[Federal Register Volume 59, Number 245 (Thursday, December 22, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-31307]


[[Page Unknown]]

[Federal Register: December 22, 1994]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AF02

 

List of Approved Spent Fuel Storage Casks: Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to add the Standardized NUHOMS Horizontal Modular System to 
the List of Approved Spent Fuel Storage Casks. This amendment allows 
the holders of power reactor operating licenses to store spent fuel in 
this approved cask under a general license.

EFFECTIVE DATE: January 23, 1995.

ADDRESSES: Copies of the environmental assessment and finding of no 
significant impact, as well as, the public comments received on the 
proposed rule are available for inspection and/or copying for a fee at 
the NRC Public Document Room, 2120 L Street, NW. (Lower Level), 
Washington, DC. Single copies of the environmental assessment and the 
finding of no significant impact are available from the individuals 
listed under the next heading below.

FOR FURTHER INFORMATION CONTACT: Mr. Gordon E. Gundersen, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, telephone (301) 415-6195, or Dr. Edward Y. S. 
Shum, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone (301) 415-7903.

SUPPLEMENTARY INFORMATION:

Background

    Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA) 
includes the following directive: ``The 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 reactor power 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.'' After 
subsequent DOE technical evaluations and based on a full review of all 
available data, the Commission approved dry storage of spent nuclear 
fuel in a final rule published in the Federal Register on July 18, 1990 
(55 FR 29181). The final rule established a new Subpart K within 10 CFR 
Part 72 entitled ``General License for Storage of Spent Fuel at Power 
Reactor Sites.''
    Irradiated reactor fuel has been handled under dry conditions since 
the mid-1940s when irradiated fuel examinations began in hot cells. 
Light-water reactor fuel has been examined dry in hot cells, since 
approximately 1960. Irradiated reactor fuel has been stored 
continuously at hot cells under dry conditions for approximately two 
decades. The NRC's experience with storage of spent fuel in dry casks 
is extensive as described in the proposed rule to establish 10 CFR Part 
72, Subpart K (May 5, 1989; 54 FR 19379). Further, the United States 
has extensive experience in the licensing and safe operation of 
independent spent fuel storage installations (ISFSIs). By the end of 
1994, six site-specific licenses for dry cask storage will have been 
issued: Virginia Power's Surry Station, issued July 2, 1986; Carolina 
Power and Light's (CP&L) HB Robinson Station, issued August 13, 1986; 
Duke Power's Oconee Station, issued January 29, 1990; Public Service of 
Colorado's Fort St. Vrain facility, issued November 4, 1991; Baltimore 
Gas and Electric's (BG&E) Calvert Cliffs Station, issued November 25, 
1992; and Northern States Power's (NSP) Prairie Island Nuclear 
Generating Plant, issued October 19, 1993. All except NSP have 
commenced operation and loaded fuel. At the end of 1994, dry storage 
spent fuel inventories of these utilities are as follows: 500 
assemblies at Virginia Power, 60 assemblies at CP&L, 530 assemblies at 
Duke Power, 1480 fuel elements at Public Service of Colorado, and 190 
assemblies at BG&E. NSP plans to begin storing fuel soon. In May 1993, 
Consumers Power's Palisades Station commenced operation and loaded fuel 
under the provisions of the general license in 10 CFR Part 72, Subpart 
K. At the end of 1994, approximately 168 assemblies are stored at 
Palisades.
    As a result of the growing use of dry storage technology, the NRC 
has gained over 35 staff years of experience in the review and 
licensing of dry spent fuel storage systems. In addition, the NRC draws 
upon the knowledge and experience of outside scientists and engineers 
recognized as experts within their respective fields in the performance 
of the independent safety analysis of the system and component designs 
submitted by applicants for dry cask licenses or certification. Reviews 
of numerous applications seeking site- specific licenses, certificates 
of compliance, or approvals of topical reports, have been conducted 
over the past eight years. More recently, the NRC published a notice of 
proposed rulemaking in the Federal Register on June 2, 1994 (59 FR 
28496), which proposed to amend 10 CFR 72.214 to include one additional 
spent fuel storage cask (i.e., the VECTRA Technologies, Inc., 
Standardized NUHOMS Horizontal Modular Storage System) on the list of 
approved spent fuel storage casks that power reactor licensees may use 
under the provisions of a general license issued by NRC in accordance 
with 10 CFR Part 72, Subpart K. The Standardized NUHOMS consists of two 
systems: (1) The NUHOMS-24P holds 24 specified pressurized-water 
reactor spent fuel assemblies and (2) The NUHOMS-52B holds 52 specified 
boiling-water reactor spent fuel assemblies.
    Subsequent to the expiration of the 75-day public comment period on 
August 16, 1994, NRC received a request, dated August 11, 1994, for a 
6-week extension of the comment period from Connie Kline of the Sierra 
Club on behalf of 12 citizen groups. The extension request asserted 
that several proprietary documents related to this rulemaking were not 
available to the public for approximately 2 weeks at the beginning of 
the comment period. The NRC granted the request on August 29, 1994 (59 
FR 44381) by extending the public comment period to September 30, 1994.
    VECTRA Technologies, Inc. (formerly Pacific Nuclear Fuel Services, 
Inc.) submitted to the NRC a Safety Analysis Report (SAR) entitled 
``Safety Analysis Report for the Standardized NUHOMS Horizontal Modular 
Storage System for Irradiated Nuclear Fuel,'' NUH-003, Revision 2, 
dated November 1993. Subsequently, VECTRA Technologies, Inc. provided 
additional information to the NRC related to the SAR. In March 1994, 
the NRC issued a draft Safety Evaluation Report (SER) entitled ``Safety 
Evaluation Report of Pacific Nuclear Fuel Services, Inc. Safety 
Analysis Report for the Standardized NUHOMS Horizontal Storage System 
for Irradiated Nuclear Fuel'' approving the SAR. The NRC issued a draft 
Certificate of Compliance by letter to Mr. Robert D. Quinn from Mr. 
Frederick C. Sturz dated April 28, 1994. These documents are part of 
the docket and record that support the proposed rule published in the 
Federal Register on June 2, 1994 (59 FR 28496).
    The objective of 10 CFR Part 72 is to protect the public health and 
safety by providing for the safe confinement of the stored fuel and 
preventing the degradation of the fuel cladding. The review criteria 
used by the NRC for review and approval of dry cask storage under 10 
CFR Part 72 consider the following factors: siting, design, quality 
assurance, emergency planning, training, and physical protection of the 
fuel. Phenomena such as earthquakes, high winds, tornados, tornado 
driven missiles, lightning, and floods are included in the review of a 
specific system, either for a certificate of compliance or a site-
specific license. In addition, applicants must demonstrate to NRC's 
satisfaction that their proposed dry cask system will resist man-made 
events such as explosions, fire, and drop or tipover accidents.\1\
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    \1\The design bases for these events and accidents are contained 
within 10 CFR Part 72.
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    Based on further NRC review and analysis of public comments, both 
the SER and Certificate of Compliance for the Standardized NUHOMS were 
modified. Section M contains a description of changes to the SER and 
Certificate of Compliance in response to public comments. The NRC finds 
that the Standardized NUHOMS, as designed and when fabricated and used 
in accordance with the conditions specified in its Certificate of 
Compliance, meets the requirements of 10 CFR Part 72. Thus, use of the 
Standardized NUHOMS, as approved by the NRC, will provide adequate 
protection of the public health and safety and the environment. With 
this final rule, the NRC is approving the use of the Standardized 
NUHOMS 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 Certificate of Compliance to 
be effective on January 23, 1995. A copy of the Certificate of 
Compliance is available for public inspection and/or copying for a fee 
at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), 
Washington, DC.

Public Responses

    In response to the proposed addition of the Standardized NUHOMS, 
239 comments in 27 letters with one supplement were received from 
individuals, public interest groups, an environmental group, an 
association, industry representatives, a city, states, and one Federal 
agency. One commenter withdrew his comments. Many of these letters 
contained similar comments that have been grouped together and 
addressed as a single issue. All comments have been grouped into 15 
broad issues designated A through O. A summary of the comments and an 
NRC analysis and response to those comments is included for each broad 
issue. The NRC has identified and responded to 89 separate issues that 
include the significant points raised by each commenter.
    A number of comments were related to the disposal of high-level 
waste and the use of dry cask storage technology in general, rather 
than to the acceptability of this particular cask. Examples of these 
comments include:

--The Federal Government's failure to resolve questions about the 
permanent storage of nuclear waste leaves both the plant and public 
with limited options: additional storage in pools, additional storage 
in dry casks, or plant shutdown. The Federal Government has an 
obligation to resolve the issue of permanent or interim storage. It 
would be difficult to overstate the need for dispatch in doing so, as 
hundreds of American communities will eventually face this problem.
--It is not fair to the public of Ohio to link Toledo Edison Company's 
attempts to continue the safe storage of its nuclear fuel with 
insistence by others that the NRC shut down Davis-Besse and every other 
nuclear plant in the country.
--Only dry storage casks that are compatible with future DOE interim or 
permanent storage operation, including transportation, should be 
approved for use under the general license and listed in 10 CFR 72.214.

    These comments deal with broad policy and program issues relating 
to the storage and disposal of high-level radioactive waste, including 
the DOE's repository program and as such are beyond the scope of this 
rule. However, there is a summary of relevant information on many of 
these broad issues in Group G. Many comments were directed at the 
Standardized NUHOMS-24P with only a few comments being specific to the 
Standardized NUHOMS-52B.
    Many commenters discussed topics that were not the subject of this 
rulemaking and thus were not specifically addressed by the NRC staff as 
a part of this final rule action. These comments express opposition to 
the use of dry cask storage and included the following suggestions and 
topics:
    (1) Nuclear plants generating radioactive waste should be shut 
down.
    (2) The production of radioactive waste should be stopped when the 
existing spent fuel pool (and off-load-reactor capacity) is full.
    (3) A formal hearing should be required at each site using dry 
storage casks.
    (4) The Davis-Besse plant should be shut down.
    (5) The use of nuclear power should be stopped and existing sites 
cleaned up.
    (6) Palisades experienced problems in using the VSC-24 cask.
    (7) Alternative forms of power should be used.
    Finally, many commenters expressed concern over the ability of dry 
cask storage designs, presumably including the Standardized NUHOMS, to 
store spent fuel safely. The following responses to these comments 
reflect a small but important portion of the NRC's review of health, 
safety, and environmental aspects of the Standardized NUHOMS to ensure 
that the cask is designed to provide protection of the public health 
and safety and environment under both normal conditions and severe, 
unlikely but credible, accident conditions. Dry cask storage systems 
are massive devices, designed and analyzed to provide shielding from 
direct exposure to radiation, to confine the spent fuel in a safe 
storage condition, and to prevent releases of radiation to the 
environment. They are designed to perform these tasks by relying on 
passive heat removal and confinement systems without moving parts and 
with minimal reliance on human intervention to safely fulfill their 
function for the term of storage. The NRC staff has concluded that the 
methods of analysis are conservative and assure that the design has 
appropriate margins of safety under both normal and accident 
conditions.

Analysis of Public Comment

    A. A number of commenters raised issues relating to cask handling 
and the ability of the cask to withstand drop and tipover accidents.
    A.1. Comment. Several commenters wanted the transfer cask 
containing the Dry Storage Canister (DSC) to be analyzed for the 
maximum possible drop, regardless of whether that drop can occur inside 
or outside the spent fuel building. One commenter alleged that a drop 
of the transfer cask into the spent fuel pool would damage fuel 
assemblies in the pool. Another commenter was concerned about jamming 
the transfer cask in the spent fuel pool. What would happen to the cask 
if jammed fuel could not be extricated? Would the entire 40 ton 
transfer cask be left in the fuel pool?
    Response. Use of the Standardized NUHOMS inside the fuel handling 
building would be conducted in accordance with the 10 CFR Part 50 
reactor operating license. These cask handling operations, including 
loading, retrieval, and training, must be evaluated by the general 
licensee as required by 10 CFR 72.212(b)(4) to ensure that procedures 
are clear and can be conducted safely. Load handling activities and 
possible load drop events with structural and radiological consequences 
related to transfer cask drops inside the spent fuel building are 
subject to the provisions of 10 CFR 50.59. Thus, the licensee must 
determine whether the activities involve any unreviewed facility safety 
question or any change in facility technical specifications. The 
transfer cask and DSC designs were evaluated by the NRC against the 
criteria for controlling heavy loads that are found in NRC's NUREG-
0612,\2\ ``Control of Heavy Loads at Nuclear Power Plants,'' and 
American National Standards Institute (ANSI) N14.6, ``Special Lifting 
Devices for Shipping Containers Weighing 10,000 Pounds or More.'' The 
lifting yoke associated with the transfer cask is a special purpose 
device designed to ANSI N14.6 criteria to ensure that the yoke can 
safely lift the wet transfer cask containing the DSC out of the spent 
fuel pool and can safely lift the dry transfer cask and DSC to the 
transport trailer. Pursuant to 10 CFR 50.59, for those reactor plants 
with a shipping cask drop analysis, the licensee must verify that the 
shipping cask drop analysis adequately describes the consequences of a 
postulated transfer cask drop and that no unreviewed safety question 
exists. For those reactor plants that may lack a shipping cask drop 
analysis, each licensee must perform a transfer cask drop analysis 
pursuant to 10 CFR 50.59.
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    \2\Copies of NUREG-0612 and NUREG/CR 1815 may be purchased from 
the Superintendent of Documents, U.S. Government Printing Office, 
Mail Stop SSOP, Washington, DC 20402-9328. Copies are also available 
from the National Technical Information Service, 5285 Port Royal 
Road, Springfield, VA 22161. A copy is also available for inspection 
and copying for a fee in the NRC Public Document Room, 2120 L 
Street, NW (Lower Level), Washington, DC 20555-0001.
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    Specific requirements for lifting the transfer cask are contained 
in the Certificate of Compliance and SER. However, movement of the 
transfer cask in the spent fuel pool building must, as required by 10 
CFR 72.212(b)(4), be evaluated by the licensee pursuant to 10 CFR 
50.59. The possibility of jamming a transfer cask while in the spent 
fuel pool is one of many issues to be evaluated under 10 CFR 50.59.
    A.2. Comment. One commenter asked why the transfer cask with the 
DSC can be lifted to 80 inches outside the spent fuel pool building 
when it has to be unloaded and inspected for damage if it drops from 
above 15 inches. Why not limit the height to 15 inches?
    Response. The transfer cask with the DSC rides on the transport 
trailer at a height of greater than 15 inches and therefore was 
analyzed for a drop from that height (80 inches). A drop from a height 
between 15 and 80 inches does not pose a public health and safety 
hazard. However, to ensure safety the NRC requires the DSC to be 
unloaded and inspected for damage.
    A.3. Comment. One commenter asked about the tipover analysis or 
drop analysis result.
    Response. The tipover, end drops, and horizontal drop analyses form 
part of the structural design basis for the Standardized NUHOMS design. 
The designer, VECTRA, described these drop and tipover analyses in SAR, 
Section 8.2.5. The NRC's evaluation of the vendor's analyses is 
described in SER, Section 3.2.2.3E. The NRC found the results of these 
analyses to be satisfactory, because the calculated stresses were all 
within the allowable criteria of the American Society of Mechanical 
Engineers (ASME) Code.
    A.4. Comment. Several commenters, citing Section 1.1.1 of the draft 
Certificate of Compliance, requested that the postulated cask drop 
accident in the plant fuel handling area be included in the list of 
parameters and analyses that will need verification by the system user 
(for the 10 CFR 50.59 safety evaluation).
    Response. As stated in Section 1.1.1 of the draft Certificate of 
Compliance, a holder of a 10 CFR Part 50 license before use of the 
general license under 10 CFR Part 72, must determine whether activities 
related to storage of spent fuel involve any unreviewed facility safety 
issues or changes in facility technical specifications as provided 
under 10 CFR 50.59. Fuel handling including the possible drop of a 
spent fuel cask is among the activities that are required to be 
verified. Fuel handling operations, including spent fuels and fresh 
fuels, are routine within the nuclear power plant and are subject to 
NRC regulation under 10 CFR Part 50. A holder of a 10 CFR Part 50 
license is required to establish operating procedures for spent fuel 
handling and to provide emergency planning to address a potential cask 
drop accident in the reactor's fuel handling area (Certificate of 
Compliance, Section 1.1.4). Therefore the NRC considers it clear that 
the spent fuel operation in the nuclear power plant should be evaluated 
to verify that the possible drop of a spent fuel cask does not raise an 
unreviewed safety issue or require a facility technical specification 
change appropriately regulated under 10 CFR Part 50.
    A.5. Comment. One commenter stated that there is no place to unload 
the spent fuel in the event of a canister breach. There is no 
indication that the canister, the canister lifting mechanism, or the 
transport mechanism to move the canister into the cask, are nuclear 
grade equipment or have been designed to prevent a single failure from 
breaching the canister and circumventing the protection provided by the 
sole barrier provided by the canister wall itself.
    Response. According to 10 CFR 72.122(1), storage systems must be 
designed to allow ready retrieval of the spent fuel in storage. A 
general licensee using an NRC-approved cask must maintain the 
capability to unload a cask. Typically, this will be done by 
maintaining the capability to unload a cask in the reactor fuel pool. 
Other options are under consideration that would permit unloading a 
cask outside the reactor pool.
    With respect to canister equipment and design, the DSC or canister 
is designed to the ASME Boiler and Pressure Vessel Code (BPVC), Section 
III, Subsection NB. The DSC provides a containment boundary for the 
radioactive material and the cladding of the fuel rods provides 
confinement of fuel pellets. Only intact fuel assemblies (rods) with no 
known cladding defects greater than pin holes and hairline cracks are 
permitted to be stored. This approach assures the structural integrity 
of the fuel to confine the fuel pellets and its retrivability. In the 
unlikely event of a breach that required the canister to be unloaded, 
the canister can be returned to the reactor spent fuel pool. Therefore, 
it is incorrect to assert that there is no place to unload a canister. 
The Horizontal Storage Module (HSM) is designed to American Concrete 
Institute (ACI) 349, which is the required code for nuclear structures 
made of reinforced concrete. The transfer cask is designed according to 
the ASME BPVC, Section III, Subsection NC; ANSI-N14.6 for heavy loads; 
ANSI-50.9 for load combinations; and NUREG/CR 1815 for impact testing. 
Because the cask itself is required to meet such exacting standards of 
construction, the transport mechanism and the trailer that move the 
canister into the HSM are not considered to be important to safety. 
Therefore, the design that meets industry standards is sufficient.
    B. A number of commenters raised issues relating to releases of 
radioactivity from surface contamination and leakage from the casks 
under normal and accident conditions.
    B.1. Comment. One commenter pointed out that the Certificate of 
Compliance Surveillance Requirement 1.2.12 does not have a section 
stating the action that is to be taken when the contamination level in 
the transfer cask exceeds limits after the DSC has been transferred to 
the concrete HSM.
    Response. The Certificate of Compliance Surveillance Requirement in 
Section 1.2.12 has been modified to clarify that decontamination of the 
transfer cask is required if the surface contamination limit is 
exceeded.
    B.2. Comment. One commenter, who was concerned with the seismic 
events at the Davis-Besse Nuclear Power Station, stated that a 
displacement pulse of 60 cm, as observed in the Lander's quake in the 
Mojave Desert northeast of Los Angeles, would completely destroy the 
HSM and allow a substantial release of radioactivity from the fuel 
within.
    Response. The potential for a seismic event is not the same at 
every reactor site in the United States. For Davis-Besse, the maximum 
ground displacement has been calculated to be 3.33 inches (8.46 cm), 
corresponding to a 0.15g maximum ground acceleration. This is 
substantially less than the displacement observed in the Lander's quake 
and appears to be well within the design of the Standardized NUHOMS. 
Each general licensee using the Standardized NUHOMS, including Davis-
Besse is required to document their evaluations to determine that the 
reactor site parameters, including seismic events, envelope the cask 
design basis, as specified in its SAR and SER.
    B.3. Comment. One commenter, citing a Wisconsin Public Service 
Commission draft environmental impact statement (EIS) for Point Beach, 
asked for an explanation of why NUHOMS and metal casks have a greater 
potential to spread contamination than the Pacific Sierra Nuclear 
Associates ventilated storage cask (VSC) system, VSC-24 cask.
    Response. The specific rationale that forms the basis of the 
statement in the Wisconsin Public Service Commission's draft EIS for 
Point Beach was not documented. The decontamination requirements for 
the two designs are comparable. The VSC-24 DSC is loaded into the 
ventilated concrete cask (VCC) forming the VSC. The VSC is then 
transported from inside the reactor auxiliary building to the storage 
pad. During moving and storage of the VCC, the exterior surface remains 
clean because it has not been exposed to contamination in the spent 
fuel pool. The NUHOMS DSC is moved in the transfer cask from the 
reactor building to the horizontal storage module in the field. Because 
the transfer cask has been in the spent fuel pool, it may have small 
amounts of external contamination that have the potential to spread 
during transit. However, any potential contamination of this type could 
not be significant. The NRC requires that the limits for surface 
contamination, workers' dose, and environmental dose must all be met 
for the operation of the ISFSI, including during any transfer 
operations. Each 10 CFR Part 50 licensee must have a radiation 
protection program to monitor operations to ensure that surface 
contamination and worker and public exposure to radiation are below 
acceptable levels and as low as is reasonably achievable (ALARA). Past 
operation of the NUHOMS shows that the doses are well below all NRC 
limits.
    B.4. Comment. One commenter ``is concerned that heat generated by 
fission product decay may provide the driving force, the presence of 
free moisture in water-logged fuel may, in a non-mechanistic way, 
provide a transport mechanism for fission product release and the 
ambient air circulating through the cask concrete structure may provide 
(an unmonitored) pathway to the biosphere.'' One commenter remained 
concerned about the possibility of insufficient drying of the fuel 
before placement in the DSC. Another commenter, citing the Battelle 
Pacific Northwest Laboratory Report PNL-5987 on the removal of moisture 
from degraded fuel during vacuum drying, contends that the mechanism 
for free moisture and radionuclide release that pertain in normal or 
upset conditions, such as conditions caused by sabotage, have not been 
simulated adequately.
    Response. The DSC is a closed vessel. There is no path available 
for release of fission products from inside the DSC to the atmosphere. 
During normal operation, the circulating air, as it passes through the 
HSM and around the outside of the DSC to remove the heat, never comes 
in contact with fission products and therefore, could not remove these 
products from the cavity of the DSC. Moreover, design basis accidents 
under upset conditions were postulated and analyzed in the SAR and SER. 
These analyses show that the heat generated from fission product decay 
is not capable of breaching the DSC and could not provide the driving 
force for a release of radioactivity. Further, it is not expected that 
any significant amount of moisture will remain in the fuel after it is 
loaded into the DSC. The fuel is dried after it has been loaded into 
the DSC and the topcover plate seal welded to the DSC shell. The 
Certificate of Compliance requires two pump-downs to a vacuum pressure 
of less than 3 mm Hg each with a holding time of greater than 30 
minutes. A stable vacuum pressure of less than 3 mm Hg further assures 
that all liquid water has evaporated in the DSC cavity.
    The safeguards issue of radiological sabotage of storage casks has 
been reviewed previously and assessed in the 1989 proposed rule to add 
Subparts K and L to 10 CFR Part 72 (54 FR 19379). The NRC has 
determined that the Standardized NUHOMS is sufficiently robust such 
that the effects of a successful attack would have low health 
consequences and are similar to the results presented in the 1989 
proposed rule. (see also response to comment N.1)
    C. A number of comments were received that focused on monitoring, 
surveillance, and inspection activities associated with dry cask 
storage of spent fuel, particularly as they relate to the Standardized 
NUHOMS.
    C.1. Comment. One commenter stated that there are neither active 
nor passive systems in place to mitigate barrier breaches, nor are 
there active radiation monitors that would indicate a breach has 
occurred. There are no monitored drains and sumps nor are there 
retention basins. The commenter stated that the cask is insufficient to 
be relied upon for the health and safety of Ohioans.
    Response. The Certificate of Compliance (Section 1.3) for the 
Standardized NUHOMS includes surveillance and monitoring requirements 
that are more than sufficient to detect cask degradation in time to 
ensure that adequate corrective actions can and will be taken. In 
addition, radiation monitoring and environmental monitoring programs 
would detect any radiation leak in excess of NRC limits from an NRC-
approved cask.
    In some instances, the NRC has required continuous monitoring where 
it is needed to determine when corrective action needs to be taken. 
Under a general license, to date, the NRC has accepted continuous 
pressure monitoring of the inert helium atmosphere as an indicator of 
acceptable performance of mechanical closure seals for dry spent fuel 
storage casks.
    However, the NRC does not consider continuous monitoring for the 
Standardized NUHOMS double-weld seals to be necessary because:
    (1) There are no known long-term degradation mechanisms which would 
cause the seal to fail within the design life of the DSC; and
    (2) The possibility of corrosion has been included in the design 
(see SER Section 3.2.2.5).
    These conditions ensure that the internal helium atmosphere will 
remain stable. Therefore, an individual continuous monitoring device 
for each HSM is not necessary. However, the NRC considers that other 
forms of monitoring, including periodic surveillance, inspection and 
survey requirements, and application of preexisting radiological 
environmental monitoring programs of 10 CFR Part 50 during the use of 
the canisters with seal weld closures can adequately satisfy NRC 
requirements.
    With respect to the use of instrumentation and control systems to 
monitor systems that are important to safety, the user of the 
Standardized NUHOMS will, as provided in Chapter 14 of the SER and in 
Section 1.3.2 of the Certificate of Compliance, be required to verify, 
the cask thermal performance on a daily basis by a temperature 
measurement, to identify conditions that threaten to approach cask 
design temperature criteria. The cask user will also be required to 
conduct a daily visual surveillance of the cask air inlets and outlets 
as required by Chapter 12 of the SER and Section 1.3.1 of the 
Certificate of Compliance.
    While the HSM and DSC are considered components important to 
safety, they are not considered operating systems in the same sense as 
spent fuel pool cooling water systems or ventilation systems that may 
require other instrumentation and control systems to ensure proper 
functioning. Due to this passive design, temperature monitoring and 
surveillance activities are appropriate and sufficient to assure 
adequate protection of the public health and safety for this design.
    Because the Standardized NUHOMS DSC is welded closed and has been 
decontaminated before being placed in a HSM, there is no routine 
radioactive liquid generation that would require a retention basin or 
sump. Water entering the storage area has no mechanism of becoming 
contaminated because the DSC is enclosed within the HSM and is expected 
to be dried by the heat generated during storage.
    C.2. Comment. One commenter expressed concern over the possible 
external corrosion of the stainless steel DSC because of exposure to 
water over decades. Another commenter expressed concern about corrosion 
of stainless steel under conditions of indefinite duration, stating 
that while stainless steel corrodes less rapidly than carbon steel, 
even the plumbing fixture industry is finding unexpected stainless 
steel pitting and corrosion under conditions far less intense than 
those in a DSC. Another commenter stated that the system is not 
designed for remote inspection of the DSC for corrosion while it is in 
the HSM and that the only way to inspect the DSC is to return it to the 
spent fuel pool. Periodic inspection of the DSC is needed to preclude 
or identify gradual canister deterioration by unknown mechanisms. 
Another commenter inquired about a checking system for the NUHOMS in 
the future. How will corrosion be evaluated on the canister (DSC) and 
the support rails inside the HSM? Is it possible for them to accumulate 
moisture and corrode together over possibly many years of storage? What 
check is required on the possibility that the canister couldn't be 
removed at the end of cask life?
    Response. The DSC is enclosed within the HSM and is not exposed to 
external water. Laboratory experiments have indicated a general 
corrosion rate of less than 0.00001 inches per year for similar 
stainless steels. The NRC believes these experiments more accurately 
bound DSC corrosion than experiences in unrelated industries. For the 
50-year design life of the DSC, the expected corrosion would therefore 
not result in exceeding a corrosion depth of 0.0005 inches. This will 
not affect the DSC performing its intended safety functions. Because of 
the low corrosion rates expected for stainless steel, periodic 
inspections for deterioration of the DSC are not considered necessary. 
Therefore, inspections are not required. The support rails for the DSC 
have an extremely hard-alloy steel applied to the sliding surface, are 
ground to a smooth finish, and are coated with a dry film lubricant to 
prevent corrosion and to reduce the coefficient of friction. 
Furthermore, the environment inside the HSM is protected from rain and 
it is kept dry by the heat load from the DSC. Therefore, it is highly 
unlikely that corrosion between the stainless steel and the hard alloy 
steel surface of the support rail will occur to any significant extent. 
These conclusions and analyses regarding the very small likelihood of 
corrosion indicate that there is reasonable assurance that the DSC can 
be removed from the HSM when required.
    C.3. Comment. One commenter questioned whether the screens between 
the casks, which are essential to cooling, will remain clear of debris 
and how they can be cleaned if they become partially clogged. Another 
commenter was concerned about how the roof screen was inspected, 
stating that it seems likely that insects, animals, and birds will be 
attracted to the warm air coming from the outlet vents. Several 
commenters remained concerned about vent blockage that can completely 
cover and block screening and vents particularly from insects such as 
paper wasps, that build huge nests, and swarms of midges that are 
common to the Great Lakes. How are the screens attached?
    Response. As stated in the Certificate of Compliance, a licensee 
using the Standardized NUHOMS must conduct a daily visual surveillance 
of the exterior of air inlets and outlets (front wall and roof bird 
screen). In addition, the licensee must perform a daily close-up 
inspection to ensure that no material accumulates between the modules 
to block the air flow. If the surveillance shows blockage of air vents, 
the licensee is required to clear the vent blockage by following 
procedures developed by each user of the Standardized NUHOMS. If the 
screen is damaged, the licensee must replace the screen. The required 
daily surveillance and temperature measurements should readily detect 
blockage of the vents or screens by insects, animals, or birds in a 
timely manner, leading to the removal of the obstruction before damage 
occurs from high temperatures. The bird screen is made of stainless 
steel wire cloth tack-welded to stainless steel strips, which are 
attached to the HSM with stainless steel wedge anchors.
    C.4. Comment. One commenter expressed concern about the presence of 
burrowing and other nuisance animals that have posed problems at other 
waste sites.
    Response. Burrowing and other nuisance animals are not expected to 
pose problems for the Standardized NUHOMS. Because of the robust system 
design, animals will not be able to get to the radioactive material or 
cause damage such that water could cause movement of the radioactive 
material. Burrowing under the concrete pad would not cause damage to 
safety-related components. Further, large-scale burrowing would likely 
be detected by the daily surveillance or other activities related to 
the operation of the storage area.
    C.5. Comment. One commenter wanted additional radiation monitoring 
because of the calculated higher dose rates over previous NUHOMS 
designs. The commenter stated that these higher dose rates are not 
consistent with the objective of maintaining occupational exposures 
ALARA, and that site-specific applications should provide detailed 
procedures and plans to meet ALARA guidelines and 10 CFR Part 20 
requirements with respect to operation and maintenance.
    Response. No additional radiation monitoring has been specifically 
identified or required for the Standardized NUHOMS. However, 10 CFR 
Parts 20, 50, and 72 require that licensees comply with ALARA. In 
addition, 10 CFR 72.212(b)(6) requires each licensee to review its 
radiation protection program to determine that their effectiveness is 
not decreased by use of the Standardized NUHOMS. Further, 10 CFR 72.212 
(b)(9) requires each licensee to conduct storage activities in 
accordance with appropriate written procedures. If the results of these 
licensee activities indicate that additional procedures are required 
then the licensee is required to implement the procedures.
    C.6. Comment. One commenter was concerned about the optical survey 
equipment used to align the transfer cask with the HSM before transfer. 
What checks are made on this optical equipment and what regulations 
apply?
    Response. The optical equipment used to align the transfer cask 
with the HSM is optional and is an operational convenience. However, 
the licensee must meet Technical Specifications 1.2.9 in the 
Certificate of Compliance. Therefore, only appropriate calibrations or 
checks to assure compliance with this technical specification are 
appropriate.
    C.7. Comment. One commenter wants to know who evaluates the 
insertion or retrieval of the DSC for excessive vibration and what is 
the result of excessive vibration. Would this allow crud to be 
released?
    Response. The NRC Certificate of Compliance, Section 1.2.9 provides 
that the cask user observe the transfer system during DSC insertion or 
retrieval to ensure that motion or excessive vibration does not occur. 
It also prescribes certain follow-up actions to be taken by the cask 
user in the event that alignment tolerances are exceeded and excessive 
vibration occurs. It is possible that excessive vibration could 
dislodge crud. However, the crud would be contained within the DSC and 
would not be released to the atmosphere because the DSC is a sealed 
vessel. Any opening of the DSC will be under controlled conditions that 
should safely contain the crud and prevent its release to the 
environment.
    C.8. Comment. Several commenters wanted the NRC to set definite 
methods for the required surveillance and monitoring of NUHOMS, 
including the daily temperature measurements, so that data are uniform 
and standardized for future reference on different modules at different 
reactor locations.
    Response. The NRC Certificate of Compliance for the Standardized 
NUHOMS has required temperature measurements. However, the licensee or 
vendor has latitude in determining how the performance-based 
temperature requirements will be met. The NRC is not convinced that the 
possible benefits of a uniform, but prescriptive, surveillance and 
monitoring system or technique would outweigh the costs of curtailing 
the freedom of cask users to design an implementation scheme suited to 
their individual needs. The collection of uniform data for possible 
future use, but without a specific regulatory need could lead to 
additional exposure to workers, or adversely affect safety without any 
offsetting benefit.
    C.9. Comment. One commenter asked about the design life of this 
NUHOMS module and on how this is documented. Will the canister be 
removed from the concrete module at a specific time and be opened?
    Response. The design life of the Standardized NUHOMS is 50 years as 
described in the SAR. The Certificate of Compliance has a 20-year 
approval period that can be renewed by NRC for another 20 years 
following a safety reevaluation. It is expected, that at the end of 
operation, the canister will be removed from the concrete module and 
will be opened in the spent fuel pool facility or an adequate dry 
environment alternative. The fuel will be transferred to an NRC-
approved shipping cask for off-site transportation and ultimate 
disposal by the DOE.
    C.10. Comment. One commenter believed it prudent to monitor 
temperature and air flow to ensure that temperature excursions are not 
experienced.
    Response. NRC believes the required temperature measurement stated 
in Specification 1.3.2 of the Certificate of Compliance, plus the daily 
visual inspection of HSM air inlets and outlets, are adequate to ensure 
that temperature excursions exceeding the design basis are not 
experienced and to determine when corrective action needs to be taken 
to maintain safe storage conditions. Therefore, air flow measurements 
are not required to assure safety.
    D. A number of commenters raised technical issues related to the 
thermal analysis of the Standardized NUHOMS and thermal performance of 
the system under normal, off-normal, and accident conditions.
    D.1. Comment. Several commenters wanted, in the interest of ALARA 
principles, the capacity for approximately 24 kW heat removal to be 
verified by using an artificial heat load. One commenter suggested that 
the NUHOMS be tested with a full heat load at a testing site such as 
Idaho National Engineering Laboratory (INEL), and not at each reactor 
site that may load it with a higher heat generation rate fuel. Another 
commenter cited the ALARA philosophy of loading the oldest fuel first 
even though design basis fuel is on site. Several commenters wanted 
deletion of the requirement (a literal interpretation of draft 
Certificate of Compliance) to calculate the temperature rise for each 
HSM loaded with canisters producing less than the design limit of 24 kW 
for the following reasons:
    (1) Users are not normally provided the vendor's analytical models 
for this calculation,
    (2) The 100  deg.F rise calculated for the design basis maximum 
heat load ensures that all safety limits are met for concrete and fuel,
    (3) Because 24 kW is the limit, virtually all the HSMs will be 
affected, which places an undue burden on the user to ``baseline'' the 
predicted delta-T by calculation considering the inherent safety 
margins of the system, and
    (4) Technical Specification 1.3.1 ensures that air flow is not 
blocked so a false measurement of low temperature rise cannot occur.
    Response. A licensee is not required by NRC to load the oldest fuel 
first but, in the interest of ALARA, it may do so. However, each time 
hotter fuel is loaded up to the maximum allowed in a DSC, the licensee 
would need to verify the heat removal performance of the system. For 
fuel producing less heat than the design limits of the system, the heat 
removal capacity of the system determined by calculation must be 
verified by temperature measurements. This process must be repeated 
each time a DSC is loaded with hotter fuel until the maximum-system 
designed heat load is reached. When loaded with spent fuel producing 24 
kW heat, the system may not have an ambient and vent outlet temperature 
difference of more than 100  deg.F for fuel cooled equal to or more 
than 5 years. This verification process is required to confirm that the 
as-built system of each licensee is performing as designed. A licensee 
could use an artificial heat source to test an initial cask at a 
bounding heat load of 24 kW before loading fuel. However, this test 
would only verify the spent fuel heat removal capacity of the system. 
It would not verify as-built performance. Experience has shown that 
adequate verification testing can be performed at the reactor site. 
Therefore, performing the verification at a testing site like INEL 
would not provide additional safety margins.
    D.2. Comment. Several commenters pointed out possible conflicting 
statements about temperature measurements in the surveillance 
requirements. In discussions about the heat removal capacity test, 
temperatures are determined only during the test period. Daily 
temperature measurements on each HSM are required to verify thermal 
performance.
    Response. These two temperature measurement programs have different 
objectives. Temperature measurements by licensees to verify the heat 
capacity calculations need only be done until equilibrium is reached. 
The daily temperature measurements by licensees are intended to 
demonstrate continued safe operation within specified limits over the 
life of the HSM and may not be the same type of measurement done in the 
initial period to verify heat removal capacity.
    D.3. Comment. One commenter was concerned about the adequacy of 
cooling under all atmospheric conditions in the country. The commenter 
cited conditions such as humidity over 90 percent, temperature over 100 
 deg.F, and no wind.
    Response. Regulatory requirements for general licensee users of dry 
storage casks are contained in 10 CFR 72.212(b). Each user must verify 
that the following conditions are not exceeded at their reactor site 
for the Standardized NUHOMS: the maximum average yearly temperature 
with solar incidence is 70  deg.F; the average daily temperature is 100 
 deg.F; and the maximum temperature is 125  deg.F with incident solar 
radiation. If the power reactor site high temperature parameters fall 
within these criteria, the Standardized NUHOMS can be safely used at 
the site.
    D.4. Comment. One commenter wants the NRC to establish procedures 
to measure temperature performance, especially the thermal performance 
of an individual module and not the combined performance of adjacent 
modules as stated on page A-23 of the draft Certificate of Compliance.
    Response. As required by the regulations, the licensees are 
required to develop detailed procedures. NRC in its regulatory oversite 
role has the opportunity to review the adequacy of the procedures. The 
requirement cited by the commenter is a requirement for the licensee to 
verify a temperature measurement of the thermal performance for each 
HSM, not the combined performance of adjacent modules. A cautionary 
statement is included in the basis of the specification to ensure that 
licensee measurements of air temperatures reflect only the thermal 
performance of an individual module and not the combined performance of 
adjacent modules.
    D.5. Comment. One commenter wanted to know how the temperature 
differences in the roof, side wall, and floor areas are incorporated 
into the daily temperature measurement.
    Response. For the first HSM to be emplaced, the user is required to 
measure the air inlet and air outlet temperature difference of the 
system at equilibrium. This measurement is to ensure that the heat 
capacity of the system will not be exceeded and that the concrete 
temperature criteria will not be exceeded. For the Standardized NUHOMS, 
this maximum heat capacity is 24 kW. The 24 kW heat load is the design 
maximum and is the basis for the thermal hydraulic calculations for the 
cask. The temperature distribution for various parts of the HSM have 
been calculated (i.e., the roof, walls, and floor) by the cask vendor. 
Temperature differences causing thermal stresses in the concrete were 
evaluated and are duly reported in both the SAR and SER. These 
calculations were reviewed by NRC as a part of the overall process for 
this design approval.
    D.6. Comment. One commenter stated that daily temperature 
measurements are not necessary to ensure convective air flow, given the 
requirement to verify that the inlets and outlets are not obstructed. 
Site-specific NUHOMS require temperature measurements when the DSC is 
placed into the HSM, 24 hours later, and again at 1 week after loading 
to ensure adequate thermal performance.
    Response. The NRC disagrees with this comment. The HSM and DSC are 
considered components important to safety in the Standardized NUHOMS. 
Daily temperature measurements of the thermal performance by the 
licensee are required to provide additional assurance that thermal 
limits are not exceeded under the general license. This requirement was 
imposed on the first cask of this type approved by the NRC and listed 
in 10 CFR 72.214 for use by a general licensee, the VSC-24 cask (58 FR 
17967; April 7, 1994) and is now applied to the Standardized NUHOMS.
    E. A number of commenters expressed concern about emergency 
planning and response contingencies.
    E.1. Comment. Several commenters expressed concern that in the 
event of problems and the need to off-load fuel (as in the recent 
situation at Palisades), a transfer cask may not be available in a 
timely manner because of inclement weather or because the transfer cask 
itself has experienced problems or is being used elsewhere. One 
commenter expressed concern at having to have a transfer cask on site 
within 40 hours of vent blockage to prevent concrete damage. If the 
transfer cask is leased from VECTRA and is not at the licensee's site, 
who is liable if something happens that would require the use of a 
transfer cask?
    Response. The NRC has analyzed all design basis accidents from the 
operation of an ISFSI and concluded that there will be no release of 
radioactive material to the environment. The 40-hour limit on vent 
blockage is intended to prevent concrete degradation that might occur 
over a long period of storage. A vent blockage accident would not 
result in the release of radioactive material because the DSC would not 
be breached. Therefore, the NRC believes that the potential risk to the 
public health and safety is extremely small during the time needed to 
obtain the use of a transfer cask. Thus, there is no requirement that a 
transfer cask be at an ISFSI site all the time.
    E.2. Comment. One commenter expressed concern that the effects of 
tornado winds and missiles during movement of the fuel in a transfer 
cask or in a storage cask on a transporter were not analyzed.
    Response. Both the vendor's SAR and NRC staff's SER address the 
effects of tornado winds and missiles during movement of the transfer 
cask with a loaded canister. These analyses show that, for tornado 
winds, there is a safety factor of 1.5 against overturning when 
subjected to Design Basis Tornado winds (a safety factor greater than 1 
will generally be adequate for public protection). The transfer cask 
stability, tornado missile penetration resistance, and shell and end 
plate stresses were calculated and shown to be below the allowable 
stresses for ASME BPVC Service Level D (accident) stresses.
    E.3. Comment. One commenter described an October 1972 storm that 
flooded the entire Davis-Besse plant site, including the (pre-
operational) reactor building. There has been subsequent flooding of 
the site, particularly during spring thaws.
    Response. Safety analyses by NRC and the cask vendor show the 
Standardized NUHOMS can withstand floods and will continue to perform 
acceptably. With regard to the Davis-Besse site, the licensee changed 
site topography during plant construction. Specifically, the area was 
built up and some dikes were added. The plant structure's ground floor 
elevation is 585 feet International Great Lakes Datum (IGLD), which is 
also the elevation of the pad. The licensing design basis for maximum 
probable static water level on the site is 583.7 feet IGLD. As noted, 
the HSM and DSC were evaluated for flood conditions as required by 10 
CFR 72.122(b). The HSM can withstand a maximum water velocity of 15 
feet per second and a static head of 50 feet of water. The DSC can 
withstand a static head of 50 feet of water. Any site that intends to 
use a Standardized NUHOMS design must evaluate the conditions at their 
site to verify compatibility with the design specifications of the 
system.
    F. A number of commenters raised issues relating to the design, 
evaluation, and operation of the Standardized NUHOMS.
    F.1 Comment. Several comments related to the fuel to be stored in 
the Standardized NUHOMS. One commenter wanted control components 
contained in assemblies addressed in the SAR and SER citing DOE 
acceptance criteria. One commenter questioned how 55,000 MWD/MTU burnup 
fuel now being used in pressurized water reactors will be handled since 
the Standardized NUHOMS-24 is rated to handle only 40,000 MWD/MTU 
burnup fuel. Another commenter, citing provisions of current site-
specific licensees for other NUHOMS designs, stated that higher burnup 
should be allowed if the decay heat and radiological source terms are 
within limits. Another commenter asserted that increased fission 
products from higher enriched fuel may potentially increase 
embrittlement of the fuel cladding and that this needs to be evaluated 
in the SER. This commenter further alleged that this would increase the 
probability of more defective fuel being loaded into dry casks.
    Response. The vendor designed the cask system for storage of 
pressurized water or boiling water reactor fuel assemblies meeting 
certain specifications. By limiting the use of the cask system to 
assemblies meeting these specifications, the vendor made a decision 
that may partially restrict the use of the cask. However, the NRC does 
not require that a cask be universal for all types of fuel or be usable 
at every reactor site. For example, none of the casks previously listed 
in 10 CFR 72.214 is usable for boiling water reactor spent fuel.
    Currently, the 55,000 MWD/MTU burnup fuel and fuel with initial 
enrichments of greater than 4% will have to remain in the spent fuel 
pool because dry spent fuel cask designs to store fuel with this higher 
burnup and initial enrichment or related to DOE acceptance criteria 
have not yet been reviewed and evaluated by the NRC.
    F.2 Comment. Several comments were related to criticality safety 
analysis. One commenter questioned the conservatism of using 7.5-year 
cooled spent fuel when 5-year-cooled fuel is the minimum specified and 
when older fuel may also be stored in the cask. Another inquired about 
criticality safety if the original basket geometry were compromised, as 
might be the case for brittle failure of a spacer disk. In the 
compromised basket geometry case, the commenter also asked about the 
difference in criticality safety for a helium atmosphere rather than a 
borated water medium. The commenter, referring to July 24, 1992, 
meeting minutes, inquired why all parties agreed not to spend any 
resources to make these criticality safety calculations.
    Response. The Standardized NUHOMS nuclear criticality safety 
analysis is based on the following: (1) Babcock and Wilcox 15  x  15/
208 pin fuel assemblies with initial enrichments up to 4.0 wt% of U-235 
and (2) General Electric 7  x  7 fuel assemblies with initial 
enrichments up to 4.0 wt% of U-235, for the Standardized NUHOMS-24P and 
NUHOMS-52B designs respectively. The age of the fuel that will actually 
be stored is not relevant in criticality safety analysis because the 
analysis assumes storage of unirradiated fresh fuel that is more 
reactive than cooled spent fuel. The Standardized NUHOMS-24P system has 
administrative controls that limit the irradiated fuel reactivity to 
less than or equal to 1.45 wt% of U-235 equivalent unirradiated fuel 
(Certificate of Compliance Section 1.2.1).
    The possibility of a criticality accident caused by the brittle 
failure of the basket should not be a significant concern. No lifting 
or handling of the DSC outside the spent fuel pool building is 
permitted if the basket temperature is lower than 0 deg.F. If the user 
does not determine the actual basket temperature, the ambient 
temperature must be used conservatively. Under these temperature 
restrictions, the basket materials will not behave in a brittle 
fashion. Consequently, the basket geometry would not be compromised by 
brittle failure. As for the criticality safety consideration related to 
a helium atmosphere versus a borated water medium, the keff of the 
fuel in a helium atmosphere is much less than the keff in borated 
water. Therefore, criticality calculations for the borated water are 
sufficient because they are more conservative and therefore would bound 
calculations using a helium atmosphere.
    F.3 Comment. Two commenters were concerned with shielding and dose 
assessments for the Standardized NUHOMS. One commenter believed that 
using 10-year-cooled fuel for the dose assessment was nonconservative 
when 5-year-cooled fuel is needed to load the DSC to produce 24 kW of 
heat. Another, referring to an NRC meeting with Pacific Nuclear Fuel 
Services, Inc. (PNFSI), wanted clarification of an NRC request to 
delete a clause allowing the utility to perform site-specific shielding 
calculations.
    Response. The cask vendor presented dose assessment results in the 
SAR for both 5- and 10-year-cooled fuel. However, for this rulemaking, 
NRC used the dose assessment for 5-year-cooled fuel for the shielding 
analysis radiation source term and for accidental releases of 
radionuclide material. NRC's use of the 5-year-cooled fuel assessment 
is conservative and bounding.
    To ensure safe storage of spent nuclear fuel in NRC-approved casks, 
the NRC specifies, in Section 1.2.1 of the Certificate of Compliance a 
number of fuel acceptance parameters. These parameters, which may 
include burnup, initial enrichment, heat load, cooling time, and 
radiological source term, define the properties of those assemblies 
that can be stored in a cask. One such parameter of interest for the 
Standardized NUHOMS is the radiological source term that forms the 
basis of the shielding analyses. For this parameter, the vendor 
proposed an alternative approach. Specifically, for fuel assemblies 
that fall outside the specified source term parameters but satisfy all 
other parameters, the vendor proposed to allow licensees to do 
individual cask shielding calculations to show compliance with the 
design basis dose rates. This could result in more assemblies in a 
licensee's inventory that would be eligible for dry storage. In the 
instance noted in the comment, the NRC did not agree with the vendor 
proposal. The Certificate of Compliance dose rate specifications 
provide a simple check to ensure that DSCs are not inadvertently loaded 
with the wrong fuel. The dose rate specifications are based on the 
shielding analyses provided by the vendor in its SAR. Because of 
differences in non-fuel components in the ends of some assemblies, dose 
rates higher than those evaluated by NRC in the SER may occur at the 
ends of casks than were assumed in the shielding analysis. The 
Certificate of Compliance specifications allow for this possibility and 
permit the licensee to store such fuel provided the licensee verifies 
proper cask fabrication, conformance with all other fuel parameters, 
and compliance with radiation protection requirements. The site-
specific calculations referred to in the comment are not shielding 
calculations, but rather are the licensee's written evaluations (or 
dose assessments) required by 10 CFR 72.212(b)(2)(iii) to establish 
that the radiation criteria for ISFSI in 10 CFR 72.104 have been met. 
The Certificate of Compliance also requires that the licensee submit a 
letter report to the NRC summarizing its actions in this type of case.
    F.4. Comment. Several commenters were concerned with fuel clad 
integrity issues. Particularly, they were concerned with potential 
problems that may arise because of differences between vertical and 
horizontal storage. One commenter noted that it was essential to 
inspect the cladding carefully for the minute hairline cracks which 
would allow the radioactivity inside to escape. Another commenter 
wanted it made clear that for fuel to be eligible for storage it 
doesn't need to be specifically inspected nor require special handling 
or storage provisions within the spent fuel pool. The commenter also 
asserted that pinhole leaks in fuel rod cladding do not constitute 
gross breaches. The commenter wanted fuel cladding integrity clarified. 
Another commenter claimed that horizontal storage of fuel rods will 
lead to cladding deterioration that would challenge the technical 
specifications of the NUHOMS cask. Another commenter was concerned 
about the possibility of fuel rod bowing that could result in weighted 
contact between the fuel cladding/crud and the DSC guide sleeve, with 
the potential for eventual bonding of the materials over the duration 
of the storage period. One commenter, noting that some of the fuel in 
the spent fuel pools could be nearly 20 years old, was concerned that 
the fuel will not be tested for leaks using specific techniques such as 
penetrating dyes, eddy current, sipping, or ultrasound before canister 
loading. A commenter wanted all fuel with known defects and all water-
logged fuel retained in the spent fuel pool until the cask integrity 
under operating conditions is fully demonstrated. Another wanted to 
know how ``grossly breached'' fuel will be ultimately handled and 
shipped off site.
    Response. In the Standardized NUHOMS, PWR fuel rods are stored in a 
horizontal orientation and do not normally deflect in the middle of any 
span so that the rods contact the DSC guide sleeve. However, the 
possibility exists that a bowed rod may come in contact with the guide 
sleeve.
    With respect to storage of BWR fuel, the fuel channel that 
surrounds the fuel bundle (rods) provides a barrier to separate coolant 
flow paths, to guide the control rod, and to provide rigidity and 
protection for the fuel bundle during handling. Therefore, the BWR fuel 
rods inside the channel do not come in contact with the guide sleeves. 
Even if there were contact with either PWR or BWR fuel rods, the 
interaction would not present a significant concern because the guide 
sleeve material is stainless steel, which has a very low rate of 
corrosion, and the DSC cavity is evacuated and back-filled with inert 
helium, which further reduces the likelihood of any corrosion or 
bonding involving the guide sleeve and fuel rods.
    The Certificate of Compliance requires that the fuel have no known 
or suspected gross cladding breaches to ensure the structural integrity 
of the fuel. Known or suspected failed fuel assemblies (rods) and fuel 
with cladding defects greater than pin holes and hairline cracks are 
not authorized in the Standardized NUHOMS. Fuel meeting this 
specification will be safely stored and will remain intact in storage 
because the dry inert atmosphere and relatively low temperature will 
prevent deterioration of the cladding. Grossly breached fuel will be 
handled in site-specific license applications.
    F.5. Comment. Quite a few comments related to the structural 
stability of the HSM, particularly its response to earthquakes. 
Commenters questioned the possibility of vertical storage of the 
Standardized NUHOMS and suggested that it would be very difficult to 
restrain the HSM if the DSC were in a vertical position. One commenter 
wanted dry storage casks constructed to Building Officials Code 
Administrators (BOCA) National Building Code (and Ohio Administrative 
Code) for structures in use group H-4, high hazard use, which includes 
radioactive materials. Commenters questioned whether ground 
acceleration as used by the NRC in its evaluation could adequately 
describe all potential earthquakes east of the Rocky Mountain Front and 
suggested that a ground acceleration of 2.5g would not be realistic for 
all sites, despite proximity to fault lines. Another commenter alleged 
a number of seismic events in the midwest which had some effect in the 
Ohio area could cause a complete failure of the cask and requested that 
the NRC insist that the cask, containment structure, and foundation pad 
be designed to substantially exceed all earthquakes with a potential 
for 0.60g. One commenter wanted to know if the module had been analyzed 
for earthquake events at all United States reactor sites, according to 
Laurand Findmun Seismic Hazard Curves. Other commenters expressed 
various concerns about the integrity and reaction of the Standardized 
NUHOMS components under earthquake conditions and asked the following 
questions:
    Could the casks crash against each other as the ground moves 
beneath them?
    Could the module shift, crack, or move off the pad?
    How are the rail support holdings evaluated?
    Could the DSC be knocked off the rails? and
    Could the module roof crack and fall on the canister?
    Response. The Standardized NUHOMS design described in the vendor's 
applications for approval and the SAR does not address vertical 
storage. Consequently, NRC neither evaluated nor approved vertical 
storage for the system. Therefore, it may not be stored vertically.
    The NRC reviewed the Standardized NUHOMS for compliance with design 
criteria that are more stringent than those of the BOCA National 
Building Code (NBC) (see response to Comment A.5). These more stringent 
criteria are included in national standards that more closely represent 
the use of the Standardized NUHOMS.
    Part 72 specifies a design basis maximum ground acceleration of 
0.25g for areas east of the Rocky Mountain Front that are not in areas 
of known seismic activity. All HSMs and DSCs are designed to withstand 
a 0.25g earthquake. Any reactor licensee who intends to use the 
Standardized NUHOMS must verify that the maximum displacements at the 
cask's location on the reactor site are within the design criteria for 
the system. The Standardized NUHOMS is free standing and not dependent 
on the pad for safety. Failure of the pad caused by seismic events will 
not cause the Standardized NUHOMS to fail. Therefore, cask safety does 
not require the pad to be designed to withstand a seismic event.
    F.6. Comment. One commenter stated that the SAR did not include 
consideration of the accident events such as: aircraft crashes, turbine 
missiles, external fires, explosions, and sabotage.
    Response. Before using the Standardized NUHOMS, the general 
licensee must evaluate them to ensure the site is encompassed by the 
design bases of the approved cask. The events listed in the comment are 
among the site-specific considerations that must be evaluated.
    The site evaluation for a nuclear plant considers the effects of 
nearby transportation and military activities. It is incumbent upon the 
user of the cask to determine if the SER for the facility encompasses 
the design basis analysis performed for the Standardized NUHOMS or any 
certified cask. The great majority of the aircraft are single-engine 
propeller airplanes which typically weigh on the order of 1,500 to 
2,000 pounds. The cask's inherent design will withstand tornado 
missiles and other design loads and also provides protection from the 
collision forces imposed by these light general aviation aircraft 
without adverse consequences. NUREG-800, Section 3.5.1.6 ``Standard 
Review Plan for Light Water Reactors,'' contains methods and acceptance 
criteria for determining if the probability of an accident involving 
larger aircraft (both Military and civilian) exceeds the acceptable 
criterion. It is incumbent upon the licensee to determine whether or 
not the reactor site parameters are enveloped by the cask design basis 
as required by 10 CFR 72.212(b)(3). These would include an evaluation 
demonstrating that the requirements of 10 CFR 72.106 have been met.
    Turbine missile analyses typically show a very low probability of a 
turbine missile breaking the turbine casing. The site's turbine missile 
analyses must be considered as part of the facility's analysis of the 
suitability of the storage location. External fires are handled by 
established fire control programs. Explosions are prevented by control 
of combustibles under the licensee's fire protection program. Sabotage 
is considered under the criteria for security programs that each 
licensee must implement. (See also response to comment N.1).
    F.7. Comment. Several commenters raised issues about the pad and 
foundation for the Standardized NUHOMS. One commenter referred to a 
previous rulemaking that stated that the NUHOMS casks required site-
specific approvals because they are constructed in place. Other 
commenters, concerned with seismic events at the Davis-Besse Nuclear 
Power Station and soil stability issues similar to cask use at the 
Palisades Plant, asserted that there was a necessary relationship of 
the Standardized NUHOMS cask or module to the pad at a specific site 
and that evaluation of it could not be based on the reactor site 
seismic analysis. Each site required singular seismic and soil analysis 
for dynamic loads and not just static loads.
    Response. The NUHOMS design referred to in the July 18, 1990, 55 FR 
29181, rulemaking includes the site-specific pad as an integral part of 
the concrete HSM and therefore it is important to safety. The 
Standardized NUHOMS considered in this rulemaking have the HSMs as 
free-standing units; that is, they have no structural connections to 
the pad. The Standardized NUHOMS does not rely on the pad to perform a 
safety function to protect public health and safety. The vendor 
analyzed the HSM containing the DSC for peak ground accelerations of 
0.25g caused by earthquakes and found that it would neither slide nor 
overturn. NRC evaluated the Standardized NUHOMS under a wide range of 
site conditions that could diminish cask safety. Further, under the NRC 
general license, before using the Standardized NUHOMS a licensee must 
verify that reactor site parameters are within the envelope of 
conditions reviewed by NRC for the cask approval. If potential 
conditions exist at the reactor site (including potential erosion, soil 
instability, or earthquakes) that could unacceptably diminish cask 
safety by any credible means, the licensee's analysis must include an 
evaluation of the potential conditions to verify that impairment of 
cask safety is highly unlikely.
    The NRC's regulations do not explicitly require a licensee using a 
cask under a general license to evaluate the cask storage pad and 
foundation under such site conditions for erosion or earthquakes. If 
conditions at the reactor site could unacceptably diminish cask safety 
by affecting the stability of the supporting foundation so as to put 
the cask in an unsafe condition, the cask may not be used unless the 
foundation is appropriately modified or a suitable location at the 
reactor site is found. Implicitly, therefore, the pad and the 
underlying foundation materials must be analyzed under site conditions 
that include erosion, soil instability, and earthquakes, even though 
the pad has no direct safety function and the cask is designed to 
retain its integrity even assuming the occurrence of a range of site 
conditions.
    The licensee has the responsibility under the general license to 
evaluate the match between reactor site parameters and the range of 
site conditions (i.e., the envelope) reviewed by NRC for an approved 
cask. Typically, the licensee will have a substantial amount of 
information already assembled in the Final Safety Analysis Report 
(FSAR) for the nuclear reactor. In addition, the envelope for the 
approved cask is identified in the NRC SER and Certificate of 
Compliance and in the cask vendor's SAR for the cask. Of course, the 
licensee should consider whether the envelope evaluated by NRC 
adequately encompasses the actual location of the cask at the reactor 
site. The licensee should also consider whether there are any site 
conditions associated with the actual cask location that could affect 
cask design and that were not evaluated in the NRC safety evaluation 
for the cask.
    The vendor analyzed the DSC and the HSM for rigid body response 
(i.e., sliding and overturning) to seismic accelerations. The resultant 
peak horizontal ground acceleration is 0.37g and the peak vertical 
acceleration is 0.17g. The margin of safety against sliding is 1.35. 
Similarly, the design seismic force will not cause the HSM to tip over 
because the stabilizing moment of the HSM is greater than the seismic 
overturning moment. The margin of safety against overturning is 1.26. 
Thus, no sliding or overturning of the HSM or DSC will occur from the 
design earthquake.
    Because the pad is not considered a safety-related item, a specific 
pad design is not being approved in this rulemaking for the 
Standardized NUHOMS.
    F.8. Comment. A few commenters had questions pertaining to the 
operation of and procedures for the Standardized NUHOMS. One commenter 
inquired whether just one module of the Standardized NUHOMS could be 
purchased by a utility, or whatever number of modules desired could be 
procured and easily added like singular casks. One commenter expressed 
concern about snow removal procedures to prevent blockage of the bottom 
vents by drifting snow. Another commenter wanted NRC to establish a 
procedure and criteria for dose rates discussed on pages A-15 and A-16 
in the draft Certificate of Compliance. Several commenters noted that a 
procedure for opening a storage cask and removing the fuel has not been 
tried before nor documented in the rulemaking. They were also concerned 
that unloading of a cask would place workers at higher risk.
    Response. The NRC Certificate of Compliance does not permit or 
limit the number of NUHOMS modules that may be purchased by a general 
license. The NRC does not regulate the commercial arrangements between 
the cask vendor and the users including any provisions on the number of 
casks that can be purchased or added to the Standardized NUHOMS.
    Under the Certificate of Compliance, Section 1.3, the user of the 
Standardized NUHOMS (general licensee) is required to conduct a visual 
surveillance of the exterior of air inlets and outlets. If the 
surveillance shows blockage of air vents, they must be cleaned in 
accordance with proper procedures. These procedures will minimize the 
potential impact to the health and safety of workers. The daily 
temperature measurements indicate proper thermal performance.
    The Certificate of Compliance requires each licensee to develop 
procedures to implement the dose criteria prescribed on pages A-15 and 
A-16. On page A-15 of the Certificate of Compliance, Section 1.26, the 
dose rate criteria to be met is equal to or less than: (a) 200 mrem/hr. 
at the top shield plug surface at centerline with water in the cavity; 
and (b) 400 mrem/hr. at the top cover plate surface at centerline 
without water in the cavity. On page A-16 of the Certificate of 
Compliance the dose rate criteria is less than or equal to: (a) 400 
mrem/hr. at 3 feet from the HSM surface; (b) 100 mrem/hr. outside of 
the HSM door on center line of the DSC; and (c) 20 mrem/hr. at the end 
shield wall exterior. Each licensee is required to develop its own 
procedures to implement these criteria. In addition, each licensee must 
develop operational procedures for the ISFSI for workers' radiation 
exposure to be ALARA.
    For the Standardized NUHOMS, removal of spent fuel from the DSC is 
addressed in Chapter 5 of the SAR and in Chapter 11 of the SER. The 
process is essentially the reverse of loading operations and would be 
performed under the reactor license radiation protection program. The 
Certificate of Compliance requires each user to develop written 
procedures for these operations and includes precautions to be 
considered for unloading. ALARA is required to be addressed by 10 CFR 
Part 20. Specification 1.1.6 of the Certificate of Compliance requires 
that pre-operational testing and training exercises include the opening 
of a DSC and returning the DSC and transfer cask to the spent fuel 
pool. The Certificate of Compliance also requires the training program 
to include off-normal events.
    F.9. Comment. One commenter, citing the May 1993 study prepared for 
the NRC by the Center for Nuclear Waste Regulatory Analyses of San 
Antonio, Texas, questioned the relatively higher temperature 
consequences of dry storage on fuel cladding. The report states that, 
``the dry environment has the potential of producing such problems as 
further fuel cladding oxidation, increased cladding stresses, and creep 
deformation as a result of rod internal pressure * * *. These possible 
spent fuel and cladding alteration modes could be quite accelerated 
under dry storage conditions, since temperatures are much higher than 
in wet storage.'' The commenter does not believe that NRC is fulfilling 
its obligation in 10 CFR 72.122(h) to see that ``spent fuel cladding 
must be protected during storage against degradation that leads to 
gross rupture.''
    Response. The May 1993 study addresses the long-term geological 
disposal of high-level waste (spent fuel) and is not directed to the 
short-term interim storage of spent fuel at nuclear power plants. The 
report evaluates processes over 10,000 years of repository performance 
for geological disposal. The conclusions of the report are not 
applicable for the interim storage period of a 20-year cask certificate 
during which spent fuels stored in the DSC have to meet the NRC's 
criteria to ensure that cladding is protected. Under normal operation 
of the ISFSI, leakage of radionuclides is not expected to occur. The 
design and the double-seal welding of the DSC covers are checked and 
tested to provide structural integrity throughout the approved storage 
period. During normal storage conditions, the licensee is required to 
conduct a radiation monitoring program to ensure protection of workers 
and the safety of the general public.
    G. A number of comments were related to broad policy and program 
issues in connection with the storage and disposal of high-level 
radioactive waste, including the DOE repository program. Some 
commenters questioned the use of dry cask storage and technology in 
general. Some commenters stated that only dry storage casks that would 
be compatible with DOE interim or final repository operations, 
including transportation, should be approved for use under a general 
license.
    G.1. Comment. One commenter does not want any more casks approved 
until a permanent Federal repository is opened. The wet fuel pool is a 
proven technology that has been successful in containing radioactivity. 
Another commenter stated that dry storage is dangerous.
    Response. The NRC, in implementing the Nuclear Waste Policy Act of 
1982, has an obligation to review dry storage technologies and to 
determine whether to approve the use of these technologies for the 
storage of spent fuel if they meet applicable safety requirements. The 
July 18, 1990, 55 FR 29181, rulemaking found that spent fuel stored in 
dry storage casks designed to meet the NRC regulatory requirements can 
safely contain radioactivity. This rulemaking adds one cask design that 
meets the safety requirements previously developed to the list of 
approved casks. The previous responses to comments, as well as the 
detailed safety and environmental analyses underlying this rulemaking 
(and described elsewhere in this notice), all reveal that the 
Standardized NUHOMS will conform to the NRC requirements and that its 
use should not pose the potential for significant environmental 
impacts.
    DOE is required by the Nuclear Waste Policy Act of 1982 to accept 
spent fuel for ultimate disposal. Moreover, the Commission made a 
generic determination in its waste Compliance Decisions (September 18, 
1990, 55 FR 38474 and August 31, 1994, 49 FR 34658) that safe disposal 
is technically feasible and will be available within the first quarter 
of the 21st century.
    Dry cask storage has significant advantages over wet storage in 
that the system is passive and requires minimal human intervention. No 
pumps, filters, or water quality monitoring are needed to maintain the 
conditions necessary for wet storage. The only monitoring required for 
the Standardized NUHOMS is daily temperature monitoring and visually 
checking inlet and outlet vents.
    G.2. Comment. A number of commenters wanted a full formal trial-
type public hearing on the use of the NUHOMS cask.
    Response. Consistent with the applicable procedure, the NRC does 
not intend to hold formal trial-type public hearings on the 
Standardized NUHOMS rule or separate hearings at each reactor site 
before the use of the dry cask technology approved by the Commission in 
this rulemaking. Rulemaking procedures, used by the NRC for generic 
approval of the Standardized NUHOMS, including the underlying NRC staff 
technical reviews and the opportunity for public input, are more than 
adequate to obtain public input and assure protection of the public 
health and safety and the environment. In this rulemaking, the NRC has 
taken additional steps to elicit and fully consider public comments on 
the Standardized NUHOMS technology. These steps included NRC 
participation in public meetings near Davis-Besse and extension of the 
public comment period by 45 days in response to public requests. This 
extension provided a total public comment period of almost 4 months.
    Section 133 of the Nuclear Waste Policy Act of 1982 authorizes the 
NRC to approve spent fuel storage technologies by rulemaking. When it 
adopted the generic process in 1990 for the review and approval of dry 
cask storage technologies, the Commission stated that ``casks * * * 
[are to] be approved by rulemaking and any safety issues that are 
connected with the casks are properly addressed in that rulemaking 
rather than in a hearing procedure'' (July 18, 1990; 55 FR 29181). 
Rulemaking under NRC rules of practice, described in 10 CFR 2.804 and 
2.805, provides full opportunity for expression of public views but 
does not use formal trial-type hearings of the kind requested by 
commenters.
    In this proceeding, rulemaking clearly provided adequate avenues 
for members of the public to provide their views regarding NRC's 
proposed approval of the Standardized NUHOMS, including the opportunity 
to participate through the submission of statements, information, data, 
opinions and arguments. In this connection, technical evaluations for 
Standardized NUHOMS and detailed documented findings of compliance with 
NRC safety, security, and environmental requirements were prepared by 
the NRC staff for public examination. In November 1993, the NRC staff 
reviewed the Standardized NUHOMS and approved the design for the 
purpose of initiating this rulemaking to grant a generic approval of 
the design. In addition, the NRC staff conducted a second review in 
response to the public comments on the Standardized NUHOMS in this 
rulemaking, again finding compliance with NRC requirements as discussed 
in this document.
    In addition to reviewing systematically and in depth the technical 
issues important to protecting public health and safety, and the 
environment, the NRC has taken extra steps to obtain and fully consider 
public views on the Standardized NUHOMS technology and has made every 
effort to respond to public concerns and questions about the 
Standardized NUHOMS compliance with NRC safety, security, and 
environmental requirements. The initial public comment period opened on 
June 2, 1994, and was scheduled to close on August 16, 1994. On August 
29, 1994, the public comment period was extended to September 30, 1994. 
The NRC also participated in an earlier meeting near the Davis-Besse 
site.
    Under these circumstances, formal hearings would not appreciably 
add to NRC's efforts to ensure adequate protection of public health, 
safety, and the environment and they are unnecessary to NRC's full 
understanding and consideration of public views on the Standardized 
NUHOMS.
    G.3. Comment. One commenter stated that because there is not now 
and there may not be a permanent high-level radioactive waste (HLWR) 
repository for commercial reactor fuel, and since the NUHOMS 24P and 
52B casks are non- transportable, any distinction between so called 
``temporary storage'' and ``permanent disposal'' of this waste is moot. 
Because of the lack of a permanent repository or Monitored Retrievable 
Storage (MRS) in the foreseeable future, a case of a serious spill and 
the resultant contamination at an environmentally unsuitable site like 
Davis-Besse where ``short and long-term adverse impacts associated with 
the occupancy and modification of (a) floodplain * * * potential 
release of radioactive material during the lifetime of the ISFSI * * * 
(and location) over an aquifer which is a major water resource'' have 
been inadequately dealt with.
    Response. This rulemaking to certify the Standardized NUHOMS is for 
interim storage of spent fuel in an approved cask for 20 years. It does 
not authorize or approve the ultimate disposal in a permanent HLRW 
repository, which is under the responsibility of the DOE. During 
interim storage, the user (holder of a Part 50 license) must protect 
the spent fuel against design basis threats, and against environmental 
conditions and natural phenomena such as tornadoes, tornado missiles, 
earthquakes, and floods. In regard to flooding, the Certificate of 
Compliance has a provision (see A-2 of Certificate of Compliance) for 
flood condition analysis to ensure that there is no release of 
radioactive material from flooding.
    G.4. Comment. One commenter stated that projected future uses of 
land and water within the region are impossible to make given the 
unknown length of time this waste may remain on site and the options 
for both cask and reactor license renewal beyond 20 and 40 years, 
respectively, and the fact that no known man-made structure can last 
for the length of time that this waste must be isolated from humans and 
the environment. If an MRS or repository ever become available, this 
waste may have to be repacked. Each handling of this waste increases 
the likelihood of an accident, spill, contamination, and worker and 
public exposures.
    Response. Projected future land and water use can be made based on 
the continued safe operation of a reactor and its associated dry cask 
storage facility. The continued operation of these facilities should 
have no greater impact on land and water use in the future than they do 
today. As previously noted, the NRC Waste Confidence decisions 
concluded there is reasonable assurance that safe disposal of spent 
fuel by the Federal Government will be available by the year 2025. 
Therefore, the spent fuel will not remain at a reactor site for the 
length of time it must be isolated from humans and the environment.
    It should be noted that the absence of significant environmental 
impacts from dry cask storage at a reactor site is the conclusion of 
NRC's environmental assessment for the Standardized NUHOMS and for 
previously approved dry casks analyzed in earlier rulemakings 
addressing 10 CFR Part 72, as well as in the Commission's Waste 
Confidence decisions in 1984 (August 31, 1984; 49 FR 34658) and 1989 
(September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence 
decision, the Commission concluded there was reasonable assurance that 
spent fuel can be safely stored at reactor sites, without significant 
environmental impacts, for at least 30 years beyond expiration of NRC 
reactor operating licenses. The 1989 Waste Confidence decision review 
reaffirmed earlier Commission conclusions on the absence of significant 
environmental impacts.
    G.5. Comment. One commenter questioned whether the NUHOMS canister 
will fit the conceptual design for the DOE multi-purpose canister 
(MPC). If DOE chooses to use vertical casks (like the VSC) at the MRS, 
will the NUHOMS inner canister fit into the vertical outer concrete 
shell in the MPC design? If local reactors choose the VSC-24 or the 
NUHOMS, will either inner metal canister fit into the overpacks for 
DOE, or will they have to be opened after storage, returned to the 
pool, the fuel put in a new canister, and the old one discarded as 
radioactive waste?
    Response. The Certificate of Compliance for the Standardized NUHOMS 
is intended for the interim storage of spent fuels and is not required 
to conform to, and has not been evaluated by NRC for conformance with, 
the conceptual design for the DOE MPC. DOE has not yet made final 
decisions regarding design or deployment of the MPC. Therefore, it is 
not possible to speculate on conformance of the Standardized NUHOMS to 
the MPC.
    G.6. Comment. One commenter asked what are the criteria for 20-year 
renewal of this cask design? How will this be checked? If the design is 
not renewed, what is the plan?
    Response. The 1989 proposed rule (May 5, 1989; 54 FR 19379) to add 
Subparts K and L to Part 72 indicated that the 20-year period 
represents what the Commission believes to be an appropriate increment 
for cask design approvals. The application for design reapproval would 
have to demonstrate the cask's ability to perform the necessary safety 
functions for the reapproval period. The application would be evaluated 
by NRC against the Commission's regulatory requirements. If a cask 
design is not reapproved, the licensee would have to remove casks from 
service as the 20-year approved storage life expired. This could mean 
removal of the spent fuel and storing it elsewhere.
    G.7. Comment. One commenter wanted to discuss the need for an 
additional cask design, including how it would better meet the need of 
the interim dry cask storage of high-level waste.
    Response. Section 218(a) of the Nuclear Waste Policy Act of 1982 
(NWPA) provides the following directive: ``The Secretary [of DOE] shall 
establish a demonstration program in cooperation with the private 
sector, for the dry storage of spent nuclear fuel at civilian nuclear 
reactor power 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.'' After subsequent DOE 
technical evaluations and based on a full review of all available data, 
the Commission approved dry storage of spent nuclear fuel in a final 
rule published in the Federal Register on July 18, 1990 (55 FR 29181). 
The final rule established a new Subpart K within 10 CFR Part 72, 
entitled ``General License for Storage of Spent Fuel at Power Reactor 
Sites.'' Therefore, there is a need for casks to be approved by NRC to 
implement the NWPA to meet the demand of the interim dry cask storage 
of spent fuels in the nuclear power plants. However, the variety of 
cask designs submitted by vendors for NRC review and approval is mostly 
dictated by economic reasons that do not involve NRC.
    H. A number of commenters wanted site-specific analyses done for 
each use of the Standardized NUHOMS despite the fact that each licensee 
must determine that the site parameters are enveloped by the cask 
design specified in the SAR, SER, and Certificate of Compliance. The 
intent of Subpart K of 10 CFR Part 72 was to grant a general license to 
licensees of power reactors to use NRC-approved dry storage casks 
listed in 10 CFR 72.214 without additional licensing review by NRC.
    H.1. Comment. A number of commenters wanted site-specific 
Environmental Impact Statements (EIS). Several commenters stated that 
an EIS should be required on any waste facility that may be permanent 
along the Great Lakes fresh water system. To say that this will have no 
adverse effect on public health and safety is a prediction most of the 
public does not accept. The commenter believes that the generic ruling 
to use a dry cask storage design at any reactor site is impossible and 
should be discarded. By relying on environmental evaluations done in 
the 1970s before Davis-Besse construction, the NRC was remiss in its 
responsibility to protect the people of Ohio from harm by its licensee. 
Another commenter wants the NRC to prepare, at a minimum, an 
Environmental Assessment (EA) for each site, including information on 
sensitive ecosystems, wildlife, demography, meteorology, and geology. 
The EA should discuss the cask's capability to withstand weather 
conditions and potential catastrophic events.
    Response. The potential environmental impacts of utilities using 
the Standardized NUHOMS (or any of the other spent fuel casks approved 
by NRC (10 CFR 72.214)) have been fully considered and are documented 
in a published Environmental Assessment (EA) covering this rulemaking. 
Further, as described below, the EA indicates that use of the casks 
would not have significant environmental impacts. Specifically, the EA 
notes the 30-plus years of experience with dry storage of spent fuel 
have shown that the previous extensive NRC analyses and findings that 
the environmental impacts of dry storage are small and succinctly 
describes the impacts, including the non-radiological impacts of cask 
fabrication (the impacts associated with the relatively small amounts 
of steel, concrete, and plastic used in the casks are expected to be 
insignificant), the radiological impacts of cask operations (the 
incremental offsite doses are expected to be a small fraction of and 
well within the 25 mrem/yr limits in NRC regulations), the potential 
impacts of a possible dry cask accident (the impacts are expected to be 
no greater than the impacts of an accident involving the spent fuel 
storage basin), and the potential impacts from possible sabotage (the 
offsite dose is calculated to be about one rem). All of the NRC 
analyses collectively yield the singular conclusion that the 
environmental impacts and risks are expected to be extremely small.
    NRC EA's for previously approved dry casks also concluded there was 
an absence of significant environmental impacts from dry cask storage 
at a reactor site when they were analyzed in earlier rulemakings 
addressing 10 CFR Part 72 as well as in the Commission's Waste 
Confidence decisions in 1984 (August 31, 1984; 49 FR 34658) and 1989 
(September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence 
decision, the Commission concluded there was reasonable assurance spent 
fuel can be safely stored at reactor sites, without significant 
environmental impacts for at least 30 years beyond expiration of NRC 
reactor operating licenses. The 1989 Waste Confidence decision review 
reaffirmed earlier Commission conclusions on the absence of significant 
environmental impacts.
    Given the Commission's specific consideration of environmental 
impacts of dry storage and the absence of any new information casting 
doubt on the conclusion that these impacts are expected to be extremely 
small and not environmentally significant, the NRC is not convinced 
that meaningful new environmental insights would be gained from either 
a new site-specific EIS or EA for each site using dry storage methods.
    The EA covering the proposed rule, as well as the finding of no 
significant impact (FONSI) prepared and published for this rulemaking, 
fully comply with the NRC environmental regulations in 10 CFR Part 51. 
The Commission's environmental regulations in Part 51 implement the 
National Environmental Policy Act (NEPA) and give proper consideration 
to the guidelines of the Council of Environmental Quality (CEQ). The EA 
and FONSI prepared as required by 10 CFR Part 51 conform to NEPA 
procedural requirements. Further analyses are not legally required.
    The regulation 10 CFR Part 72, Subpart K, already authorizes dry 
cask storage and approves dry casks for use by utilities to store spent 
fuel at reactor sites. See 10 CFR 72.214 for a listing of information 
on Cask Certificate Nos. 1000 through 1003, 1005, and 1007. The purpose 
of this final rule is to add one more cask to the list of casks already 
approved by NRC. The cask added to the list in Sec. 72.214 by this 
final rule complies with all applicable NRC safety requirements.
    Finally, this final rulemaking applies to the use of this cask by 
any power reactor within the United States.
    H.2. Comment. One commenter stated that the January 30, 1994, reply 
from NRC's Robert Bernero to Mr. Adamkus, EPA, is completely 
inadequate, as is the March 1994 ``Draft Environment Assessment and 
Finding of No Significant Impact'' because no consideration is given to 
the site's unsuitability even for LLRW per NRC's own admission, and 
``new information which could alter the original site evaluation 
findings'' is ignored.
    Response. This final rule does not provide any site-specific NRC 
approval or address site-specific parameters that are peculiar to a 
particular reactor site. The rule only adds one cask design, the 
Standardized NUHOMS, to the list of approved casks available for use by 
a power plant licensee in accordance with the conditions of the general 
license in Part 72. Pursuant to those conditions, each licensee must 
determine whether or not the reactor site parameters (including 
earthquake intensity and tornado missiles) are encompassed by the cask 
design bases considered in the cask SAR and SER. The EA and FONSI for 
this rule are limited in scope to the Standardized NUHOMS in a generic 
setting.
    Unlike interim storage prescribed in 10 CFR Part 72, the in-ground 
disposal of radioactive material, whether high-level or low-level waste 
(HLW or LLW), must take into account the geologic, hydrologic, and 
geochemical characteristics of the site or region to isolate the 
radioactive waste from the accessible environment. Site criteria for 
in-ground disposal of radioactive wastes enable an applicant to choose 
an appropriate site, one with a combination of favorable conditions 
that will be a natural barrier to retard or attenuate the migration of 
any leaked radioactive material over a long period to control releases 
within acceptable limits. The disposal period for LLW is on the order 
of 500 years, and for HLW it is greater than 10,000 years. Therefore, 
site characteristics are investigated and assessed for interim spent 
fuel storage under Part 72, not to determine their suitability as a 
barrier to release of radioactive material, but rather to determine the 
frequency and the severity of external natural and artificial events 
that could affect the safety of an ISFSI. Unlikely, but credible, 
severe events are considered to determine the safety of the storage 
cask design.
    H.3. Comment. One commenter stated that the NRC has not approved 
technologies for the use of spent fuel at the sites of * * * without 
the need for additional site reviews. If that were so, no additional 
site review would have been necessary at Palisades, nor would an SAR 
revision or a Certificate of Compliance amendment be called for right 
after the VSC-24 was certified.
    Response. The approval and use of dry storage technologies under 
the provisions of the general license are relatively new. Questions 
were raised by members of the public about the possible effects of 
earthquakes and erosion at the Palisades site on the safe storage of 
spent fuel in the VSC-24 dry casks. As the agency which is responsible 
for questions about compliance with regulatory requirements, which 
oversees such matters as the ``cop on the beat,'' the NRC began an 
independent assessment to more closely examine the behavior of the pad 
at Palisades under normal conditions, under the long-term effects of 
erosion, and under conditions of a postulated earthquake that might 
cause the sand below or around the pad to move. The results of NRC's 
assessment were documented in the NRC Final Safety Assessment of 
Independent Spent Fuel Storage Installation (ISFSI) Support Pad (TAC 
No. M88875). As is the case at all sites, NRC requires the cask user to 
determine if the design basis for the storage technology being 
considered encompasses the site parameters at the location where the 
fuel is to be stored. The review at Palisades confirmed this to be the 
case. As experience with use of this new design is gained, 
modifications to the design described in the SAR are expected and 
allowed under the provisions of 10 CFR 72.48.
    H.4. Comment. One commenter wanted the environmental impacts of 
alternatives, such as: renewable energy sources, conservation of 
energy, shutting down the nuclear power plants, and wind and solar 
power evaluated.
    Response. Energy production is not the subject of this rulemaking 
and alternative sources of energy are, therefore, not reasonable 
alternatives requiring evaluation. This rulemaking is limited to the 
addition of the Standardized NUHOMS to the list of approved casks in 10 
CFR 72.214.
    H.5. Comment. One commenter stated that the NRC is ignoring the 
regulatory requirements of a site-specific license as to the 
feasibility of using the cask or of modifying its design.
    Response. This rulemaking does not cover site-specific NRC 
licensees; however, the NRC is not ignoring them. Under NRC 
regulations, the utility has two options in using dry cask storage of 
spent fuel: (1) The licensee may apply for a site-specific license from 
NRC; or (2) the licensee may use an NRC-approved cask under the general 
license provisions of Subpart K of 10 CFR Part 72. However, not all 
licensees may be able to use the general license provisions, either 
because the fuel type they possess is not storable in any cask listed 
in 10 CFR 72.214 or because none of the cask designs envelope the 
reactor site parameters. The NRC is also not ignoring site-specific 
license considerations relating to modifying cask designs. Quite the 
contrary, the criteria that apply to modifications of an NRC-approved 
cask such as the Standardized NUHOMS are the same as the criteria that 
apply to modifications of site-specific ISFSIs.
    H.6. Comment. Because the populations of several states and 
provinces, including two-thirds of the population of Quebec, are based 
along the St. Lawrence Seaway, one commenter wanted an Economic Impact 
Statement conducted with a cost/benefit analysis citing possible 
adverse impact on tourism and sport fishing.
    Response. A regulatory analysis, which considers both benefits and 
impacts of adding the Standardized NUHOMS to the list of NRC-approved 
casks under Subpart K of 10 CFR Part 72, was prepared in support of 
this rulemaking action. It was included as a part of the notice of 
proposed rulemaking and is also included in this final rulemaking 
notice. However, this regulatory analysis reflects the limited scope of 
this rulemaking. Because the rulemaking does not provide any site-
specific NRC approvals, NRC did not evaluate site-specific economic 
impacts.
    H.7. Comment. One commenter wanted to restrict the use of the cask 
to reactor sites that have responded on schedule to NRC Generic Letter 
88-20, Supplement 4, ``Individual Plant Examination of External Events 
(IPEEE).''
    Response. IPEEE response submittals will not address dry cask 
storage and are not necessary for Standardized NUHOMS use.
    H.8 Comment. One commenter stated that NUHOMS must not receive 
generic approval because site-specific characteristics must be 
considered. The commenter stated that placing this cask on the shores 
of Lake Erie is potential ecocide and the cask is not terrorist-proof. 
Another commenter stated that the potential engineering problems of 
storing high-level nuclear waste in a variety of climatic and geologic 
regions of the United States are not considered.
    Response. A utility's use of the Standardized NUHOMS, for the 
storage of spent fuel in casks at a reactor site, would not have a 
significant impact on the environment. This finding is supported by the 
NRC safety and environmental evaluations for the Standardized NUHOMS, 
including the applicant's demonstration of compliance of the cask with 
NRC requirements, as well as by the 1990 rulemaking on dry cask storage 
and the 1984 and 1989 waste confidence proceedings. Because the 
Standardized NUHOMS can only be used by a licensee if the site 
parameters are enveloped by the cask design basis, as specified in the 
SAR and SER, cask storage of spent fuel near the shore of Lake Erie 
within the specified parameters would not have a significant impact on 
the environment.
    I. The following comments relate to the transportability of dry 
storage casks to an off-site location.
    I.1. Comment. One commenter questioned how the cask transport 
methods used at both on-site and off-site locations are related.
    Response. In this rulemaking, the NRC reviewed the cask vendor's 
proposed means for transporting the Standardized NUHOMS canister and 
transfer cask outside the reactor buildings to the on-site storage pad 
under the storage requirements of 10 CFR Part 72. This on-site movement 
occurs within an owner-controlled area where access can be limited and 
where operations would be safely managed by the general licensee. The 
NRC did not review the Standardized NUHOMS for transport off-site, for 
example to a DOE MRS or repository. Generally, off-site transport of 
spent fuel occurs in public places where the shipper has fewer access 
restrictions and limited control of the surroundings. Off-site spent 
nuclear fuel shipments must be made in a transportation cask approved 
by the NRC pursuant to NRC's regulations found in 10 CFR Part 71, 
``Packaging and Transportation of Radioactive Material,'' and must also 
comply with pertinent Department of Transportation (DOT) regulations. 
At this time, the NRC is approving the Standardized NUHOMS for storage 
only.
    I.2. Comment. One commenter, citing a Wisconsin Public Service 
Commission EIS for Point Beach, questioned the statement, ``The 
baskets' heavier weight and larger diameter make the transportability 
of an intact NUHOMS canister to an MRS site or repository 
questionable.''
    Response. The NRC has not reviewed the Standardized NUHOMS in this 
rulemaking for off-site transportation.
    I.3. Comment. One commenter wanted to know the relationship between 
the Standardized NUHOMS and the NUHOMS MP187 now applying for a 
Certificate of Compliance. Is the MP187 transportable? Will the 
canister of all models fit into the transport overpack? Wouldn't a 
utility be better off waiting for the transportable cask rather than 
choosing a storage only cask that may have compatibility problems with 
an MPC system?
    Response. The MP-187 transportation overpack uses a canister 
similar to the Standardized NUHOMS. However, it is the subject of a 
separate NRC review as part of a site-specific licensing application. 
Both the Standardized NUHOMS and the MP-187 share many common design 
features. However, they are separate applications, and the NRC has not 
been asked by the cask vendor to review whether the Standardized NUHOMS 
can be transported in the NUHOMS MP187 transportation overpack.
    The issue of whether a utility should consider the transportability 
of dry storage casks is beyond the scope of this rulemaking.
    I.4. Comment. One commenter cited a report given at the HLW 
Conference at Las Vegas, in 1990, ``Integrated Spent Fuel Storage and 
Transportation Systems using NUHOMS,'' by PNFSI (page 671): ``While 
subsequent transfer of an intact DSC from a NUHOMS on-site transfer 
cask directly to an OCRWM rail/barge is feasible, this method of 
transfer is not preferred since the assemblies would be oriented top 
down and the DSC bottom shield plug and grapple ring assembly would be 
orientated top up, thus complicating the canister opening and fuel 
handling process at the MRS or geologic repository following 
shipment.'' Has NRC evaluated this situation? Has it been rectified?
    Response. Because the cask vendor applied for certification of the 
Standardized NUHOMS only as a storage cask under 10 CFR Part 72, 
transportation of this cask is not a subject of this rulemaking. 
Therefore, the NRC review of the standardized NUHOMS did not consider 
the particular transportation problem described in the comment.
    J. Several commenters supported the rule stating that it is 
beneficial to the NRC and licensees, and it is consistent with NRC's 
direction to avoid unnecessary site-specific licensing reviews. Others 
disagreed and asked specific questions about NRC's approval and 
oversight process.
    J.1. Comment. One commenter stated that the NRC statement, ``The 
proposed rule will not have adverse effect on public health and 
safety,'' cannot be guaranteed and, therefore, even though it may be 
convenient for the nuclear industry and the NRC to avoid site-specific 
approvals, in this case these are essential for maintaining public 
safety. Another commenter following the same theme questioned how the 
following determination was made: ``this cask, when used in accordance 
with the conditions specified in the Certificate of Compliance and NRC 
regulations, will meet the requirements of 10 CFR Part 72; thus, 
adequate protection of the public health and safety would be ensured.''
    Response. Dry storage casks approved by the NRC for use under the 
general license are of a robust design that relies on generic cask 
features to ensure protection of the public health and safety. 
Additional NRC site-specific approvals are unnecessary. NRC oversight 
and inspections are sufficient to ensure that general licensees 
implement NRC conditions on cask use. If specific concerns are raised, 
the NRC also has the authority to look into them and respond as 
necessary to protect public health and safety. The NRC has established 
specific requirements in 10 CFR Part 72 that must be met in order to 
obtain a Certificate of Compliance for a cask. The details of the 
review and the bases for the NRC concluding that the cask meets the 
requirements of 10 CFR Part 72 are provided in the SER. The goal of dry 
cask storage technology is to store spent fuel safely. That goal, and 
the effectiveness of the technology, have been demonstrated empirically 
and experimentally. Different cask designs may require different types 
of analysis to demonstrate their safety. Therefore, different review 
methods may be appropriate to reach that conclusion. In each case, the 
level of review performed is that needed to provide assurance of 
adequate protection of the public health and safety.
    J.2. Comment. Several commenters expressed concern over the 
exemption to 10 CFR 72.234(c) granted to VECTRA to begin transfer cask 
fabrication (but not use) ``to have the necessary equipment available 
for use by Davis-Besse Nuclear Power Station (DBNPS) in mid-1995, and 
thus enable DBNPS to maintain complete full-core off-load capability in 
its spent fuel pool following the refueling outage scheduled for early 
1996.'' One commenter said that seeking public comment and providing 
comments is an exercise in futility because cask approval seems to be a 
fait accompli. Another commenter wants no exemptions for fabrication 
before certification to be allowed, stating that problems have 
developed when all these exemptions are allowed.
    Response. The NRC granted VECTRA's request for an exemption to 
fabricate the transfer cask before issuance of the Certificate of 
Compliance under its NRC-approved quality assurance program. NRC's 
exemption decision made a special effort to clarify that fabrication 
was entirely at VECTRA's financial risk and did not ensure favorable 
consideration of VECTRA's application. The NRC's finding, based on the 
SAR for the Standardized NUHOMS and the NRC's SER, concluded that 
beginning fabrication before the issuance of the Certificate of 
Compliance would pose no undue risk to public health and safety. Use of 
the transfer cask is dependent on satisfactory completion of NRC's 
certification process.
    The NRC staff carefully considers the public comments received in 
rulemakings to determine whether changes are needed to the proposed 
rule. As noted elsewhere in this notice, several public comments 
received in this and other cask-approval rulemakings have resulted in 
changes to the SER and the Certificate of Compliance. For this reason, 
the public comments provide useful inputs to the NRC's safety approval 
process.
    J.3. Comment. One commenter wanted a Regulatory Guide outlining the 
requirements of an SAR for cask certification (CSAR). Requirements for 
a CSAR have not been clarified. Specific criteria for a TR (TSAR) by a 
vendor for a generic Certificate of Compliance need to be set.
    Response. Regulatory Guide 3.61, ``Standard Format and Content for 
a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask,'' 
dated February 1989, provides guidance for the preparation of a TSAR. 
Regulatory Guide 3.62, ``Standard Format and Content for the Safety 
Analysis Report for Onsite Storage of Spent Fuel Storage Casks,'' dated 
February 1989, provides guidance in preparing an SAR locating an ISFSI 
at a reactor site. Both Regulatory Guides identify similar information 
that can be potentially useful to prospective applicants for cask 
certification.
    J.4. Comment. One commenter wanted to know why Pacific Nuclear 
divested itself of any ownership or relationship to the VSC design in 
January 1992. How does this affect proprietary material shared in these 
two closely related designs? How does it affect their relationship to 
the DOE MPC system?
    Response. The key individual involved in the design and development 
of the VSC-24, who was also involved in the design and development of 
the NUHOMS design, left Pacific Nuclear and formed a new company, 
Pacific Sierra Nuclear, for the commercial manufacture and marketing of 
the VSC-24 storage system. The NRC has experienced no difficulty 
obtaining the required safety information, including proprietary 
information or answers to its questions from either firm, either before 
or after divestiture. The NRC is not aware of any relationship between 
the vendors. In addition, the NRC fully reviewed the health and safety 
aspects of each vendor's cask design independently. The NRC did not 
rely on any assumed relationship between the two vendors. Concerning 
their relationship to the DOE MPC system, each vendor has to establish 
its own relationship with DOE.
    J.5. Comment. One commenter wanted to know how long any model of 
NUHOMS has been used and if fuel has been taken out and evaluated. Has 
the 24P or 52B ever been used anywhere and for how long? If not, this 
is a test of a new cask at a reactor site.
    Response. The NUHOMS-24P is being used at Duke Power Company, 
Oconee Nuclear Station, under a site-specific license issued January 
29, 1990, and at Baltimore Gas and Electric Company, Calvert Cliffs 
Nuclear Station, under a site-specific license issued November 25, 
1992. Monitoring and surveillance of the system is being performed 
under the conditions of the site-specific license. However, there has 
been no need for fuel to be removed for evaluation.
    The NUHOMS-52B has not been used yet. Pre-operational testing of 
the first cask system put in place under the general license is to be 
performed in accordance with Certificate of Compliance, Attachment A, 
``Conditions for Systems Use.'' Monitoring and surveillance of the 
system will be performed under the conditions of the Certificate of 
Compliance.
    The first use of the Standardized NUHOMS-52B will not place plant 
workers, the public, or the environment at risk. Conditions of use for 
the Standardized NUHOMS-52B ensure adequate safety of the workers, the 
public, and the environment. The Standardized NUHOMS-52B has been 
designed and will be fabricated to well established criteria of the 
ASME B&PV and ACI codes. It uses construction materials that have well 
known and documented properties to provide the necessary structural 
strength and radiation shielding to meet regulatory requirements. While 
the Standardized NUHOMS-52B is not identical to the NUHOMS-24P, many 
parallels in design and function can be drawn to demonstrate that the 
Standardized NUHOMS-52B will perform as intended.
    J.6. Comment. One commenter stated that even though dry cask 
storage passes all NRC rules and is one of the least expensive methods, 
it would seem that a different location or more expensive storage 
method is worth lives, resources, and property.
    Response. Based on numerous NRC reviews and growing experience with 
dry cask storage technologies, the NRC has concluded that spent fuel 
can be safely stored in dry casks without significant risk to the 
public health and safety. More expensive storage techniques or 
alternative storage locations would not provide any significant 
additional public protection. Further, the storage location is a matter 
of Congressional policy as reflected in Section 218(a) of the Nuclear 
Waste Policy Act of 1982, which includes the following directive: ``The 
Secretary [of 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 III(a) also finds that the generators of the 
spent fuel have the primary responsibility to provide for the interim 
storage of the spent fuel until it is accepted by the DOE.
    The type of spent fuel stored in the dry cask storage systems is 
one factor that allows the cost of the systems to be lower. Because the 
fuel has cooled a number of years, passive cooling can be used rather 
than active cooling as is required for fuel just removed from the 
reactor. Passive cooling reduces the cost by not having active 
components such as pumps, heat exchanger, water filters, and the 
maintenance required for these components.
    J.7. Comment. One commenter opposed licensing any dry cask storage 
system other than the DOE multi-purpose canister (MPC) because it 
minimizes handling individual fuel assemblies, standardizes 
compatibility between storage sites and DOE, and reduces cost. Multiple 
cask designs lead to less expertise in production, operation, and 
accident management. Federal regulations need to be amended to mandate 
only the use of the MPC.
    Response. The DOE MPC system will not be available for general use 
until well after 1997. In the meantime, additional storage capacity is 
needed now at several reactor sites. Once the MPC is available for 
general use, most utilities might use it. However, given the 
demonstrated and immediate need of some reactors for an additional 
storage capacity, and given NRC's responsibility to implement dry cask 
storage under a general license pursuant to NWPA of 1982, it would not 
be prudent for NRC now to require use of MPC designs that not even DOE 
has yet approved.
    The NRC does not agree that the number of cask designs has a 
significant effect on the level of expertise available because standard 
engineering and scientific skills such as mechanical and civil 
engineers and health safety specialists can be hired as needed.
    K. Several commenters had concerns about decommissioning issues.
    K.1. Comment. One commenter, citing the draft SER, stated that 
decommissioning and decontamination of reactors and reactor sites 
remain uncertain at best. ``At this time, it is not known whether 
demolition and removal of the HSM can be performed by conventional 
methods * * *. The reinforced structure of the HSM, for example, will 
require considerable effort to demolish.'' The commenter continues by 
indicating that in its typical fashion of putting off until tomorrow 
what it cannot deal with today, the NRC considers ``ease of 
decommissioning (a) secondary consideration.''
    Response. The demolition of the HSM will be more difficult than a 
typical building because of the large amount of reinforced steel it 
contains. However, it is technically feasible and represents a likely 
level of effort similar to that required to demolish a bank vault. Bank 
vaults are routinely demolished without extraordinary effort. The HSM 
may become slightly radioactive from being exposed to a neutron 
radiation field during the spent fuel storage period, which would 
require some containment during demolition to prevent the spread of 
contamination. Recognizing this, the NRC considers decommissioning a 
secondary consideration compared to the safety afforded by storage of 
spent fuel in dry casks.
    K.2. Comment. One commenter questioned how, where to, and when the 
spent fuel and casks will go? How does the decommissioning of NUHOMS 
affect the reactor decommissioning plan if no repository is sited and 
the pool must remain open? Another commenter expressed concern that 
after the operating facility has been decommissioned, the spent fuel 
pool may not be available for use in recovery of a breached DSC.
    Response. The Commission determined in the Waste Confidence 
decisions that sufficient repository capacity will be available, in the 
first quarter of the 21st century, to accept spent fuel that is already 
in storage or that will be generated during the lifetime of the reactor 
licensed by NRC. In addition, the Commission determined that spent fuel 
can be safely stored at reactors until it is disposed. The bases for 
these determinations are extensively discussed in the Waste Confidence 
decisions (54 FR 39765; September 28, 1989 and 49 FR 34658; August 31, 
1984) and remain applicable today.
    To operate the dry spent fuel storage area under the provisions of 
the general license, a license to possess or operate a nuclear power 
reactor under 10 CFR Part 50 is required. If the reactors were 
decommissioned and the license terminated, and if the spent fuel were 
to remain on site, a specific license issued under 10 CFR 72.40 would 
be required. At the time of application for a specific license and 
before the Part 50 license was terminated, the licensee would have to 
address the subject of how the fuel will be repackaged for shipment to 
an MRS or repository. (None of the casks now listed in 10 CFR 72.214 
are approved for transportation). Decommissioning and termination of a 
Part 50 license for a given reactor site must take into account the 
proper disposal of any spent fuel.
    L. A number of positive and negative comments were received about 
the application of 10 CFR 72.48 or Item 9 of the Certificate of 
Compliance to general licensees.
    L.1. Comment. Several commenters questioned the application of 10 
CFR 72.48 to Certificate of Compliance holders for use by a general 
licensee. Some commenters believe that this regulation is being 
inappropriately applied to general licensees and cask vendors. These 
commenters believe that the regulation was intended to apply to site-
specific licenses issued under 10 CFR 72.40 only. One commenter cited 
the parallel application of 10 CFR 50.59 to 10 CFR Part 50 licensees. 
Any changes to the Certificate of Compliance and the supporting SAR and 
SER need public input using the rulemaking process. Who would make the 
decisions in using the terms ``unreviewed safety questions,'' 
``significant increase,'' and ``significant environmental impact''? 
Other commenters liked this addition, stating that non-safety-
significant changes can be made in a timely and cost effective manner. 
Several commenters supported the incorporation of item number 9 (in 
72.48 type language) in the draft Certificate of Compliance. One 
commenter wanted similar provisions made for general license holders 
with recordkeeping requirements applicable to the general license 
rather than the certificate holder. Changes requiring an amendment to 
the certificate should be initiated by the certificate holder only.
    Response. The NRC will not allow changes in the Certificate of 
Compliance under 10 CFR 72.48. However, the general licensee may make 
changes in the SAR under 10 CFR 72.48, unless it involves an unreviewed 
safety question, a significant increase in occupational exposure, or a 
significant unreviewed environmental impact. The general licensee must 
make the determinations, in the first instance, that are necessary for 
application of 10 CFR 72.48. The licensee must also retain its 
evaluations on its records (which are subject to NRC review).
    Supporting this application of 10 CFR 72.48 to the general license 
are the words of 10 CFR 72.48(a)(1) which provides as follows: ``The 
holder of a license issued under this part may: (i) Make changes in the 
ISFSI * * * described in the Safety Analysis Report, * * * (iii) * * * 
without prior Commission approval, unless the proposed change, test or 
experiment involves a change in the license conditions incorporated in 
the license, an unreviewed safety question, a significant increase in 
occupational exposure, or a significant unreviewed environmental 
impact.'' Also supporting the interpretation is 10 CFR 72.210 which 
provides as follows: ``A general license is hereby issued for the 
storage of spent fuel in an independent spent fuel storage installation 
at power reactor sites to persons authorized to possess or operate 
nuclear power reactors under Part 50 of this chapter.'' The NRC staff 
is considering a rulemaking to amend NRC regulations to explicitly 
state that 10 CFR 72.48 applies to general licensees.
    L.2. Comment. One commenter stated that the CFR is silent on how a 
vendor can change a cask SAR and certificate after the final rule. It 
should be made clear for the vendor that this cask SAR (CSAR) is 
generic for all United States sites. All seismic, control component, 
distance, changes in length and weight, changes in transfer devices, 
etc., need to be clearly defined in the proposed rulemaking for the 
cask and the CSAR before public comment. Who would be liable if a 
utility requested the vendor to change a certified cask design?
    Response. The cask vendor can apply to the NRC for a change to the 
cask certificate and SAR after the final rule is published in the 
Federal Register. The vendor must propose the generic revisions to the 
certificate and SAR and request NRC review of the proposed revision. 
The NRC will evaluate the proposed revision in an SER, and if 
appropriate, prepare a draft revised Certificate of Compliance. These 
documents would then be placed in the NRC Public Document Room and a 
proposed rule would be published requesting public comments on the 
proposed revised Certificate of Compliance. After consideration of 
public comments (and assuming an appropriate basis exists), a final 
rule would be published incorporating the revision in the revised 
Certificate of Compliance.
    The SAR (CSAR) is not necessarily generic for all United States 
operating reactor sites as the comment appears to suggest. The SAR is 
pertinent for those sites that have parameters that are incorporated by 
the cask design bases analyzed in the SAR. From a practical standpoint, 
it is difficult for a cask vendor to foresee all possible combinations 
of seismic, control component, distance, changes in length and weight, 
changes in transfer devices, etc. Revisions are expected when the 
vendor submits its initial application for approval. The vendor is 
responsible for the certified cask design.
    L.3. Comment. One commenter wanted an explanation for not allowing 
buyer substitution of material for a Certificate of Compliance and that 
these references should be deleted from fabrication specifications and 
drawings. Does this mean that no changes in any materials are allowed 
once the design is certified? If so, explain this in reference to new 
models of the VSC-24 as far as materials, coatings, etc.?
    Response. Under 10 CFR Part 72, the licensee is permitted to make 
changes in the ISFSI as described in the SAR provided the changes do 
not involve an unreviewed safety question. The licensee and cask 
certificate holder must have a quality assurance (QA) program that 
provides control over activities affecting quality of the identified 
structures, systems, and components to an extent commensurate with the 
importance to safety and to ensure conformance with the approved 
design. The NRC does not want buyers (who may not be the licensee or 
certificate holder) of cask materials to automatically be able to 
substitute material without the necessary safety evaluations. Rather, 
the licensee, through the cask certificate holder, has the ultimate 
responsibility for approving any changes to ensure conformance with the 
approved design. For structures, systems, and components identified as 
important to safety, if alternative materials are desired to be used 
and those specific materials form the basis of the safety evaluation, 
it would be appropriate to identify those materials in the cask 
application. Alternatively, the certificate holder may seek an 
amendment to the SAR and, if necessary, a change to the Certificate of 
Compliance. For other structures, systems, or components that are 
needed for the design to be used or are otherwise prudent, but do not 
perform a safety function and were not relied upon in the basis for 
design approval, appropriate changes may be permitted provided the 
licensee and the Certificate of Compliance holder document the 
appropriate evaluations and use their quality assurance programs to 
implement the change. New models of the VSC-24 casks are not the 
subject of this rulemaking.
    L.4. Comment. One commenter questioned how the draft Environmental 
Assessment and Finding of No Significant Impact would remain valid if 
changes to cask design and procedures can be made. Tests or experiments 
could be conducted under draft Certificate of Compliance Item No. 9 
(see also 10 CFR 72.48) leading to the use of a cask that does not meet 
the conditions specified in the Certificate of Compliance. These 
changes may adversely impact site-specific public health, safety, and 
the environment.
    Response. Given the limiting criteria of 10 CFR 72.48, it is 
unlikely that any change would materially change the environmental 
analysis. The licensee's authority under 10 CFR 72.48 does not permit 
any changes that involve unresolved safety issues, changes to the 
conditions for cask use in the Certificate of Compliance, significant 
increase in occupational exposure, or significant environmental impact. 
In the Environmental Assessment supporting this rulemaking to approve 
the Standardized NUHOMS, the NRC staff evaluated various types of 
accidents that could happen to the ISFSI facility. The NRC staff's 
evaluation encompassed design basis accidents and concluded that no 
radioactive material will be released to the environment. The NRC staff 
also evaluated a worst-case accident and found that the environmental 
impact is insignificant. Therefore, it is unlikely that the potential 
impact from changes to cask design or tests or experiments under the 
control of the licensee would introduce new environmental 
considerations or impacts that differ from or exceed those as analyzed 
in the Environmental Assessment. Changes in environmental impacts, as a 
result of changes to the cask design or procedures, must be evaluated 
by the licensee. The licensee's evaluations are available for 
inspection by the NRC.
    M. A number of technical clarifications and editorial issues were 
raised.
    M.1. Comment. One commenter stated that both the SAR and SER on 
which the Certificate of Compliance is based should be dated, as was 
the case for the VSC-24 Certificate of Compliance. If not, the public 
will be commenting on an unfinished document that can be endlessly 
revised.
    Response. Both the draft SER and the SAR are dated November 1993. 
These documents were revised based on public comments.
    M.2. Comment. One commenter wanted page one of the Certificate of 
Compliance revised to change the name ``Pacific Nuclear'' to 
``VECTRA''.
    Response. The Certificate of Compliance has been revised to reflect 
this.
    M.3. Comment. One commenter pointed out a typographical error on 
page A-19 of the draft Certificate of Compliance. In the Basis 
paragraph, the sentence starting, ``Acceptable damage may occur * * *'' 
should read ``Unacceptable damage may occur * * *''
    Response. The Certificate of Compliance has been revised to correct 
this.
    M.4. Comment. One commenter requested clarification of Technical 
Specification 1.2.16 on page A-25 of the draft Certificate of 
Compliance, as to whether the Yearly Average Ambient Temperature is a 
surveillance requirement or an action statement. It is unclear what 
action should be taken if either of the two specified limits (Yearly 
average temperature <70  deg.F or average daily ambient temperature 
<100  deg.F) is exceeded.
    Response. The Yearly Average Ambient Temperature specification is a 
site-specific parameter that the user must verify in accordance with 
the requirement of 10 CFR 72.212(b)(3) in order to use the system under 
the general license. There is no surveillance requirement or further 
action to be taken.
    Certificate of Compliance Section 1.1.1, ``Regulatory Requirements 
for General License,'' also includes verification of some of the same 
site-specific temperature parameters and has been amended to include 
the 100 deg.F or less average daily ambient temperature parameter. 
Therefore, this specification mentioned in the comment (draft 
Certificate of Compliance Section 1.2.16) was deleted.
    M.5. Comment. Apparently in reference to a December 4, 1991, letter 
from PNFSI that stated ``The NUHOMS Certification Safety Analysis 
Report (CSAR) was * * *,'' one commenter believed that the use of the 
term CSAR was a good idea and should have been used by the NRC. The 
utility SAR should be called SAR as it was and the vendor SAR should be 
called CSAR just as NUHOMS did in 1990. Also, the acronyms topical 
report (TR), TSAR, and SAR are being used interchangeably and they need 
clear definition. This would eliminate confusion on the issue by those 
involved.
    Response. The NRC staff generally agrees with the comment. However, 
the required documents that form the basis of the NRC staff's safety 
review are clearly identified in the SER and Certificate of Compliance.
    M.6. Comment. One commenter wanted the term ``certificate holder'' 
eliminated because it is ambiguous and misleading.
    Response. The term ``certificate holder'' has been changed to 
``holder of a Certificate of Compliance'' to be consistent with the 
regulations.
    M.7. Comment. One commenter wanted the draft Certificate of 
Compliance clarified as to who is responsible for the use of seismic 
restraints at each reactor site, the vendor or the utility, citing the 
ambiguous term ``certificate holder.''
    Response. The utility is responsible for determining the need for 
seismic restraints in the spent fuel building based on seismic 
conditions at the site (Certificate of Compliance, Section 1.2.17).
    M.8. Comment. Several commenters stated that the limits on both 
neutron and gamma emission rates as well as neutron and gamma spectra 
(Attachment A, Section 1.2.1 of draft Certificate of Compliance) result 
in excluding some fuel assemblies that would actually produce lower 
dose rates. The problem for fuel qualification stems from the fact that 
the neutron dose rate does not decrease as rapidly as the gamma dose 
rate during cooling because of the longer lived isotopes. Thus, a high 
burned fuel assembly excluded on the basis of high neutron source term 
may remain excluded, even though with extra cooling time the combined 
neutron/gamma dose rate could be less than the design basis case. Some 
fuel may not qualify because it exceeds the spectra requirements, even 
though the energy groups exceeding the limits may not be significant 
contributors to the dose rates. Combined neutron/gamma dose rates are 
the real concern; it is recommended that the limits on source term be 
replaced by limits based on dose equivalence. The fuel specification 
should allow other combinations of fuel enrichment, burnup, and cooling 
time that would not result in exceeding the fuel cladding temperature 
or dose rates.
    Response. The NRC staff agrees that alternative fuel specifications 
could be beneficial. However, this commenter did not provide a specific 
alternative, and the NRC staff has not evaluated any other alternative 
at this time because VECTRA did not include this approach in the SAR. 
Therefore, no other approach is considered for this rulemaking.
    M.9. Comment. One commenter suggested wording changes to the draft 
Certificate of Compliance in Attachment A, Section 1.2.6, Action b, as 
follows: ``Visually inspect placement of top shield plug. Re-install or 
adjust position of top shield plug if it is not properly seated.'' The 
commenter also proposed wording changes to Action c of the same section 
as follows: ``Install additional temporary shielding or implement other 
ALARA actions, as appropriate.''
    Response. The NRC staff agrees with the first comment and has added 
the suggested words to the Certificate of Compliance, Section A.1.2.6, 
Action b. It is not necessary to change Action c because 10 CFR Part 20 
ALARA already applies to these activities.
    M.10. Comment. One commenter wanted draft Certificate of 
Compliance, Attachment A, Section 1.2.6, Action d deleted. The user 
should be permitted to analyze and document higher dose rates under 10 
CFR 72.48, which is available for NRC review. Another commenter wanted 
the complete Section 1.2.6 of Attachment A to the draft Certificate of 
Compliance deleted. Given that HSM dose rates are specified, a 
specification for DSC dose rates is not necessary because only the 
workers involved in the canister closure operations are affected by 
them and they are already covered by the reactor radiation protection 
program. One commenter wanted draft Certificate of Compliance, 
Attachment A, Section 1.2.11 deleted. Given that HSM dose rates are 
specified, a specification for transfer cask dose rates is not 
necessary because only the workers involved are affected, not the 
general public. The commenter also stated that if Section 1.2.11 cannot 
be deleted the action statement should be revised to read as follows: 
``If specified dose rates are exceeded, place temporary shielding 
around the affected areas of the transfer cask or implement other ALARA 
actions, as appropriate. Review the plant records of the fuel 
assemblies which have been placed in the DSC to ensure they conform to 
the fuel specifications of Section 1.2.1. The report to the NRC should 
be deleted with the user being able to analyze and document the higher 
dose rates under 10 CFR 72.48, which is available for NRC review.''
    Response. The dose rate limits are for design purposes. The dose 
rate is limited to ensure that the DSC has not inadvertently been 
loaded with fuel not meeting the vendor/applicant spent fuel 
specifications. The NRC will require reporting if the specified dose 
limits are exceeded. For these reasons, the NRC will not grant the 
above requests.
    M.11. Comment. One commenter stated that the requirement for a 
dissolved boron concentration in the DSC of 2000 ppm is in excess of 
the 1810 ppm site-specific license. The 1810 ppm dissolved boron is 
sufficient to ensure reactivity below 0.95 K-eff (95/95 tolerance level 
with uncertainties) assuming 24 fresh fuel assemblies. For the unlikely 
worst case with water density of 0.2 to 0.7 gm/cc (a condition not 
achievable for fresh fuel), reactivity remains below 0.98 K-eff. The 
pool dissolved-boron verification- measurement frequency should be 
changed from not to exceed 48 hours to once per month to be consistent 
with 10 CFR Part 50 requirements.
    Another commenter stated that the NUHOMS-24P canister was designed 
using burnup credit, the basis for licensing is ``credit for soluble 
boron.'' The burnup-enrichment curve requirement (Figure 1-1, draft 
Certificate of Compliance) should be removed until the NRC accepts 
burnup credit and the pool boron specification (Section 1.2.15, draft 
Certificate of Compliance) is removed.
    The NRC has not yet approved the use of burnup credit in 
criticality analyses for spent fuel storage and transportation casks. 
The applicant did, however, analyze credit for burnup as an alternative 
design acceptance basis for the NUHOMS-24P DSC, pending further 
consideration of burnup credit by NRC. As discussed in the SER, the 
NUHOMS-24P DSC criticality safety is approved based on, among others, 
the key assumptions of loading with irradiated fuel assemblies with 
equivalent enrichment <1.45 wt% U-235, misloading unirradiated fuel 
with maximum enrichment of 4.0 wt% U-235, and soluble boron in water 
for wet loading and unloading. The NRC considered the use of the 
burnup- enrichment curve, Certificate of Compliance, Figure 1-1, as a 
fuel selection criteria, to be prudent. Its use adds additional 
unanalyzed conservatism in the criticality safety margin. It is 
comparable to previous NUHOMS-24P approvals. Its use would also be 
consistent with the requirement that storage cask designs be, to the 
extent practicable, compatible with removal of the stored spent fuel 
from the reactor site, transportation, and ultimate disposition by DOE. 
Therefore, the NRC disagreed with the commenter's request to allow 
Standardized NUHOMS-24P users the option of using these burnup- 
enrichment curve.
    Response. The comment appears to refer to the use of a NUHOMS 24P 
associated with a site-specific license. The ``standardized NUHOMS 24P 
and 52B'' are the subject of this general rulemaking and should not be 
confused with a site license. The SER for this rulemaking is clear 
about conditions for use, i.e., 2000 ppm boron concentration is 
required to ensure that the keff remains below 0.95. The SAR for 
this rulemaking does not request, nor does the SER grant, exemption 
from the requirement of keff = 0.95 for all accident conditions, 
including misloading of 24 unirradiated fuel assemblies and optimum 
moderation density.
    The NRC has not yet approved the use of burnup credit in 
criticality analyses for spent fuel storage and transportation casks. 
The applicant did, however, analyze credit for burnup as an alternative 
design acceptance basis for the NUHOMS-24P DSC, pending future 
acceptance of burnup credit by NRC. As discussed in the SER, the 
NUHOMS-24P DSC criticality safety is approved based on, among other 
assumptions, the key assumptions of loading with irradiated fuel 
assemblies with equivalent enrichment <1.45 wt% U-235, misloading 
unirradiated fuel with maximum enrichment of 4.0 wt% U-235, and soluble 
boron in water for wet loading and unloading. The NRC still considers 
the use of the burnup-enrichment curve, Certificate of Compliance 
Figure 1-1, as a fuel selection criteria, to be prudent. Its use adds 
additional unanalyzed conservatism in the criticality safety margin. It 
is comparable to previous NUHOMS-24P approvals. Its use would also be 
consistent with the requirement that storage cask designs be, to the 
extent practicable, compatible with removal of the stored spent fuel 
from the reactor site, transportation, and ultimate disposition by DOE. 
Therefore, the NRC disagrees with the commenters request to allow 
Standardized NUHOMS-24P users the option of using the burnup-enrichment 
curve.
    M.12. Comment. Several commenters stated that the listing of 
specific fuel types in the draft Certificate of Compliance is overly 
restrictive. Allowance should be made for very similar fuel types or a 
``fuel qualification table'' as proposed by the vendor should replace 
the listing.
    Response. The NRC agrees that allowance should be made for very 
similar types of fuel to be stored. The Certificate of Compliance 
provides this flexibility. The ``fuel qualification table'' 
consideration at this time is not subject to this rulemaking.
    M.13. Comment. One commenter citing the first paragraph of page A-
27 of the draft Certificate of Compliance states that the postulated 
adiabatic heatup would result in concrete temperatures being exceeded 
in approximately 40 hours. As a result, it is appropriate and 
conservative to perform the visual surveillance to verify no vent 
blockage on a daily basis to ensure that a blockage existed for less 
than 40 hours. The last sentence in the first paragraph should reflect 
that the module needs to be removed from service if it cannot be 
established that the blockage is less than 40 hours, not 24 hours. A 
24-hour surveillance interval will adequately verify this. One 
commenter cited an inconsistency in Section 3 of the draft Certificate 
of Compliance. Section 3.1 indicates that a module must be removed from 
service if a vent blockage is in existence for longer than 24 hours. 
Surveillance Section 1.3.2 indicates that a module must be removed from 
service if the concrete accident temperature criterion has been 
exceeded for more than 24 hours. A vent blockage of less than 24 hours 
would not cause the temperature limit to be exceeded, as explained in 
Section 1.3 and the objective for the 24-hour frequency required by 
surveillance 1.3.1. The apparent conflict between Section 1.3 and the 
action for Surveillance Requirement 1.3.2 should be resolved. It 
appears that Surveillance Requirement 1.3.2 actions are appropriate.
    Response. The Certificate of Compliance has been clarified to 
reflect the comment.
    M.14. Comment. One commenter stated that Section 1.2.14 to 
Attachment A of the draft Certificate of Compliance is unnecessary 
because the time to transfer the DSC from the transfer cask to the HSM 
would normally require less than 8 hours. During this time, even with 
temperatures above 100 deg.F without the solar shield, any increase in 
fuel clad temperature and neutron shield temperature would be small and 
therefore not detrimental. Additionally, the transfer cask is open to 
the atmosphere and would not pressurize.
    Response. The vendor, VECTRA, has proposed this limiting condition 
of operation in lieu of showing what detrimental effect might occur on 
the cladding or neutron shield, should the ambient conditions involve 
temperatures above 100 deg.F. The NRC concurs with this condition as 
cited in Attachment A, Section 1.2.14 of the Certificate of Compliance.
    N. Several commenters raised safeguards/sabotage issues.
    N.1. Comment. One commenter cited the World Trade Center bombing 
and the ease with which a disturbed individual recently breached 
security and remained undetected at a U.S. reactor. Explosive 
technology has become very sophisticated in the last 15 years since the 
NRC and Sandia Laboratories studied the effect of sabotage on shipping 
casks in the March 1979, NUREG-0459, ``Generic Adversary 
Characteristics Summary Report.'' Another commenter made reference to 
an experiment with balloons which failed. Yet another commenter 
questioned the degree of protection in the spent fuel pool versus dry 
cask storage. Will the cask be in a vital area? Will safeguards be 
reviewed as part of the security plan? What is the effect on the 
security of these casks?
    Response. The NRC reviewed potential issues related to possible 
radiological sabotage of storage casks at reactor site ISFSIs in the 
1990 rulemaking that added Subparts K and L to 10 CFR Part 72 (55 FR 
29181; July 18, 1990). NRC regulations in 10 CFR Part 72 establish 
physical protection and security requirements for an ISFSI located 
within the owner controlled area of a licensed power reactor site. 
Spent fuel in the ISFSI is required by 10 CFR 72.212(b)(5) to be 
protected against the design basis threat for radiological sabotage 
using provisions and requirements as specified in 72.212(b)(5). Each 
utility licensed to have an ISFSI at its reactor site is required to 
develop security plans and install a security system that provides high 
assurance against unauthorized activities that could constitute an 
unreasonable risk to the public health and safety. The security systems 
at an ISFSI and its associated reactor are similar in design features 
to ensure the detection and assessment of unauthorized activities. 
Alarm annunciations at the ISFSI are monitored by the security alarm 
stations at the reactor site. Response to intrusion is required. Each 
ISFSI is periodically inspected by NRC and annually audited by the 
licensee to ensure that the security systems are operating within their 
design limits. The validity of the threat is continually reviewed, with 
a formal evaluation every six months by the NRC.
    The NRC is currently conducting a study into the consequences of a 
vehicle bomb detonated in the vicinity of an ISFSI. Following 
completion of this study the NRC will make a determination as to 
whether additional physical protection is warranted. In the interim, 
the NRC staff believes that the inherent nature of the fuel, along with 
the degree of protection provided by the approved storage means for 
spent fuel, provides adequate protection against a vehicle bomb.
    N.2. Comment. One commenter wanted the emergency plan updated to 
include initiating events caused by unnatural occurrences, such as 
sabotage, particularly for this fuel storage option. The commenter 
believes that the NRC should determine if upgraded or new security 
barriers are necessary for the David-Besse site.
    Response. Under 10 CFR 72.212 requirements, each general licensee 
must protect the spent fuel against the design basis threat of 
radiological sabotage. Also, 10 CFR 72.212 requires each general 
licensee to review the reactor emergency plan to determine whether its 
effectiveness is decreased, and if so, to prepare the necessary changes 
and obtain the necessary approvals. Therefore, the comment is already 
essentially incorporated into NRC regulations.
    O. Several commenters had fabrication, quality assurance, and 
inspection concerns.
    O.1. Comment. One commenter raised questions about NRC oversight 
and requirements for proper cask fabrication by licensees. This is 
based on tests of the faulty welds at the Palisades plant conducted in 
July 1994 just before the cask was filled, but the test was not 
reviewed.
    Response. The ultimate responsibility to ensure proper cask 
fabrication belongs to the user of the cask. Each Part 50 licensee 
(general licensee) must have its own quality assurance (QA) program in 
place to oversee vendor activities. The QA requirements apply to 
design, purchase, fabrication, handling, inspection, testing, 
operation, maintenance, repair, modifications of structures, systems 
and components, and decommissioning that are important to safety. In 
addition, certified cask vendors have NRC-approved QA programs that 
control the implementation of these quality activities in a manner 
appropriate to the safety significance of these activities. In turn, 
the general licensee reviews, approves, and oversees its vendor's QA 
programs and activities. The NRC inspects both the general licensee and 
the subtiered vendors for compliance with the respective QA program 
requirements and for the adequacy of the activities performed.
    The faulty welds at Palisades in a loaded cask happened because the 
radiographs were not read initially. If the radiographs were read in a 
timely manner, the cask should not have been loaded without corrective 
action first being taken. NRC oversight and involvement in the process 
contributed to timely detection of the defective cask weld.
    O.2. Comment. One commenter wants clarification of the quality 
assurance program. NRC should have a regulatory guide for vendors with 
strong criteria for audits and subcontractors, and NRC inspection 
reports of fabricating facilities need to be put in the PDR. How will a 
subcontractor of NUHOMS vendor be checked by NRC in the future? If a 
vendor is going to continuously change subcontractors, the NRC should 
inspect each cask and carefully inspect the vendor QA manual.
    Response. Chapters 11 or 13 of Regulatory Guides 3.62 and 3.61, 
respectively, provide guidance on acceptable quality assurance 
programs. These chapters state that a QA program meeting the 
requirements of Appendix B of 10 CFR Part 50 or Subpart G of 10 CFR 
Part 72 will be accepted by NRC. Both Parts 50 and 72 require an audit 
program. An NRC Branch Technical Position titled ``Quality Assurance 
Programs for an Independent Spent Fuel Storage Installation (ISFSI) 10 
CFR 72,'' implements the NRC review of quality assurance programs 
submitted by applicants. NRC inspection reports are routinely placed in 
the PDR except for reports containing sensitive information. Inspection 
reports of NUHOMS fabrication are available in the PDR.
    O.3. Comment. One commenter wanted to know if any nonconformances 
have been discovered in inspection reports of any fabrication of the 
NUHOMS canister. If so, what? How was this resolved? How has the QA 
program for NUHOMS been reviewed? Is there a manual? How will 
contractors and subcontractors be checked?
    Response. A notice of nonconformance is documented in NRC 
Inspection Report No. 721004/93-07 dated August 23, 1993. The NRC staff 
conducted inspections in three phases at Duke Power Company, its 
contractor (Pacific Nuclear Fuel Services, Inc.) and subcontractor 
(Rancor, Inc.), concerning the QA activities with regard to the NUHOMS-
24P dry spent fuel storage canisters. The NRC staff found that 
implementation of Duke Power Company QA Program was satisfactory, in 
general. However, certain NRC requirements under Subpart G of 10 CFR 
Part 72 were not met. QA activities cited in the inspection report were 
documentation of nonconforming materials, parts, or components; quality 
assurance records; control of purchased material, equipment, and 
services; control of measuring and test equipment; instructions, 
procedures, and drawings; licensee inspection; and audits. 
Nonconformance corrective actions were taken and documented by Duke 
Power Company. The NRC staff found these corrective actions acceptable 
and so stated in letters dated January 13, 1994, and April 4, 1994. The 
corrective actions taken and the implementation of the QA Program are 
reviewed in periodic inspections by the NRC staff.
    The latest version of the QA manual is ``VECTRA Technologies, Inc., 
Quality Assurance Manual,'' Revision 1, transmitted July 25, 1994, 
which reflects the corporation's new name and organization and includes 
additional changes to update the manual and clarify QA recordkeeping 
commitments. The NRC staff found Revision 1 acceptable and so stated in 
its letter dated August 23, 1994. In its review, the NRC staff compared 
Revision 1 of the VECTRA QA Manual with Revision 3, Edition 2, of the 
PNSI QA manual, which the NRC staff found acceptable by letter dated 
January 28, 1993.
    Contractors and subcontractors of cask vendors (or licensees) are 
subject to periodic QA inspections performed by the NRC staff.
    O.4. Comment. One commenter wanted to know if there is a possible 
problem, and if there was, how it was resolved, with a material defect 
in Swagelok tube fittings for NUHOMS?
    Response. The NRC is not aware of any material defect problem with 
Swagelok tube fittings on NUHOMS designs. There is no reliance on the 
Swagelok fittings as part of the confinement boundary for the NUHOMS 
canister. The fittings are covered by a metal plate that is welded on 
after the canister is vacuum dried. Therefore, if there is a failure in 
the fitting it would be the responsibility of the licensee to repair or 
replace it so that the DSC can be loaded properly, but its failure 
would not cause a public health and safety concern.

Finding of No Significant Environmental Impact: Availability

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
Subpart A of 10 CFR Part 51, 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 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 environmental assessment and finding of no significant 
impact on which this determination is based are available for 
inspection at the NRC Public Document Room, 2120 L Street NW. (Lower 
Level), Washington, DC. Single copies of the environmental assessment 
and finding of no significant impact are available from Mr. Gordon E. 
Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington DC, 20555, telephone (301) 415-6195.

Paperwork Reduction Act Statement

    This final rule does not contain a new or amended information 
collection requirement subject to the Paperwork Reduction Act of 1980 
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the 
Office of Management and Budget approval number 3150-0132.

Regulatory Analysis

    The Commission has prepared a regulatory analysis on this 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the Commission. Interested persons may 
examine a copy of the regulatory analysis at the NRC Public Document 
Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of 
the analysis may be obtained from Mr. Gordon E. Gundersen, Office of 
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 
Washington DC, 20555, telephone (301) 415-6195.

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. 
This rule affects only licensees owning and operating nuclear power 
reactors and cask vendors. The owners of nuclear power plants do not 
fall within the scope of the definition of ``small entities'' set forth 
in Section 601(3) of the Regulatory Flexibility Act, 15 U.S.C. 632, 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 rules 10 CFR 50.109 and 10 
CFR 72.62 do not apply to this final rule. A backfit analysis is not 
required for this final rule because this amendment does not involve 
any provisions that would impose backfits as defined in 10 CFR 
50.109(a)(1) or 72.62(a).

List of Subjects in 10 CFR Part 72

    Manpower training programs, Nuclear materials, Occupational safety 
and health, Reporting and recordkeeping requirements, Security 
measures, Spent fuel.

    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 AND HIGH-LEVEL RADIOACTIVE WASTE

    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. 2902, 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).

    2. In Sec. 72.214, Certificate of Compliance 1004 is added in 
numerical order to read as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
Certificate Number: 1004
SAR Submitted by: VECTRA Technologies, Inc.
SAR Title: Safety Analysis Report for the Standardized NUHOMS 
Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 
2
Docket Number: 72-1004
Certification Expiration Date: (20 years after final rule effective 
date)
Model Numbers: NUHOMS-24P for Pressurized Water Reactor fuel; NUHOMS-
52B for Boiling Water Reactor fuel.
* * * * *
    Dated at Rockville, Maryland this 15th day of December, 1994.

    For the Nuclear Regulatory Commission.
James M. Taylor,
Executive Director for Operations.
[FR Doc. 94-31307 Filed 12-21-94; 8:45 am]
BILLING CODE 7590-01-P