[Federal Register Volume 71, Number 237 (Monday, December 11, 2006)]
[Rules and Regulations]
[Pages 71463-71472]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-20962]



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  Federal Register / Vol. 71, No. 237 / Monday, December 11, 2006 / 
Rules and Regulations  

[[Page 71463]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AH93


List of Approved Spent Fuel Storage Casks: NUHOMS[supreg] HD 
Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

-----------------------------------------------------------------------

SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to add the NUHOMS[supreg] HD cask system to the list of 
approved spent fuel storage casks. This final rule allows the holders 
of power reactor operating licenses to store spent fuel in this 
approved cask system under a general license.

DATES: Effective Date: The final rule is effective on January 10, 2007.

ADDRESSES: Publicly available documents related to this rulemaking may 
be viewed electronically on the public computers located at the NRC's 
Public Document Room (PDR), Room O1F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland. The PDR reproduction contractor 
will copy documents for a fee. Selected documents can be viewed and 
downloaded electronically via the NRC's rulemaking Web site at http://ruleforum.llnl.gov.
    Publicly available documents created or received at the NRC are 
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/NRC/reading-rm/adams.html. From this site, the public can 
gain entry into the NRC's Agencywide Document Access and Management 
System (ADAMS), which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are any 
problems in accessing the documents located in ADAMS, contact the NRC 
PDR Reference staff at (800) 397-4209, (301) 415-4737, or by e-mail to 
[email protected].

FOR FURTHER INFORMATION CONTACT: Jayne M. McCausland, Office of Federal 
and State Materials and Environmental Management Programs, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
6219, e-mail [email protected].

SUPPLEMENTARY INFORMATION:

Background

    Section 218(a) of the Nuclear Waste Policy Act of 1982, as amended 
(NWPA), requires that ``[t]he Secretary [of the Department of Energy 
(DOE)] shall establish a demonstration program, in cooperation with the 
private sector, for the dry storage of spent nuclear fuel at civilian 
nuclear power reactor sites, with the objective of establishing one or 
more technologies that the [Nuclear Regulatory] Commission may, by 
rule, approve for use at the sites of civilian nuclear power reactors 
without, to the maximum extent practicable, the need for additional 
site-specific approvals by the Commission.'' Section 133 of the NWPA 
states, in part, that ``[t]he Commission shall, by rule, establish 
procedures for the licensing of any technology approved by the 
Commission under Section 218(a) for use at the site of any civilian 
nuclear power reactor.''
    To implement this mandate, the NRC approved dry storage of spent 
nuclear fuel in NRC-approved casks under a general license by 
publishing a final rule in 10 CFR Part 72 entitled ``General License 
for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 
18, 1990). This rule also established a new Subpart L within 10 CFR 
Part 72, entitled ``Approval of Spent Fuel Storage Casks,'' containing 
procedures and criteria for obtaining NRC approval of spent fuel 
storage cask designs.

Discussion

    On May 5, 2004, and as supplemented on July 6, August 16, October 
11, October 28, November 19, 2004; February 18, March 7, April 14, May 
20, May 24, August 16, 2005; and January 24, February 15, and September 
19, 2006, the certificate holder, Transnuclear, Inc. (TN), submitted an 
application to the NRC to add the NUHOMS[supreg] HD cask system to the 
list of NRC-approved casks for spent fuel storage in 10 CFR 72.214. The 
NUHOMS[supreg] HD System provides for the horizontal storage of high 
burnup spent pressurized water reactor fuel assemblies in a Dry 
Shielded Canister (DSC) that is placed in a horizontal storage module 
(HSM) utilizing an OS-187H transfer cask (TC). The system is an 
improved version of the Standardized NUHOMS[supreg] System described in 
Certificate of Compliance (CoC) No. 1004. The NUHOMS[supreg] HD System 
has been optimized for high thermal loads, limited space, and radiation 
shielding performance. The -32PTH DSC included in this system is 
similar to the -24PTH DSC submitted for licensing as Amendment No. 8 to 
the Standardized NUHOMS[supreg] System. The -32PTH DSC will be 
transferred during loading operations using the OS-187H TC. The OS-187H 
TC is very similar to the OS-197 and OS-197 TCs described in the final 
safety analysis report for the Standardized NUHOMS[supreg] System. The 
-32PTH DSC will be stored in an HSM, designated the HSM-H. The HSM-H is 
virtually identical to the HSM-H submitted for licensing as Amendment 
No. 8 to the Standardized NUHOMS[supreg] System. The NRC staff 
performed a detailed safety evaluation of the proposed CoC request and 
found that an acceptable safety margin is maintained. In addition, the 
NRC staff has determined that there continues to be reasonable 
assurance that public health and safety and the environment will be 
adequately protected.
    The NRC published a direct final rule (71 FR 25740; May 2, 2006) 
and the companion proposed rule (71 FR 25782) in the Federal Register 
to add the NUHOMS[supreg] HD cask system to the listing in 10 CFR 
72.214. The comment period ended on July 17, 2006. Six comment letters 
were received on the proposed rule. The comments were considered to be 
significant and adverse and warranted withdrawal of the direct final 
rule. A notice of withdrawal was published in the Federal Register on 
July 13, 2006; 71 FR 39520.
    Based on NRC review and analysis of public comments, the staff has 
modified, as appropriate, Technical Specifications (TS) and the 
Approved Contents and Design Features, for the NUHOMS[supreg] HD 
system. The staff has also modified its preliminary Safety Evaluation 
Report (SER). In particular, regarding the potential for the dry

[[Page 71464]]

shielded canister to corrode in a coastal marine environment, TN 
committed to specifying a weathering steel for Independent Spent Fuel 
Storage Installations (ISFSIs) located near a coastal marine 
environment. The staff made corresponding changes to the SER and added 
a requirement to TS 4.4.1 to capture this commitment for the HSM-H.
    The proposed TS and SER have been revised in response to Comment 2. 
Specifically, based on questions from the staff regarding this issue, 
TN committed in a letter dated September 19, 2006, to add the following 
to Section 3.4.1.4 of the Safety Analysis Report (SAR) for the NUHOMS 
HD design: ``If an independent spent fuel storage installation site is 
located in a coastal salt water marine atmosphere, then any load-
bearing carbon steel DSC support structure rail components of any 
associated HSM-H shall be procured with a minimum 0.20 percent copper 
content for corrosion resistance.'' This commitment has also been 
captured in NUHOMS[supreg] HD TS 4.4.1 for the HSM-H, and the staff 
made corresponding changes to SER Section 3.2.1 to document its 
evaluation.
    The NRC finds that the TN NUHOMS[supreg] HD cask system, as 
designed and when fabricated and used in accordance with the conditions 
specified in its CoC, meets the requirements of 10 CFR Part 72. Thus, 
use of the TN NUHOMS[supreg] HD cask system, as approved by the NRC, 
will provide adequate protection of public health and safety and the 
environment. With this final rule, the NRC is approving the use of the 
TN NUHOMS[supreg] HD cask system under the general license in 10 CFR 
Part 72, Subpart K, by holders of power reactor operating licenses 
under 10 CFR Part 50. Simultaneously, the NRC is issuing a final SER 
and CoC that will be effective on January 10, 2007. Single copies of 
the CoC and SER are available for public inspection and/or copying for 
a fee at the NRC Public Document Room, O-1F21, 11555 Rockville Pike, 
Rockville, MD.

Discussion of Amendments by Section

Section 72.214 List of Approved Spent Fuel Storage Casks

    CoC No. 1030 is added to the list of approved spent fuel storage 
casks.

Summary of Public Comments on the Proposed Rule

    The NRC received six comment letters on the proposed rule. The 
commenters included representatives from industry and members of the 
public. Copies of the public comments are available for review in the 
NRC's Public Document Room, O-1F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland.
Comments on the Transnuclear, Inc., NUHOMS[supreg] HD Cask System
    Several of the commenters provided specific comments on the NRC 
staff's preliminary SER and the TS. To the extent possible, the 
comments on a particular subject are grouped together. The listing of 
the Transnuclear, Inc., NUHOMS[supreg] HD cask system within 10 CFR 
72.214, ``List of approved spent fuel storage casks,'' has not been 
changed as a result of the public comments. A review of the comments 
and the NRC staff's responses follow:
    Comment 1: Three commenters raised issues with using Boral[supreg] 
for criticality control. One commenter pointed to documented widespread 
evidence of Boral degradation; e.g., in Spain, Boral was banned from 
all casks after evidence of Boral's swelling and hydrogen generation 
was found in laboratory testing, and in the U.S., Boral has exhibited 
swelling, blistering, and instances of major hydrogen gas generation in 
dry cask fuel storage applications. Two commenters noted that NRC 
issued Generic Safety Issue No. 196 to study the Boral degradation 
problem. Other remarks concerning Boral are noted as follows: (1) The 
problem has been occurring for 20 to 30 years; (2) Boral problems occur 
on a random basis, and it is impossible to predict the product's 
performance because of uncertainty in the level of porosity in the 
aluminum boron carbide core of the cladded product; (3) Boral was the 
material choice in past years mainly because there were no economical 
alternatives; (4) The use of Boral was understandable 10 or even 5 
years ago because fully dense metallic neutron absorbers were not 
commercially available then, but now aluminum alloy-based neutron 
absorbers with high boron content are produced by several suppliers; 
(5) Boral is used today only because of its cost savings to the cask 
supplier, and it is not worth putting the health and safety of workers 
who load the cask at risk; (6) From a metallurgical point of view, the 
most consistent performance will be demonstrated from an aluminum boron 
carbide neutron absorbing product which exhibits 100 percent of 
theoretical density, and only a fully dense neutron absorber will 
completely eliminate the potential of swelling and hydrogen gas 
generation phenomenon.
    Response: The NRC is aware that canisters containing 
BORALTM may generate hydrogen while the canister is 
submerged in the spent fuel pool during short-term loading operations. 
This was observed at the Columbia Generating Station in 2002. 
BORALTM will react with the spent fuel pool water during 
loading operations and generate hydrogen. The magnitude of the hydrogen 
generation could depend on many factors, such as pool water chemistry, 
batch-to-batch variations, time-at-temperature, etc. The hydrogen 
generation does not decrease the efficacy of the material as a neutron 
absorber. As is the case with most casks licensed by the NRC, the SAR 
for the NUHOMS[supreg] HD describes hydrogen generation mitigating 
procedures. Vendors of casks certified by NRC have recommended that the 
utilities monitor for hydrogen gas during loading operations and state 
that a purge be used when hydrogen gas concentration exceeds 2.4 
percent prior to or during root-pass welding of the lid.
    The NRC is aware that BORALTM can swell or blister under 
high temperatures and hydrostatic pressures as was observed in Spain. 
In October 2003, the NRC received a letter from the Empresa Nacional de 
Residuos Radiactivos, S. A. (ENRESA) concerning this matter in the 
Spanish cask. However, it is our understanding that the Equipos 
Nucleares, S.A (ENSA) test conditions, under which blistering was 
observed, were conducted at high heat-up rates and high hydrostatic 
pressures, well beyond those for operating conditions for the dry cask 
storage systems in the U.S. It is also our understanding that the high 
heat-up rates and hydrostatic pressures did not permit the liquid to 
drain prior to expanding, thereby leading to blistering. This was due 
to low porosity of the BORALTM matrix structure which does 
not facilitate water egress under the conditions mentioned above. The 
letter from ENRESA concerning this matter in the Spanish cask and the 
BORALTM blistering never stated that BORALTM has 
been banned from use in Spain. It should be noted that no U.S. vendors 
or utilities have reported any BORALTM blistering during 
loading operations or manufacturer acceptance testing of a cask.
    The staff in the Spent Fuel Storage and Transportation Division 
have shared data and reports with the staff in the office of Nuclear 
Regulatory Research concerning GSI-196, BORALTM degradation. 
All data, reports, and letters (domestic and foreign) provided to 
ascertain criticality implications of BORALTM degradation in 
the context of dry cask storage of spent fuel have shown that the 
efficacy

[[Page 71465]]

was not reduced in BORALTM used in dry cask storage systems.
    Blistering or swelling in BORALTM has been reported to 
occur under wet storage conditions in the spent fuel pools at both 
domestic and foreign reactors. For example, in September 2003, FPL 
Energy Seabrook, LLC, reported bulging of the BORALTM coupon 
used to monitor the performance of the spent fuel pool racks. The 
bulging of this coupon was due to blistering. FPL's examination and 
analysis of the coupon indicated no loss in the B-10 areal density.
    Neutron attenuation and radiography measurements have been 
conducted on the BORALTM test coupons--both seal-welded and 
vented--subjected to multiple wetting/drying cycles and varying heat-up 
rates to simulate wet storage and typical cask loading conditions. In 
the many test reports reviewed by the NRC staff, blistering usually 
occurred in the low-porosity (low B4C content) coupons. The data 
reported that the boron-10 areal density in the blistered specimens 
remained unaffected. Thus, neutron attenuation efficacy was not 
affected in the BORALTM. It should be noted that the 
Seabrook licensee, who reported blistering in the BORALTM 
coupons after about 7 years of wet storage in the spent fuel pool, 
reportedly demonstrated that BORALTM suffered no loss of 
effectiveness as a neutron absorber.
    The NRC is aware that other neutron absorber materials are now 
available to the cask vendors; however, the NRC does not recommend any 
brand of material to the vendors. To date, tests have shown that the 
BORALTM material still performs its intended function with 
or without the blisters being present.
    The NRC staff does not dispute the advantages of the near-
theoretical-density neutron absorber materials, which have become 
available in recent years. However, blistering has not been shown to 
affect dose to workers involved in the cask loading process. 
Additionally, if hydrogen gas is detected during the loading 
operations, the vendors and licensees can use mitigating procedures to 
vent and purge the cask. This procedure is recommended prior to 
welding; thus, worker safety can be ensured.
    The NRC staff does agree that this problem of blistering and 
hydrogen generation has not been reported in the absorber materials 
that have a 100-percent dense matrix. However, the NRC has reviewed 
evaluations by the Energy Power Research Institute (EPRI) and cask 
vendors, and for the most part, the boron areal density (10B/
cm2) in the blistered specimens remained unaffected. Thus, 
neutron attenuation was not affected, and there was no impact on 
BORAL's effectiveness as a neutron absorber.
    Comment 2: One commenter stated that the structural steel frame 
used to support the DSC poses a serious risk to public health and 
safety. The commenter made the following points: (1) From contact with 
the air and humidity in the environment, these structurals can corrode 
from the inside as well as from the outside. Particularly at coastal 
sites, anything that can corrode, will corrode. Even stainless steel 
develops stress corrosion cracks. (2) The upright tubes make up the 
only support structure for the fuel-filled canister. They cannot be 
inspected from the outside of the NUHOMS because they cannot be seen. 
All primary supports must be inspected periodically, and it is a fatal 
flaw to have a fuel storage canister perched about 6 feet in the air on 
top of a steel frame which cannot be inspected at all. It is a 
dangerous sort of design for unrestricted use around our country, 
including the plants in salt air environments.
    Response: Regarding Part (1), above, it is widely recognized that 
corrosion is a significant concern in coastal marine environments due 
to the wind borne salts deposited upon structures. Based on questions 
from the staff regarding this issue, TN committed in a September 19, 
2006, letter to add the following to Section 3.4.1.4 of the SAR for the 
NUHOMS[supreg] HD design: ``If an independent spent fuel storage 
installation site is located in a coastal salt water marine atmosphere, 
then any load-bearing carbon steel DSC support structure rail 
components of any associated HSM-H shall be procured with a minimum 
0.20 percent copper content for corrosion resistance.'' This commitment 
has also been captured in NUHOMS[supreg] HD TS 4.4.1 for the HSM-H. 
Consequently, the TN design incorporates a requirement to use 
atmospheric corrosion resisting steels (a.k.a., weathering steels) when 
the spent fuel storage site is near a coastal marine environment.
    A significant body of technical literature exists, which provides 
corrosion rate data for a variety of steel alloys exposed to the 
elements at coastal sites. From this data, TN recognized that 
weathering steels provide ample corrosion resistance in a coastal 
marine atmosphere. This corrosion resistance would assure that the 
accumulated corrosion loss over a 20-year license period would be 
immaterial to the structural integrity of the support steel inside the 
HSM-H.
    It should be noted that the data used to determine the required 
corrosion allowance are for samples fully exposed to the elements. It 
is known that samples that are fully shielded from the sun and rain 
show a significantly lower corrosion rate than fully exposed samples. 
The structural steel of the HSM-H is entirely enclosed inside a 
ventilated concrete structure that totally shields the steel from 
sunlight and precipitation. TN chose to employ the higher corrosion 
rate data for fully exposed samples as the basis for their corrosion 
allowance. This provides an added degree of conservatism to their 
design.
    In addition to the use of corrosion-resisting steels, TN has 
specified the application of a corrosion resistant coating over the 
support steel. The coating may be one of several systems. One system 
consists of an inorganic zinc primer with an epoxy overcoat. This is an 
industry-recognized, high performance, and long-lived industrial 
coating system that is designed to withstand very severe environments. 
Although the coating is specified, it is not credited in the corrosion 
rate calculations that are part of the structural steel design margins.
    The staff finds that the use of corrosion-resisting steel with a 
calculated corrosion rate derived from a more severe exposure 
environment is appropriate. Additionally, the staff finds that the use 
of a coating system, and the fact that the steel is enclosed in a dry, 
interior-like environment, provide additional protection against 
corrosion. Thus, the staff finds that this TN design provides 
reasonable assurance that the system will not experience any 
significant corrosion during the 20-year license period at a coastal 
spent fuel storage site.
    Regarding Part (2), the commenter is correct that the canister, in 
some models of the HSM, is supported in the vertical direction by a 
series of columns or legs, six in total, that are made of structural 
steel tubing. These columns are part of a three-dimensional welded and 
bolted frame anchored vertically and horizontally to the reinforced 
concrete storage module. The three pairs of columns that are each less 
than 3.5 feet long support a cross beam which then provides support at 
three locations for each of the two support rails. The framing design 
concept is similar to that used in structural steel framing of multi-
story buildings, tankage support systems, and other applications where 
a three-dimensional framing concept is appropriate. In this case, since 
the frame

[[Page 71466]]

is provided with lateral supports at the location of each column to the 
reinforced concrete horizontal storage module, the frame is considered 
to be a braced-frame and, therefore, has limited lateral deflection 
that can occur at the top of the frame. The design concept is not 
considered to be unique, out-of-the-ordinary, or a dangerous design 
configuration for this intended use. The design conditions that 
represent the environment in which the frame must function have been 
incorporated into the design criteria. In other models of the HSM, the 
support rails are supported directly on the reinforced concrete storage 
module by embedded anchors. The NUHOMS[supreg] HD support rails are 
supported and anchored in this manner.
    The commenter used the term ``primary support'' and indicated that 
all primary supports must be inspected periodically. While the 
NUHOMS[supreg] HD can be used at a nuclear power plant, the 
certification of the dry spent fuel storage system is carried out under 
10 CFR part 72 and not 10 CFR part 50. Consequently, the assertion made 
by the commenter that ``all primary supports'' must be inspected 
periodically may be in reference to a requirement in 10 CFR 50.55a(f), 
for inservice testing requirements for nuclear power reactor facilities 
for various classes of components. These 10 CFR part 50 requirements do 
not apply to the passive systems that are under the jurisdiction of 10 
CFR part 72. The design criteria used for the design of the 
NUHOMS[supreg] HD system, to support the canisters in the horizontal 
storage module, are sufficiently robust so that periodic inservice 
inspections of these structural components are not deemed to be 
necessary. It is correct that there is a requirement that is identified 
in 10 CFR 72.122(f) related to testing and maintenance of systems and 
components that are important to safety. Such systems and components 
are to be designed to permit inspection. The NUHOMS[supreg] HD rail 
support system could be visually inspected by remote operations using 
fiber optics into the HSM-H via the vent system, or the HSM-H can be 
opened, the canister extracted into the transfer cask, and the rail 
supports inspected, after appropriate radiation surveys and procedures 
are met. The environmental concern in Part (2) of the comment is 
addressed in Part (1) response.
    Comment 3: A commenter raised the following concern with respect to 
flooding: Section 4.6.3 of the Generic Technical Specification states 
that flood ``levels up to 50 feet and water velocity of 5 fps'' are 
allowed. The commenter was concerned about the flooding condition in 
which the floodwater rises to fill the inlet ducts in NUHOMS[supreg] 
(all of the air inlet ducts in the NUHOMS[supreg] module lie at the 
ground level). He questioned that if the floodwater rises high enough 
to block off the air flow through the inlet ducts, the DSC would not 
cool and concluded that without the ventilation airflow, the DSC would 
overheat and may even explode from pressure buildup. It seemed to the 
commenter that TN considered only the case of deep submergence flood in 
the safety evaluation, which is not a risky condition because the DSC 
is cooled by the flood water. The commenter further stated that low 
flood level is a risky condition since the DSC is several feet above 
the ground, and a flood of any height that remains below the DSC will 
choke off the ventilation air and cause the DSC to overheat. The 
commenter was surprised that NRC would issue ``general certification'' 
to a ventilated cask like this one to be used in flood plains, 
considering that there are many ``nukes'' on river basins that are in 
the potential flood zone. The commenter further stated that the 
condition of partial height flood should be given full technical 
consideration.
    Response: Regarding low level floods in the situation when the 
bottom vents are blocked, evaporative cooling will cool the upper 
volume of the HSM and the DSC as demonstrated below. A thermal analysis 
of a typical HSM and DSC with a fuel heat load of 24kW in accident 
conditions demonstrates that the DSC support steel maximum temperature 
is 615 [deg]F, and the DSC shell maximum temperature is 642 [deg]F. 
These component temperatures would provide evaporation of the water in 
the bottom of the HSM. The evaporated water would cool the DSC and the 
upper volume of the HSM. The staff notes that the NUHOMS[supreg] HD 
technical specification maximum heat load is 34.8 kW. Even at the 
higher heat loads, staff believes that evaporative cooling will prevent 
the DSC from overheating. In addition, the flood water will help cool 
the submerged portion of the HSM cavity. Therefore, the staff concludes 
that the DSC will not overheat, and the resulting DSC internal 
pressures will not exceed the design pressure.
    Comment 4: One commenter believed that TS 4.6.3 was unclear in the 
statement that NRC has allowed ``seismic loads of up to 0.3 g 
horizontal and up to 0.2 g vertical'' on the system. The commenter 
asked for the location in the storage facility to which the g-loads 
correspond, either at the C.G. of the storage system or at the pad 
surface on the module's centerline, and also asked if the g-load limits 
include the effect of soil-structure interaction alluded to in 
Paragraph 4.2.2. Another commenter assumed that the 0.3 g horizontal 
and 0.2 g vertical seismic events (per page 4-7 of Design Features in 
the Certificate) are free-field accelerations at the site and stated 
that they will get amplified at the pad due to soil-structure 
interaction. The on-the-pad accelerations will be further magnified at 
the rails due to the flexibility of the DSC support structure. Combined 
with the rattling impulse from the fuel, the commenter believed that a 
canister may roll off the rails.
    Response: The permissible seismic loads of 0.3 g horizontal and 0.2 
g vertical noted by the first commenter are the maximum values at the 
top of the HSM-H or the top of the supporting basemat or pad the 
NUHOMS[supreg] HD system is allowed to be subjected to. The design of 
the HSM-H and the NUHOMS[supreg] HD system is based on the amplified 
response spectra value of 0.37 g in the orthogonal horizontal direction 
and 0.20 g in the vertical direction on the 0.3 g and 0.2 g values 
respectively. The 0.30 g horizontal and 0.20 g vertical values also 
reflect the resulting maximum permitted accelerations at the top of the 
basemat or pad after a soil-structure interaction analysis has been 
performed, if necessary, by the cask system user for the specific site 
using the site-specific free field g-values. The fact left unstated is 
that where a soil-structure interaction analysis must be performed by 
the user, the resulting amplified response value at the center of 
gravity of the loaded HSM-H must not exceed 0.37 g in the horizontal 
direction and 0.20 g in the vertical direction. Based on the proposed 
rule, if either of these values were exceeded, the NUHOMS[supreg] HD 
system could not be used.
    The interpretation of the second commenter is not what is reflected 
in the TS as discussed above. The TS g-values are not generally 
consistent with the free-field acceleration values at most sites.
    The design conditions have included analyses of the canister in 
place on the rail support system under the design lateral loads from 
the seismic events, and there is no canister roll off from the rail 
support system.
    Comment 5: One commenter found that the DSC support structure is 
not restrained against all four walls of the concrete module. A 45-ton 
container resting unsecured on the rails that are not braced against 
the four walls is a physically unstable arrangement. The commenter 
asked if this configuration had been analyzed to ensure that failure

[[Page 71467]]

from resonance would not occur during earthquakes. The commenter stated 
that he could not find any evidence of such an evaluation in the TSAR 
or the NRC's SER.
    Response: It is unclear to the NRC staff what the source and basis 
are for these comments. The comments do not relate to the 
NUHOMS[supreg] HD system. There is no document identified as the TSAR 
(Topical Safety Analysis Report) associated with this docket 
application (72-1030). This terminology was associated with 
applications submitted in the late 1980s and early 1990s (e.g., TN-24 
and TN-32 cask systems). The commenter's description of the DSC support 
structure does not match that of the NUHOMS[supreg] HD system. For the 
NUHOMS[supreg] HD system, the DSC support structure consists of a pair 
of structural steel rails of 12-inch deep wide-flange sections that are 
anchored to the reinforced concrete horizontal storage module at the 
bottom flanges and connected by two struts and are, therefore, 
considered braced. This configuration is provided in the SAR for the 
NUHOMS[supreg] HD system. The seismic analysis determined that 
amplified accelerations are based on the frequency analysis, so that 
any issue of resonance has been incorporated into the analysis and then 
into the design of the individual members.
    Comment 6: One commenter believed that being able to remove the 
container at the end of 20 years of licensed life should be an 
important safety consideration. The commenter inquired and found that 
no plant that has loaded a NUHOMS[supreg] in the country has ever 
attempted to remove the container after a few years of storage. The 
commenter wanted to know what would happen if the aging of the rails 
and container's surfaces due to years of weathering were to cause the 
canister to bind to the rails.
    Response: The canister itself is constructed of stainless steel. 
The top of the support beam has a stainless steel cover plate welded 
along its entire length. This stainless steel plate forms the surface 
upon which the canister rests and also serves as a sliding surface for 
canister installation or removal operations. This plate may be 
lubricated if desired.
    Long-term experiments, where stainless steel samples were exposed 
to the weather at coastal marine sites, have demonstrated that 
stainless steel is highly resistant to atmospheric corrosion under 
those conditions. In the case of the TN NUHOMS[supreg] HD design, the 
canister and related support rails are shielded from direct exposure to 
the weather (being enclosed in a ventilated enclosure). This sheltering 
from the direct weather would result in little, if any, corrosion 
compared to the already insignificant amounts that could occur if these 
components were fully exposed to the weather. Absent corrosion, there 
is no likelihood that the canister would bind to the support rails. 
Because of this, and the fact that a lubricant (grease) could be 
applied to the rails, if desired, the staff believes it to be highly 
unlikely that any difficulty would arise during a removal operation, 
even after an extended period of time.
    Comment 7: A commenter asked what would happen if uneven settlement 
of the pad from the heavy weight of the module were to cause the 
canister to bind to the rails.
    Response: Uneven settlement of the pad, commonly referred to as 
differential settlement, is not expected to occur. If it were to occur, 
it is highly unlikely that it would result in any differential movement 
between the two supporting rails for the canister that would cause the 
canister to bind to the rails. The reinforced concrete pad and the 
reinforced concrete horizontal storage module represent a very stiff 
structural combination, so that relative movement between the support 
rails cannot be logically projected based on the structural response 
from any differential settlement across the supporting base pad. 
Further, the adequacy of the pad to support the horizontal storage 
module, without detrimental settlements, is required under the 
requirements of 10 CFR 72.212. The adequacy must be maintained under 
static and dynamic loads of the storage cask system, considering 
potential amplification of earthquakes through soil-structure 
interaction, soil liquefaction, and other soil instabilities due to 
vibratory ground motion, if these conditions exist at a site. Binding 
of the canister to the support rails from settlement or differential 
movement is not expected under any design condition.
    Comment 8: A commenter asked what would happen if the 60 kips of 
permissible extraction force to remove the container are not 
sufficient. The commenter stated that this scenario is ignored in the 
Technical Specification of TN's TSAR.
    Response: See also response to Comment 5 regarding a document 
misidentified as TN's TSAR. If settlement or differential settlement of 
a limited magnitude were to develop over the years, the transport 
trailer is equipped with hydraulic jacks or positioners and an 
alignment system, identified as the skid positioning system that is 
normally used for the alignment of the transfer cask. This same system 
can be used to accommodate effects resulting from limited settlement or 
differential settlement between the basemat or storage pad and the 
approach slab. If a situation were to develop where the support skid 
positioning system could not accommodate the magnitude of the movement, 
the approach slab can be modified or other measures taken.
    Comment 9: A commenter stated that the NUHOMS[supreg] HSM is much 
heavier and bigger than the previous models, noting that each loaded 
module weighs over 200 tons and questioned whether the ground 
underneath the NUHOMS[supreg] housing would settle over the years under 
the weight of the modules. The commenter also cited NRC's SER on page 
3-7: ``It is assumed that an axial load of 80 kips is required for 
insertion and 60 kips for extraction,'' and stated that this seems 
backwards. More force will be needed to extract the canister than to 
insert it (when the rail is new and greased). The commenter questioned 
how a safety concern would be addressed if because of settlement and 
weather effects, 60 kips is not enough to pull the canister out, and 
how the NUHOMS[supreg] would be emptied of fuel if the canister binded 
to the rails. The commenter believed that this would be a huge concern 
to people living near the NUHOMS[supreg] sites. He further stated that 
the minimum the NRC should do is to require that a demo of canister 
extractions at a couple of sites loaded with NUHOMS for 10 years (or 
more) be done to prove that the horizontally loaded canister can be 
successfully extracted.
    Response: With regard to the commenter's concern about the weight 
of NUHOMS[supreg] HSM, the 80-kip insertion load, and the 60-kip 
extraction load, it is noted that as stated in the SER on page 3-7, 
these are the design load conditions under normal operation loading 
conditions. In the off-normal operation loading condition, the 
extraction force can be allowed to reach 80 kips under that design 
condition. The dry cask storage system has been evaluated against the 
regulatory requirements for retrievability of the spent fuel, and a 
demonstration of canister extraction from the horizontal storage module 
is not deemed necessary at some time after 10 years of storage. The 
extraction system has been determined to be capable of functioning 
during the term of the certificate.
    Comment 10: A commenter stated that he could not find any 
evaluation of safety for the following scenarios when the DSC is being 
inserted into the HSM:
    Scenario 1: The transfer cask skid has been unfastened from the 
trailer and the

[[Page 71468]]

transfer cask lid has been removed making the DSC axially unrestrained, 
but before the skid has been fastened to the HSM and the hydraulic ram 
has been engaged to the DSC grapple ring. An earthquake during this 
period, depending on its magnitude, has the potential to cause 
uncontrolled DSC movement and cause a significant radiation exposure 
event to the workers that could be potentially deadly to the workers.
    Scenario 2: The DSC has been installed in the HSM, but the HSM lid 
(a heavy circular lid that also restrains the DSC in the axial 
direction) is not yet in place. An earthquake during this period could 
cause a major radiation exposure event that could be potentially deadly 
to the workers.
    Response: Scenario 1: For the described scenario, the position of 
the transfer cask for the NUHOMS[supreg] HD system, before the lid is 
removed, is on the transfer trailer, with the cask within several feet 
of the open HSM-H cavity, after the centerlines of the HSM-H and the 
cask have been verified to be approximately coincident. The lid of the 
cask is then removed. The transfer trailer is then backed to within a 
few inches of the face of the HSM-H, the trailer brakes are set, and 
the tractor is disconnected from the trailer and moved away. The 
transfer trailer vertical jacks are positioned to locate the vertical 
position of the cask in its approximate insertion orientation. The skid 
tie-down bracket fasteners are removed, and the position of the cask is 
corrected, as needed for alignment, using the hydraulic skid 
positioning system. Then, the optical survey equipment and reference 
marks are used for adjusting the final alignment. The skid positioning 
system is then used for that final alignment, and the canister is 
inserted into the HSM-H access opening docking collar. The transfer 
cask is then secured to the HSM-H using the cask restraints.
    A large seismic event, during the period of time from when the 
transfer cask lid is removed and is several feet from the HSM-H, and 
before the transfer cask is anchored to the HSM-H with a sufficiently 
large horizontal axial component, could overcome the frictional 
resistance that keeps the canister inside the transfer cask. This would 
not, however, be an uncontrolled DSC movement, since the DSC inside the 
transfer cask has only an approximately \1/4\-inch radial gap, which 
controls the movement to essentially longitudinal/axial movement with 
the maximum lateral position of the DSC changing by approximately \1/
64\-inch for each inch of longitudinal/axial movement. The 
longitudinal/axial movement is limited by the distance of several feet 
between the transfer cask opening and the face of the HSM-H. A 
longitudinal/axial movement of 3 to 5 feet of the DSC from the transfer 
cask opening would not constitute an uncontrolled DSC movement, since 
that longitudinal/axial movement is limited by the face of the HSM-H 
module.
    The possibility of the hypothesized scenario is considered to be 
much less than what is considered significant for design accident 
conditions arising from handling and storage of spent nuclear fuel. The 
seismic event, to produce the hypothesized movement, must have a large 
enough component of acceleration in the longitudinal/axial direction of 
the positioned transfer cask that can be at any point on the compass, 
and the event must occur within a time period of 2 to 4 hours. On an 
annual basis, this would occur only three to five times per year for a 
given facility. If such a remote accidental event were ever to occur, 
plant operations personnel would respond by placing temporary shielding 
with equipment over any exposed portion of the DSC.
    Scenario 2: The operations' procedures identify that upon 
disengagement of the transfer cask from the HSM-H, the canister's axial 
seismic restraint is installed. This is a design feature that uses a 
structural steel embedment in the reinforced concrete of the HSM-H as 
the anchor point for the retainer device. The commenter's assumption 
that the HSM-H lid or door is the axial retainer for the canister is 
incorrect.
    Comment 11: One commenter stated that the DSC is pushed into the 
HSM module using a simple hydraulic ram that has no redundant load 
handling features. A simple failure such as loss of hydraulic pressure 
during the pushing operation would leave the DSC in a partially 
inserted configuration. The commenter believed that a single failure 
proof ram system should be required or TN should demonstrate that a ram 
failure halfway through the DSC pushing process can be dealt with using 
credible recovery measures. The commenter did not believe that NRC has 
ever considered this issue or that TN has ever been asked to provide an 
answer.
    Response: The functioning of the ram operating system is not 
considered to be a system that is safety related since the canister is 
confined and shielded during the period of ram operations. A failure in 
the location, as hypothesized by the commenter, presents an operational 
problem, but no significant issues are created. The corrective action 
would be to repair the operating system of the ram. NRC has considered 
this scenario, and the NRC agrees with the safety classification of the 
ram assembly that it is ``Not Important To Safety'' as identified in 
Table 2-5 of the applicant's SAR.
    Comment 12: A commenter stated that the DSC, according to NRC's 
SER, can survive the drop from 80 inches height, but was concerned 
about how a dropped DSC would be lifted from the pad. The DSC seems to 
have no lifting or handling attachments except for the grapple, which 
is useable only to engage the ram for a horizontal push.
    Response: The commenter is correct in that there are no lifting or 
handling attachments other than the grapple ring for a loaded canister. 
The DSC is placed into the transfer cask within the fuel pool and then 
is loaded with spent fuel. Then, after removal from the fuel pool and 
preparation for transfer, the closed cask is moved on the transfer 
trailer in a horizontal orientation to a location outside the fuel 
handling building. The transfer trailer and cask with the DSC closed 
inside are moved to the pad area. The DSC is not lifted out of the 
transfer cask, but is pushed out of the cylindrical transfer cask 
directly into the HSM-H in a horizontal position, with the transfer 
cask coupled to the HSM-H, creating a connecting tunnel space 
completely enclosing the DSC. This operating procedure makes the 
possibility of a dropped DSC on the pad extremely unlikely and an 
accident that is beyond the design basis accident. If a beyond design 
accident condition were to arise where a loaded and unshielded DSC had 
to be lifted, the first step would be to provide temporary shielding 
and probably execute a remote lift in the horizontal position with a 
device brought in for special use. Such special procedures can be 
developed for an accident condition response. It should be noted that 
the 80-inch side drop is for the DSC inside the transfer cask.
    Comment 13: A commenter stated that NRC should require a stiff 
foundation underneath the NUHOMS[supreg] to support the weight of the 
NUHOMS[supreg]. At present, the commenter sees nothing in the proposed 
certificate that requires a strong support foundation to be built. He 
believes this to be a serious oversight.
    Response: The weight of the NUHOMS HD[supreg] system, as installed 
in-place, including the HSM-H, the DSC, and the spent fuel, is to be 
supported by the ISFSI basemat or pad. That structure is identified in 
accordance with 10 CFR 72.3 as ``Not Important to Safety.'' The basemat 
or pad is designed, constructed,

[[Page 71469]]

maintained, and tested as a commercial grade item designed to be in 
compliance with 10 CFR 72.212(b)(2). This regulation requires that the 
user of the NUHOMS[supreg] HD cask system must evaluate and establish 
that the following criteria are met:
    (1) The cask storage pads and areas have been designed to 
adequately support the static and dynamic loads of the storage casks, 
considering potential amplification of earthquakes through soil-
structure interaction and soil liquefaction potential or other soil 
instability due to vibratory ground motion.
    (2) For the HSM-H loaded with a filled -32 PTH DSC, the weight is 
approximately 207.5 tons that is distributed over the pad area, which, 
as a minimum, is approximately 200 square feet.
    (3) The static load bearing pressure on the supporting soil 
material would normally be approximately 2075 pounds per square foot, a 
common value used for residential and commercial building foundations 
on fine-grained soils.
    (4) The loading on the foundation is not considered to be 
structurally significant or unusually high.
    Comment 14: A commenter expressed the following concerns pertaining 
to storing fuel horizontally in a hot state:
    (1) After searching the public filings by TN on this docket and 
Docket No. 72-1004, the commenter could not find a single evaluation of 
the consequences of storing fuel horizontally over long periods of 
time. In discussions between Westinghouse and a utility, the conclusion 
that they reached was that ``additional analyses and evaluation will be 
needed to determine whether it is permissible to store Westinghouse's 
fuel horizontally.''
    (2) A lot of fuel is already in NUHOMS[supreg] at many sites. What 
is happening to all of the fuel stored outside of the fuel supplier's 
(Westinghouse's) specifications is unknown because the condition cannot 
be examined.
    Response: In response to (1), after searching the TN filings, one 
document was found in which Westinghouse stated that ``* * * additional 
analyses and evaluation may be needed * * *.'' The NRC staff 
independently performed a generic analysis of spent fuel stored 
horizontally under the design service condition and for the service 
life of the NUHOMS[supreg] storage system. This analysis looked at the 
structural capability of the spent fuel materials to perform in the 
horizontal position without degrading spent fuel performance.
    There are two sources of stress in the fuel cladding, when in the 
horizontal orientation, that could result in creep. These are internal 
pressurization of the fuel rod and gravity. Two possible sources of 
deformation of the cladding, bending and creep, are possible under the 
horizontal position. The bending stress and the hoop stress are both 
considerably less than the yield stress under internal pressure and a 
horizontal position. The bending deflection, at the center of the span 
between the grid spacers, due to the downward gravitational load of the 
fuel, is approximately 3 millimeters. No changes occur in the stresses 
or radial growth as a result of storage in the horizontal position. The 
creep deformation is self limiting under both stresses due to the 
decreasing temperature of the fuel with time. If the initial maximum 
temperature is kept below 400 [deg]C, as recommended by Interim Staff 
Guidance (ISG)-11, then the creep deformation under the maximum 
allowable pressurization is less than 1 percent over a 20-year storage 
period. No cladding failure is expected at this strain level. The 
additional downward load, due to the gravitational force from the 
unsupported, approximately 300 grams of fuel between the grid spacer 
supports, increases the longitudinal stress by no more than 1 percent 
of the material strength and results in a minuscule increase of the 
hoop stress. Therefore, no more additional creep is expected in the 
horizontal orientation than in the vertical orientation.
    In response to (2), the cask vendors specify the range of 
parameters for the fuel to be stored in the CoC. The worst case fuel is 
analyzed as in paragraph (1), above. The fuel is evaluated when it is 
removed from the reactor to determine if it falls in the specified 
envelope. If it is in this envelope, no adverse fuel performance is 
expected.
    Comment 15: A commenter stated that, in the future, the fuel that 
will be stored will have burned longer in the reactor. The commenter 
believed that the NRC should perform a careful safety evaluation before 
permitting even more fuel, particularly well burned fuel, to be stored 
horizontally. The commenter cited NRC's SER on page 4-6 that reads: 
``The NUHOMS HD DSC only undergoes a one-time temperature drop during 
backfilling of the DSC with helium gas. Because this is a one-time 
event, the DSC does not undergo any thermal cycling.'' The commenter 
stated that the SER evidently assumes that the fuel will never be 
unloaded, unpackaged, and reloaded after it has been vacuum dried and 
backfilled. If that is the underlying basis of the SER, the commenter 
believes that the certificate should be restricted to only once-through 
loading such that there is no likelihood of thermal cycling of the 
fuel.
    Response: The staff has performed a safety evaluation and analyzed 
the effects of these parameters on the storage of fuel as provided in 
the guidance contained in ISG-11, Rev. 3. Higher burnup fuels will have 
the following characteristics:
    (1) A higher cladding stress caused by a higher internal pressure 
due to an increased fission gas release from the pellets;
    (2) A higher hydrogen content in the cladding resulting in a 
decrease in mechanical properties; and
    (3) A higher heat generation rate.
    As long as the fuel burnup is below the approved in-reactor burnup 
limit (currently 62.5 GWd/MTU) and is maintained in a nonoxidizing 
atmosphere below 400 [deg]C, there are no active degradation mechanisms 
that would cause cladding breaches to occur under normal storage 
conditions. In addition, the structural review must include mechanical 
properties of the cladding at the limit of the approved burnup to 
determine the behavior of the fuel under off-normal and accident 
conditions.
    The staff has evaluated the issue of thermal cycling on the 
behavior of irradiated fuel. Two issues of concern were thermal shock 
during reflood, if wet unloading occurs, and hydride reorientation. 
Reflood analysis is required in every SAR to evaluate the ability of 
the cladding to tolerate the thermal shock to the cladding due to the 
rapid submergence of the hot fuel in the cool pool water. For the 
NUHOMS[supreg] HD unloading operation, the maximum fuel cladding 
temperature during cask reflood is calculated to be significantly less 
than the vacuum drying condition because of the presence of water 
vapor. Consequently, during cask reflood, a lower temperature rise is 
expected when compared with that for the cask vacuum drying operations.
    Hydride reorientation, which might degrade the mechanical 
properties of the cladding, occurs when hydrogen goes into solution and 
is subsequently precipitated under stress during cooling. A number of 
studies indicate that thermal cycling may contribute to the phenomena 
of reorientation. To limit the occurrence of hydride reorientation in 
the cladding during storage, drying, etc., ISG-11, Rev. 3, limits the 
number of thermal cycles that the fuel can experience to 10 or less. 
Thermal cycling is only a concern if thermal

[[Page 71470]]

cycling takes place early in the storage period when the fuel is 
relatively hot.
    Under normal storage conditions, there are no mechanisms to degrade 
the fuel to the point where a loaded cask would have to be opened 
prematurely. At later times in the storage period, when unloading and 
repackaging are expected to occur, the temperatures will be at a lower 
maximum temperature due to the reduced decay heat, and as a result, 
less hydrogen (the solubility decreases exponentially with temperature) 
will be able to go into solution during these operations. In addition, 
the maximum stress in the rods will be less than at the initial vacuum 
drying, due to the lower temperature during unloading and repackaging. 
As a result, hydride reorientation, and consequently thermal cycling, 
is not of concern during unloading later in the storage period.
    Comment 16: A commenter stated that ``NRC's SER says that--The 
application performed dynamic impact analysis using LS-DYNA 3D on a 
cask-pad-soil finite element model * * *.'' The commenter believed that 
this was not true and noted that the FSAR shows that the applicant used 
a cookbook approach, developed by EPRI in the time when LS-DYNA was not 
widely used, which is considered to be unconservative by most experts. 
The commenter further stated that, according to the experts he 
consulted, a true LS-DYNA analysis would have shown much greater g-
loads under an 80-inch drop. Therefore, the SAR analysis on which the 
NRC has relied is inadequate and unconservative.
    Response: The analytical method used by the applicant referred to 
by the commenter was performed as described in the NRC's SER using 
NUREG/CR-6608, dated February 1998, using LS-DYNA 3D. This is a 
commercial finite element dynamic analysis software package capable of 
three-dimensional representations. The DYNA 3D software package used in 
the development of the analysis procedure described in NUREG/CR-6608 by 
Lawrence Livermore National Laboratory is the comparable software 
package that has been used in the national laboratories. The analytical 
approach used in NUREG/CR-6608 is considered by NRC as an acceptable 
method of evaluation for low-velocity impacts such as a dropped cask. 
It is recognized that, in this approach, the transfer cask internals 
that include the canister, the fuel basket, and the spent fuel are 
modeled only by their mass and their mass distribution.
    Comment 17: A commenter believed that the tornado missile analysis 
in Chapter 11 of the NUHOMS[supreg] FSAR does not consider the damaging 
scenario of missile impact. The commenter stated that the analysis 
assumes impact over the concrete walls. The most dangerous impact would 
occur if the missile were to hit the fasteners that keep the door of 
the HSM in place. If the fastener fails from the missile impact, then 
the door will come loose and the canister will be uncovered, exposing 
people nearby to radiation. The commenter did not see any evaluation of 
this scenario in TNs FSAR or NRC's SER.
    Response: The scenario proposed by the commenter, while not 
specifically identified, is encompassed by and bounded by the scenarios 
specifically discussed in the referenced documents. First, it is 
necessary to have an accurate understanding of the physical 
configuration of the door of the HSM-H and the opening for the door on 
the front wall of the HSM-H base assembly. The door thickness is a 
total of 2.53 feet made up of 0.65 feet of steel, and the remainder is 
made of concrete. Approximately 97 percent of the total thickness of 
the door is inside the plane of the outside face of the HSM-H, filling 
the recessed hole. The door is supported within the hole on two radial 
bearing pads that support the door on the 1.875-foot thickness of 
concrete of the 2.53-foot door thickness. The door is not supported in 
the vertical direction by the fasteners that the commenter addressed. 
The failure of one of those fasteners, as a result of a local missile 
impact, would not dislodge the door from the HSM-H base unit, and the 
door's radiation shielding capability would remain. Since the relevant 
missiles used to evaluate local missile damage effects all have 
physical dimensions and resulting damage zone dimensions much less than 
the spacing of the subject fasteners, multiple fastener loss is not 
likely. The fasteners' minimum spacing is approximately 5 feet, whereas 
the missiles considered relevant have maximum dimensions of 
approximately 1.5 feet. Even with multiple fastener failures, the thick 
door assembly will most likely remain in the deeply recessed opening 
after a local missile strike on the door's steel exterior, since the 
door assembly would have to move axially outward nearly 2 feet in order 
for the HSM-H to be rendered to a condition with an open door.
    Comment 18: A commenter expressed concern with the way the canister 
is stored. The commenter stated that it seems that the canister is 
lying on a couple of rails, and it is held in place by gravity and 
nothing else (no straps, no frame, no structurals to restrain it).
    Response: The commenter is correct that the canister is supported 
by two structural support rails. These are configured to create a 
cradle for the canister. The two rails of the cradle are each oriented 
at 30 degrees off the vertical centerline through the DSC, as it is in 
the stored horizontal position. With the 60-degree angle between the 
rail supports, a simple calculation demonstrates that a side load, 
through the center of gravity of the DSC, would have to exceed 
approximately 0.55 grams to disturb the at-rest position of the stored 
cask. This value, for lateral load, exceeds the control limits that are 
placed on this system, regarding the sites where the system could be 
used. That results in a design transverse load of 0.41 grams on the 
DSC. In the longitudinal direction, the DSC is restrained from movement 
on the rail support system by the axial retainer system that restrains 
DSC movement, with respect to the HSM-H.
    Comment 19: A commenter understood that the fuel is stored in the 
canister in a non-fixed manner and that during an earthquake, the fuel 
would move in the canister. The commenter inferred from reading the SAR 
that most of the canister's weight is in the fuel. He stated that if 
most of the weight is free to move about in the canister, then there is 
a risk of the canister rolling over and falling down during an 
earthquake.
    Response: The maximum values for comparing weight distribution for 
a loaded DSC are that 46.6 percent of the total weight of a loaded DSC 
is the weight of the spent fuel and the other 53.4 percent is the 
weight of the canister, the internal basket, and other hardware of the 
cask. The internal fuel basket is a cellular structure that provides a 
storage position 8.7 inches by 8.7 inches in cross-section for each of 
the 32 spent fuel assemblies that are stored. The orthogonal grid of 
the assemblage of these 32 cells is circumscribed by a circle created 
by metallic basket rails that transition from the grid configuration to 
a circle concentric with the inside surface of the canister. The radial 
space from the fuel basket and basket rails to the inside face of the 
canister is one-eighth of an inch. This configuration does not allow 
gross freedom of movement of the stored fuel, but only provides 
sufficient space to allow for loading and unloading of the spent fuel 
and for the thermal growth that is expected. Consequently, there is 
minimal lateral displacement of the spent fuel that can occur inside 
the canister.
    Comment 20: One commenter stated that he did not find a time 
history analysis in Appendix 3.9.9.10.2 of the

[[Page 71471]]

SAR to determine if canister bouncing or rolling might occur. He also 
stated that it did not appear that the effect of soil-structure 
interaction was mentioned.
    Response: As described in Section 3.9.9.10.2 of Appendix 3.9.9 of 
the SAR, the seismic design basis for the HSM-H and the stored spent 
fuel in the canister is based on the maximum peak accelerations at the 
top of the basemat, or pad structure, not exceeding 0.3 grams in the 
horizontal direction or 0.20 grams in the vertical direction. For the 
sites where soil-structure interaction analysis is considered 
important, the user of the NUHOMS[reg] HD system will have 
to determine that these values are not exceeded. Additionally, as 
indicated in the TS, Section 4.0, Design Features, amplified seismic 
response spectra from such an analysis would be produced. The HSM-H 
system, with the stored canister, is based on a limit of 0.37 grams in 
both transverse and longitudinal directions and 0.20 grams in the 
vertical direction, at the center of gravity of the HSM-H, with respect 
to the amplified response spectra. Within these limits of 
accelerations, there will be no uncontrolled motion of the canister 
that would result in a safety issue.

Summary of Final Revisions

    The proposed TS and SER have been revised in response to Comment 2 
to capture and document TN's commitment to add the following to Section 
3.4.1.4 of the SAR for the NUHOMS[reg] HD design: ``If an 
independent spent fuel storage installation site is located in a 
coastal salt water marine atmosphere, then any load-bearing carbon 
steel DSC support structure rail components of any associated HSM-H 
shall be procured with a minimum 0.20 percent copper content for 
corrosion resistance.''

Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995 (Pub. 
L. 104-113) requires that Federal agencies use technical standards that 
are developed or adopted by voluntary consensus standards bodies unless 
the use of such a standard is inconsistent with applicable law or 
otherwise impractical. In this final rule, the NRC is adding the 
NUHOMS[reg] HD cask system to the list of NRC-approved cask 
systems for spent fuel storage in 10 CFR 72.214. This action does not 
constitute the establishment of a standard that establishes generally 
applicable requirements.

Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the Commission on June 30, 1997, 
and published in the Federal Register on September 3, 1997 (62 FR 
46517), this rule is classified as Compatibility Category ``NRC.'' 
Compatibility is not required for Category ``NRC'' regulations. The NRC 
program elements in this category are those that relate directly to 
areas of regulation reserved to the NRC by the Atomic Energy Act of 
1954, as amended (AEA), or the provisions of Title 10 of the Code of 
Federal Regulations. Although an Agreement State may not adopt program 
elements reserved to NRC, it may wish to inform its licensees of 
certain requirements via a mechanism that is consistent with the 
particular State's administrative procedure laws but does not confer 
regulatory authority on the State.

Finding of No Significant Environmental Impact: Availability

    Under the National Environmental Policy Act of 1969, as amended, 
and the NRC regulations in Subpart A of 10 CFR Part 51, the NRC has 
determined that this rule, if adopted, would not be a major Federal 
action significantly affecting the quality of the human environment 
and, therefore, an environmental impact statement is not required. This 
final rule adds an additional cask to the list of approved spent fuel 
storage casks that power reactor licensees can use to store spent fuel 
at reactor sites without additional site-specific approvals from the 
Commission. The EA and finding of no significant impact on which this 
determination is based are available for inspection at the NRC Public 
Document Room, 11555 Rockville Pike, Rockville, MD. Single copies of 
the EA and finding of no significant impact are available from Jayne M. 
McCausland, Office of Federal and State Materials and Environmental 
Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone (301) 415-6219, e-mail [email protected].

Paperwork Reduction Act Statement

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

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

Regulatory Analysis

    On July 18, 1990 (55 FR 29181), the Commission issued an amendment 
to 10 CFR Part 72. The amendment provided for the storage of spent 
nuclear fuel in cask systems with designs approved by the NRC under a 
general license. Any nuclear power reactor licensee can use cask 
systems with designs approved by the NRC to store spent nuclear fuel if 
it notifies the NRC in advance, the spent fuel is stored under the 
conditions specified in the cask's CoC, and the conditions of the 
general license are met. In that rule, four spent fuel storage casks 
were approved for use at reactor sites and were listed in 10 CFR 
72.214. That rule envisioned that storage casks certified in the future 
could be routinely added to the listing in 10 CFR 72.214 through the 
rulemaking process. Procedures and criteria for obtaining NRC approval 
of new spent fuel storage cask designs were provided in 10 CFR Part 72, 
Subpart L.
    The alternative to this action is to withhold approval of this new 
design and issue a site-specific license to each utility that proposes 
to use the casks. This alternative would cost both the NRC and 
utilities more time and money for each site-specific license. 
Conducting site-specific reviews would ignore the procedures and 
criteria currently in place for the addition of new cask designs that 
can be used under a general license, and would be in conflict with NWPA 
direction to the Commission to approve technologies for the use of 
spent fuel storage at the sites of civilian nuclear power reactors 
without, to the maximum extent practicable, the need for additional 
site reviews. This alternative also would tend to exclude new vendors 
from the business market without cause and would arbitrarily limit the 
choice of cask designs available to power reactor licensees. This final 
rulemaking will eliminate the above problems and is consistent with 
previous Commission actions. Further, the rule will have no adverse 
effect on public health and safety.
    The benefit of this rule to nuclear power reactor licensees is to 
make available a greater choice of spent fuel storage cask designs that 
can be used under a general license. The new cask

[[Page 71472]]

vendors with casks to be listed in 10 CFR 72.214 benefit by having to 
obtain NRC certificates only once for a design that can then be used by 
more than one power reactor licensee. The NRC also benefits because it 
will need to certify a cask design only once for use by multiple 
licensees. Casks approved through rulemaking are to be suitable for use 
under a range of environmental conditions sufficiently broad to 
encompass multiple nuclear power plants in the United States without 
the need for further site-specific approval by NRC. Vendors with cask 
designs already listed may be adversely impacted because power reactor 
licensees may choose a newly listed design over an existing one. 
However, the NRC is required by its regulations and NWPA direction to 
certify and list approved casks. This rule has no significant 
identifiable impact or benefit on other Government agencies.
    Based on the above discussion of the benefits and impacts of the 
alternatives, the NRC concludes that the requirements of the final rule 
are commensurate with the Commission's responsibilities for public 
health and safety and the common defense and security. No other 
available alternative is believed to be as satisfactory, and thus, this 
action is recommended.

Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
NRC certifies that this rule will not, if issued, have a significant 
economic impact on a substantial number of small entities. This final 
rule affects only the licensing and operation of nuclear power plants, 
independent spent fuel storage facilities, and TN. The companies that 
own these plants do not fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
Small Business Size Standards set out in regulations issued by the 
Small Business Administration at 13 CFR part 121.

Backfit Analysis

    The NRC has determined that the backfit rule (10 CFR 50.109 or 10 
CFR 72.62) does not apply to this final rule because this amendment 
does not involve any provisions that would impose backfits as defined. 
Therefore, a backfit analysis is not required.

Congressional Review Act

    Under the Congressional Review Act of 1996, the NRC has determined 
that this action is not a major rule and has verified this 
determination with the Office of Information and Regulatory Affairs, 
Office of Management and Budget.

List of Subjects in 10 CFR Part 72

    Administrative practice and procedure, Criminal penalties, Manpower 
training programs, Nuclear materials, Occupational safety and health, 
Penalties, Radiation protection, Reporting and recordkeeping 
requirements, Security measures, Spent fuel, Whistleblowing.


0
For the reasons set out in the preamble and under the authority of the 
Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 
1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting the 
following amendments to 10 CFR part 72.

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE

0
1. The authority citation for part 72 continues to read as follows:

     Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-
486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L. 
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135, 
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148, 
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153, 
10155, 10157, 10161, 10168); sec. 1704, 112 Stat. 2750 (44 U.S.C. 
3504 note); sec. 651(e), Pub. L. 109-58, 119 Stat. 806-10 (42 U.S.C. 
2014, 2021, 2021b, 2111).
    Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), 
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 
10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat. 
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub. 
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also 
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2224 (42 U.S.C. 10101, 
10137(a), 10161(h)). Subparts K and L are also issued under sec. 
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 
(42 U.S.C. 10198).


0
2. In Sec.  72.214, Certificate of Compliance 1030 is added to read as 
follows:


Sec.  72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1030.
    Initial Certificate Effective Date: January 10, 2007.
    SAR Submitted by: Transnuclear, Inc.
    SAR Title: Final Safety Analysis Report for the 
NUHOMS[reg] HD Horizontal Modular Storage System Irradiated 
Nuclear Fuel.
    Docket Number: 72-1030.
    Certificate Expiration Date: January 11, 2027.
    Model Number: NUHOMS[reg] HD-32PTH.

    Dated at Rockville, Maryland, this 22nd day of November, 2006.

    For the Nuclear Regulatory Commission.
William F. Kane,
Acting Executive Director for Operations.
 [FR Doc. E6-20962 Filed 12-8-06; 8:45 am]
BILLING CODE 7590-01-P