSAR.
Comment 29: The commenter stated he was surprised and disappointed
that the CoC uses a product designation name like ``-32PT,'' where the
``T'' stands for transportable; and uses the words, ``* * * and T is to
designate that the DSC is intended for transportation in a 10 CFR 71
approved package,'' when this CoC pertains only to storage. The
commenter stated that from personal experience, foreign utilities in
particular do not always recognize the distinction. The commenter
questioned the purpose for using this designation or making this
statement.
Response: The use of the term ``transportable'' in the SER, SAR, or
CoC is descriptive of the intended function. The use of this
terminology in a dry storage cask application or an NRC SER/CoC does
not represent a certification under 10 CFR Part 71 for the transport of
radioactive materials. This CoC does not authorize transportation under
Part 71.
Summary of Final Revisions
Section 72.214 List of Approved Spent Fuel Storage Casks
Certificate No. 1004 is revised by adding the effective date of
Amendment Number 5 and adding Model Number NUHOMS[reg]-32PT.
Good Cause To Dispense With Deferred Effective Date Requirement
The NRC finds that good cause exists to waive the 30-day deferred
effective date provisions of the Administrative Procedure Act (5 U.S.C.
553(d)). The primary purpose of the delayed effective date requirement
is to give affected persons; e.g., licensees, a reasonable time to
prepare to comply with or take other action with respect to the rule.
In this case, the rule does not require any action to be taken by
licensees. The regulation allows, but does not require, use of the
amended TN Standardized NUHOMS[reg]-24P, -52B, -61BT, and -
24PHB cask system for the storage of spent nuclear fuel. The TN
Standardized NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask
system, amended to include the new dry shielded canister designated -
32PT, meets the requirements of 10 CFR Part 72 and is ready to be used.
A general licensee has made plans to load the NUHOMS[reg]-
32PT casks in January 2004 to preserve full core off-load capability at
its site. The general licensee is currently in a refueling outage and
needs to load fuel into the storage casks once done. The amended TN
Standardized NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask
system, as approved by the NRC, will continue to provide adequate
protection of public health and safety and the environment.
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 the 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.
Voluntary Consensus Standards
The National Technology Transfer Act of 1995 (Pub. L. 104-113)
requires that Federal agencies use technical standards that are
developed or adopted by voluntary consensus standards bodies unless the
use of such a standard is inconsistent with applicable law or otherwise
impractical. In this final rule, the NRC is revising the Standardized
NUHOMS[reg]-24P, -52B, -61BT, and -24PHB cask system design
listed in Sec. 72.214 (List of NRC-approved spent fuel storage cask
designs). This action does not constitute the establishment of a
standard that establishes generally applicable requirements.
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 is not a major Federal action significantly
affecting the quality of the human environment and, therefore, an
environmental impact statement is not required. This final rule amends
the CoC for the TN Standardized NUHOMS[reg]-24P, -52B, -
61BT, and -24PHB cask system within the list of approved spent fuel
storage casks that power reactor licensees can use to store spent fuel
at reactor sites under a general license. The amendment modifies the
present cask system design to add another dry shielded canister,
designated NUHOMS[reg]-32PT DSC, to the authorized contents
of the Standardized NUHOMS[reg]-24P, -52B, -61BT, and -24PHB
cask system. This canister is designed to accommodate 32 PWR assemblies
with or without Burnable Poison Rod assemblies. It is designed for use
with the existing NUHOMS[reg] Horizontal Storage Module and
NUHOMS[reg] Transfer Cask. The environmental assessment and
finding of no significant impact on which this determination is based
are available for inspection at the NRC Public Document Room, One White
Flint North, 11555 Rockville Pike, Room O-1F23, Rockville, MD. Single
copies of the environmental assessment and finding of no significant
impact are available from Jayne M. McCausland, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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
If a means used to impose an information collection does not
display a currently valid OMB control number, the NRC may not conduct
or sponsor, and a person is not required to respond to, the information
collection.
Regulatory Analysis
On July 18, 1990 (55 FR 29181), the NRC issued an amendment to 10
CFR Part 72 to provide for the storage of spent nuclear fuel under a
general license in cask designs approved by the
[[Page 858]]
NRC. Any nuclear power reactor licensee can use NRC-approved cask
designs to store spent nuclear fuel if it notifies the NRC in advance,
spent fuel is stored under the conditions specified in the cask's CoC,
and the conditions of the general license are met. A list of NRC-
approved cask designs is contained in Sec. 72.214. On December 22,
1994 (59 FR 65920), the NRC issued an amendment to Part 72 that
approved the Standardized NUHOMS[reg]-24P and -52B cask
system design by adding it to the list of NRC-approved cask designs in
Sec. 72.214. Amendments No. 3 and 6 added the -61BT DSC and the -24PHB
DSC, respectively, to the system. On June 29, 2001, the certificate
holder, Transnuclear, Inc., submitted an application to the NRC to
amend CoC No. 1004 to permit a Part 72 licensee to add another DSC,
designated NUHOMS[reg]-32PT DSC, to the authorized contents
of the Standardized NUHOMS[reg]-24P, -52B, and -61BT cask
system. This canister is designed to accommodate 32 PWR assemblies with
or without Burnable Poison Rod Assemblies. It is designed for use with
the existing NUHOMS[reg] Horizontal Storage Module and
NUHOMS[reg] Transfer Cask.
The alternative to this action is to withhold approval of this
amended cask system design and issue an exemption to each general
licensee. This alternative would cost both the NRC and the utilities
more time and money because each utility would have to submit a request
for an exemption, and the NRC would have to review each request.
Approval of this final rule eliminates the problem described and is
consistent with previous NRC actions. Further, the direct final rule
will have no adverse effect on public health and safety. This direct
final rule has no significant identifiable impact or benefit on other
Government agencies. On the basis of this 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 alternative is believed to be satisfactory.
Therefore, this action is recommended.
Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this rule does not have a
significant economic impact on a substantial number of small entities.
The final rule affects only the licensing and operation of nuclear
power plants, independent spent fuel storage facilities, and
Transnuclear, Inc. These entities do not fall within the scope of the
definition of ``small entities'' set forth in the Regulatory
Flexibility Act or the NRC's size standards (10 CFR 2.810).
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. Therefore, a backfit
analysis is not required for this final rule because this amendment
does not impose any provisions that would impose backfits as defined in
10 CFR Chapter I.
Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996, the NRC has determined that this action is not a
major rule and has verified this determination with the Office of
Information and Regulatory Affairs, Office of Management and Budget.
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).
Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),
10168(c),(d)). Section 72.46 also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2244, (42 U.S.C. 10101,
10137(a), 10161(h)). Subparts K and L are also issued under sec.
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252
(42 U.S.C. 10198).
0
2. Section 72.214, Certificate of Compliance 1004 is revised to read as
follows:
Sec. 72.214 List of approved spent fuel storage casks.
* * * * *
Certificate Number: 1004.
Initial Certificate Effective Date: January 23, 1995.
Amendment Number 1 Effective Date: April 27, 2000.
Amendment Number 2 Effective Date: September 5, 2000.
Amendment Number 3 Effective Date: September 12, 2001.
Amendment Number 4 Effective Date: February 12, 2002.
Amendment Number 5 Effective Date: January 7, 2004.
SAR Submitted by: Transnuclear, Inc.
SAR Title: Final Safety Analysis Report for the Standardized
NUHOMS[reg] Horizontal Modular Storage System for Irradiated
Nuclear Fuel.
Docket Number: 72-1004.
Certificate Expiration Date: January 23, 2015.
Model Number: Standardized NUHOMS[reg]-24P,
NUHOMS[reg]-52B, NUHOMS[reg]-61BT,
NUHOMS[reg]-24PHB, and NUHOMS[reg]-32PT.
* * * * *
Dated at Rockville, Maryland, this 19th day of December, 2003.
For the Nuclear Regulatory Commission.
William D. Travers,
Executive Director for Operations.
[FR Doc. 04-313 Filed 1-6-04; 8:45 am]
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