[Federal Register Volume 69, Number 4 (Wednesday, January 7, 2004)]
[Rules and Regulations]
[Pages 849-858]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-313]



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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 
jmm2@nrc.gov.

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 jmm2@nrc.gov.

Paperwork Reduction Act Statement

    This final rule does not contain a new or amended information 
collection requirement subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the 
Office of Management and Budget, Approval Number 3150-0132.

Public Protection Notification

    If a means used to impose an information collection does not 
display a currently valid OMB control number, the NRC may not conduct 
or sponsor, and a person is not required to respond to, the information 
collection.

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]
BILLING CODE 7590-01-P