# List of Approved Spent Fuel Storage Casks: Addition

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## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** December 22, 1994

## Text

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 st

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A94-31307. Public record. Not legal advice.