List of Approved Spent Fuel Storage Casks: Addition

NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AF02

List of Approved Spent Fuel Storage Casks: Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its

regulations to add the Standardized NUHOMS Horizontal Modular System to

the List of Approved Spent Fuel Storage Casks. This amendment allows

the holders of power reactor operating licenses to store spent fuel in

this approved cask under a general license.

EFFECTIVE DATE: January 23, 1995.

ADDRESSES: Copies of the environmental assessment and finding of no

significant impact, as well as, the public comments received on the

proposed rule are available for inspection and/or copying for a fee at

the NRC Public Document Room, 2120 L Street, NW. (Lower Level),

Washington, DC. Single copies of the environmental assessment and the

finding of no significant impact are available from the individuals

listed under the next heading below.

FOR FURTHER INFORMATION CONTACT: Mr. Gordon E. Gundersen, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, telephone (301) 415-6195, or Dr. Edward Y. S.

Shum, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear

Regulatory Commission, Washington, DC 20555, telephone (301) 415-7903.

SUPPLEMENTARY INFORMATION:

Background

Section 218(a) of the Nuclear Waste Policy Act of 1982 (NWPA)

includes the following directive: ``The Secretary [of the Department of

Energy (DOE)] shall establish a demonstration program in cooperation

with the private sector, for the dry storage of spent nuclear fuel at

civilian nuclear reactor power sites, with the objective of

establishing one or more technologies that the [Nuclear Regulatory]

Commission may, by rule, approve for use at the sites of civilian

nuclear power reactors without, to the maximum extent practicable, the

need for additional site-specific approvals by the Commission.'' After

subsequent DOE technical evaluations and based on a full review of all

available data, the Commission approved dry storage of spent nuclear

fuel in a final rule published in the Federal Register on July 18, 1990

(55 FR 29181). The final rule established a new Subpart K within 10 CFR

Part 72 entitled ``General License for Storage of Spent Fuel at Power

Reactor Sites.''

Irradiated reactor fuel has been handled under dry conditions since

the mid-1940s when irradiated fuel examinations began in hot cells.

Light-water reactor fuel has been examined dry in hot cells, since

approximately 1960. Irradiated reactor fuel has been stored

continuously at hot cells under dry conditions for approximately two

decades. The NRC's experience with storage of spent fuel in dry casks

is extensive as described in the proposed rule to establish 10 CFR Part

72, Subpart K (May 5, 1989; 54 FR 19379). Further, the United States

has extensive experience in the licensing and safe operation of

independent spent fuel storage installations (ISFSIs). By the end of

1994, six site-specific licenses for dry cask storage will have been

issued: Virginia Power's Surry Station, issued July 2, 1986; Carolina

Power and Light's (CP&L) HB Robinson Station, issued August 13, 1986;

Duke Power's Oconee Station, issued January 29, 1990; Public Service of

Colorado's Fort St. Vrain facility, issued November 4, 1991; Baltimore

Gas and Electric's (BG&E) Calvert Cliffs Station, issued November 25,

1992; and Northern States Power's (NSP) Prairie Island Nuclear

Generating Plant, issued October 19, 1993. All except NSP have

commenced operation and loaded fuel. At the end of 1994, dry storage

spent fuel inventories of these utilities are as follows: 500

assemblies at Virginia Power, 60 assemblies at CP&L, 530 assemblies at

Duke Power, 1480 fuel elements at Public Service of Colorado, and 190

assemblies at BG&E. NSP plans to begin storing fuel soon. In May 1993,

Consumers Power's Palisades Station commenced operation and loaded fuel

under the provisions of the general license in 10 CFR Part 72, Subpart

K. At the end of 1994, approximately 168 assemblies are stored at

Palisades.

As a result of the growing use of dry storage technology, the NRC

has gained over 35 staff years of experience in the review and

licensing of dry spent fuel storage systems. In addition, the NRC draws

upon the knowledge and experience of outside scientists and engineers

recognized as experts within their respective fields in the performance

of the independent safety analysis of the system and component designs

submitted by applicants for dry cask licenses or certification. Reviews

of numerous applications seeking site- specific licenses, certificates

of compliance, or approvals of topical reports, have been conducted

over the past eight years. More recently, the NRC published a notice of

proposed rulemaking in the Federal Register on June 2, 1994 (59 FR

28496), which proposed to amend 10 CFR 72.214 to include one additional

spent fuel storage cask (i.e., the VECTRA Technologies, Inc.,

Standardized NUHOMS Horizontal Modular Storage System) on the list of

approved spent fuel storage casks that power reactor licensees may use

under the provisions of a general license issued by NRC in accordance

with 10 CFR Part 72, Subpart K. The Standardized NUHOMS consists of two

systems: (1) The NUHOMS-24P holds 24 specified pressurized-water

reactor spent fuel assemblies and (2) The NUHOMS-52B holds 52 specified

boiling-water reactor spent fuel assemblies.

Subsequent to the expiration of the 75-day public comment period on

August 16, 1994, NRC received a request, dated August 11, 1994, for a

6-week extension of the comment period from Connie Kline of the Sierra

Club on behalf of 12 citizen groups. The extension request asserted

that several proprietary documents related to this rulemaking were not

available to the public for approximately 2 weeks at the beginning of

the comment period. The NRC granted the request on August 29, 1994 (59

FR 44381) by extending the public comment period to September 30, 1994.

VECTRA Technologies, Inc. (formerly Pacific Nuclear Fuel Services,

Inc.) submitted to the NRC a Safety Analysis Report (SAR) entitled

``Safety Analysis Report for the Standardized NUHOMS Horizontal Modular

Storage System for Irradiated Nuclear Fuel,'' NUH-003, Revision 2,

dated November 1993. Subsequently, VECTRA Technologies, Inc. provided

additional information to the NRC related to the SAR. In March 1994,

the NRC issued a draft Safety Evaluation Report (SER) entitled ``Safety

Evaluation Report of Pacific Nuclear Fuel Services, Inc. Safety

Analysis Report for the Standardized NUHOMS Horizontal Storage System

for Irradiated Nuclear Fuel'' approving the SAR. The NRC issued a draft

Certificate of Compliance by letter to Mr. Robert D. Quinn from Mr.

Frederick C. Sturz dated April 28, 1994. These documents are part of

the docket and record that support the proposed rule published in the

Federal Register on June 2, 1994 (59 FR 28496).

The objective of 10 CFR Part 72 is to protect the public health and

safety by providing for the safe confinement of the stored fuel and

preventing the degradation of the fuel cladding. The review criteria

used by the NRC for review and approval of dry cask storage under 10

CFR Part 72 consider the following factors: siting, design, quality

assurance, emergency planning, training, and physical protection of the

fuel. Phenomena such as earthquakes, high winds, tornados, tornado

driven missiles, lightning, and floods are included in the review of a

specific system, either for a certificate of compliance or a site-

specific license. In addition, applicants must demonstrate to NRC's

satisfaction that their proposed dry cask system will resist man-made

events such as explosions, fire, and drop or tipover accidents.\1\

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\1\The design bases for these events and accidents are contained

within 10 CFR Part 72.

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Based on further NRC review and analysis of public comments, both

the SER and Certificate of Compliance for the Standardized NUHOMS were

modified. Section M contains a description of changes to the SER and

Certificate of Compliance in response to public comments. The NRC finds

that the Standardized NUHOMS, as designed and when fabricated and used

in accordance with the conditions specified in its Certificate of

Compliance, meets the requirements of 10 CFR Part 72. Thus, use of the

Standardized NUHOMS, as approved by the NRC, will provide adequate

protection of the public health and safety and the environment. With

this final rule, the NRC is approving the use of the Standardized

NUHOMS under the general license in 10 CFR Part 72, Subpart K, by

holders of power reactor operating licenses under 10 CFR Part 50.

Simultaneously, the NRC is issuing a final Certificate of Compliance to

be effective on January 23, 1995. A copy of the Certificate of

Compliance is available for public inspection and/or copying for a fee

at the NRC Public Document Room, 2120 L Street, NW. (Lower Level),

Washington, DC.

Public Responses

In response to the proposed addition of the Standardized NUHOMS,

239 comments in 27 letters with one supplement were received from

individuals, public interest groups, an environmental group, an

association, industry representatives, a city, states, and one Federal

agency. One commenter withdrew his comments. Many of these letters

contained similar comments that have been grouped together and

addressed as a single issue. All comments have been grouped into 15

broad issues designated A through O. A summary of the comments and an

NRC analysis and response to those comments is included for each broad

issue. The NRC has identified and responded to 89 separate issues that

include the significant points raised by each commenter.

A number of comments were related to the disposal of high-level

waste and the use of dry cask storage technology in general, rather

than to the acceptability of this particular cask. Examples of these

comments include:

--The Federal Government's failure to resolve questions about the

permanent storage of nuclear waste leaves both the plant and public

with limited options: additional storage in pools, additional storage

in dry casks, or plant shutdown. The Federal Government has an

obligation to resolve the issue of permanent or interim storage. It

would be difficult to overstate the need for dispatch in doing so, as

hundreds of American communities will eventually face this problem.

--It is not fair to the public of Ohio to link Toledo Edison Company's

attempts to continue the safe storage of its nuclear fuel with

insistence by others that the NRC shut down Davis-Besse and every other

nuclear plant in the country.

--Only dry storage casks that are compatible with future DOE interim or

permanent storage operation, including transportation, should be

approved for use under the general license and listed in 10 CFR 72.214.

These comments deal with broad policy and program issues relating

to the storage and disposal of high-level radioactive waste, including

the DOE's repository program and as such are beyond the scope of this

rule. However, there is a summary of relevant information on many of

these broad issues in Group G. Many comments were directed at the

Standardized NUHOMS-24P with only a few comments being specific to the

Standardized NUHOMS-52B.

Many commenters discussed topics that were not the subject of this

rulemaking and thus were not specifically addressed by the NRC staff as

a part of this final rule action. These comments express opposition to

the use of dry cask storage and included the following suggestions and

topics:

(1) Nuclear plants generating radioactive waste should be shut

down.

(2) The production of radioactive waste should be stopped when the

existing spent fuel pool (and off-load-reactor capacity) is full.

(3) A formal hearing should be required at each site using dry

storage casks.

(4) The Davis-Besse plant should be shut down.

(5) The use of nuclear power should be stopped and existing sites

cleaned up.

(6) Palisades experienced problems in using the VSC-24 cask.

(7) Alternative forms of power should be used.

Finally, many commenters expressed concern over the ability of dry

cask storage designs, presumably including the Standardized NUHOMS, to

store spent fuel safely. The following responses to these comments

reflect a small but important portion of the NRC's review of health,

safety, and environmental aspects of the Standardized NUHOMS to ensure

that the cask is designed to provide protection of the public health

and safety and environment under both normal conditions and severe,

unlikely but credible, accident conditions. Dry cask storage systems

are massive devices, designed and analyzed to provide shielding from

direct exposure to radiation, to confine the spent fuel in a safe

storage condition, and to prevent releases of radiation to the

environment. They are designed to perform these tasks by relying on

passive heat removal and confinement systems without moving parts and

with minimal reliance on human intervention to safely fulfill their

function for the term of storage. The NRC staff has concluded that the

methods of analysis are conservative and assure that the design has

appropriate margins of safety under both normal and accident

conditions.

Analysis of Public Comment

A. A number of commenters raised issues relating to cask handling

and the ability of the cask to withstand drop and tipover accidents.

A.1. Comment. Several commenters wanted the transfer cask

containing the Dry Storage Canister (DSC) to be analyzed for the

maximum possible drop, regardless of whether that drop can occur inside

or outside the spent fuel building. One commenter alleged that a drop

of the transfer cask into the spent fuel pool would damage fuel

assemblies in the pool. Another commenter was concerned about jamming

the transfer cask in the spent fuel pool. What would happen to the cask

if jammed fuel could not be extricated? Would the entire 40 ton

transfer cask be left in the fuel pool?

Response. Use of the Standardized NUHOMS inside the fuel handling

building would be conducted in accordance with the 10 CFR Part 50

reactor operating license. These cask handling operations, including

loading, retrieval, and training, must be evaluated by the general

licensee as required by 10 CFR 72.212(b)(4) to ensure that procedures

are clear and can be conducted safely. Load handling activities and

possible load drop events with structural and radiological consequences

related to transfer cask drops inside the spent fuel building are

subject to the provisions of 10 CFR 50.59. Thus, the licensee must

determine whether the activities involve any unreviewed facility safety

question or any change in facility technical specifications. The

transfer cask and DSC designs were evaluated by the NRC against the

criteria for controlling heavy loads that are found in NRC's NUREG-

0612,\2\ ``Control of Heavy Loads at Nuclear Power Plants,'' and

American National Standards Institute (ANSI) N14.6, ``Special Lifting

Devices for Shipping Containers Weighing 10,000 Pounds or More.'' The

lifting yoke associated with the transfer cask is a special purpose

device designed to ANSI N14.6 criteria to ensure that the yoke can

safely lift the wet transfer cask containing the DSC out of the spent

fuel pool and can safely lift the dry transfer cask and DSC to the

transport trailer. Pursuant to 10 CFR 50.59, for those reactor plants

with a shipping cask drop analysis, the licensee must verify that the

shipping cask drop analysis adequately describes the consequences of a

postulated transfer cask drop and that no unreviewed safety question

exists. For those reactor plants that may lack a shipping cask drop

analysis, each licensee must perform a transfer cask drop analysis

pursuant to 10 CFR 50.59.

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\2\Copies of NUREG-0612 and NUREG/CR 1815 may be purchased from

the Superintendent of Documents, U.S. Government Printing Office,

Mail Stop SSOP, Washington, DC 20402-9328. Copies are also available

from the National Technical Information Service, 5285 Port Royal

Road, Springfield, VA 22161. A copy is also available for inspection

and copying for a fee in the NRC Public Document Room, 2120 L

Street, NW (Lower Level), Washington, DC 20555-0001.

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Specific requirements for lifting the transfer cask are contained

in the Certificate of Compliance and SER. However, movement of the

transfer cask in the spent fuel pool building must, as required by 10

CFR 72.212(b)(4), be evaluated by the licensee pursuant to 10 CFR

50.59. The possibility of jamming a transfer cask while in the spent

fuel pool is one of many issues to be evaluated under 10 CFR 50.59.

A.2. Comment. One commenter asked why the transfer cask with the

DSC can be lifted to 80 inches outside the spent fuel pool building

when it has to be unloaded and inspected for damage if it drops from

above 15 inches. Why not limit the height to 15 inches?

Response. The transfer cask with the DSC rides on the transport

trailer at a height of greater than 15 inches and therefore was

analyzed for a drop from that height (80 inches). A drop from a height

between 15 and 80 inches does not pose a public health and safety

hazard. However, to ensure safety the NRC requires the DSC to be

unloaded and inspected for damage.

A.3. Comment. One commenter asked about the tipover analysis or

drop analysis result.

Response. The tipover, end drops, and horizontal drop analyses form

part of the structural design basis for the Standardized NUHOMS design.

The designer, VECTRA, described these drop and tipover analyses in SAR,

Section 8.2.5. The NRC's evaluation of the vendor's analyses is

described in SER, Section 3.2.2.3E. The NRC found the results of these

analyses to be satisfactory, because the calculated stresses were all

within the allowable criteria of the American Society of Mechanical

Engineers (ASME) Code.

A.4. Comment. Several commenters, citing Section 1.1.1 of the draft

Certificate of Compliance, requested that the postulated cask drop

accident in the plant fuel handling area be included in the list of

parameters and analyses that will need verification by the system user

(for the 10 CFR 50.59 safety evaluation).

Response. As stated in Section 1.1.1 of the draft Certificate of

Compliance, a holder of a 10 CFR Part 50 license before use of the

general license under 10 CFR Part 72, must determine whether activities

related to storage of spent fuel involve any unreviewed facility safety

issues or changes in facility technical specifications as provided

under 10 CFR 50.59. Fuel handling including the possible drop of a

spent fuel cask is among the activities that are required to be

verified. Fuel handling operations, including spent fuels and fresh

fuels, are routine within the nuclear power plant and are subject to

NRC regulation under 10 CFR Part 50. A holder of a 10 CFR Part 50

license is required to establish operating procedures for spent fuel

handling and to provide emergency planning to address a potential cask

drop accident in the reactor's fuel handling area (Certificate of

Compliance, Section 1.1.4). Therefore the NRC considers it clear that

the spent fuel operation in the nuclear power plant should be evaluated

to verify that the possible drop of a spent fuel cask does not raise an

unreviewed safety issue or require a facility technical specification

change appropriately regulated under 10 CFR Part 50.

A.5. Comment. One commenter stated that there is no place to unload

the spent fuel in the event of a canister breach. There is no

indication that the canister, the canister lifting mechanism, or the

transport mechanism to move the canister into the cask, are nuclear

grade equipment or have been designed to prevent a single failure from

breaching the canister and circumventing the protection provided by the

sole barrier provided by the canister wall itself.

Response. According to 10 CFR 72.122(1), storage systems must be

designed to allow ready retrieval of the spent fuel in storage. A

general licensee using an NRC-approved cask must maintain the

capability to unload a cask. Typically, this will be done by

maintaining the capability to unload a cask in the reactor fuel pool.

Other options are under consideration that would permit unloading a

cask outside the reactor pool.

With respect to canister equipment and design, the DSC or canister

is designed to the ASME Boiler and Pressure Vessel Code (BPVC), Section

III, Subsection NB. The DSC provides a containment boundary for the

radioactive material and the cladding of the fuel rods provides

confinement of fuel pellets. Only intact fuel assemblies (rods) with no

known cladding defects greater than pin holes and hairline cracks are

permitted to be stored. This approach assures the structural integrity

of the fuel to confine the fuel pellets and its retrivability. In the

unlikely event of a breach that required the canister to be unloaded,

the canister can be returned to the reactor spent fuel pool. Therefore,

it is incorrect to assert that there is no place to unload a canister.

The Horizontal Storage Module (HSM) is designed to American Concrete

Institute (ACI) 349, which is the required code for nuclear structures

made of reinforced concrete. The transfer cask is designed according to

the ASME BPVC, Section III, Subsection NC; ANSI-N14.6 for heavy loads;

ANSI-50.9 for load combinations; and NUREG/CR 1815 for impact testing.

Because the cask itself is required to meet such exacting standards of

construction, the transport mechanism and the trailer that move the

canister into the HSM are not considered to be important to safety.

Therefore, the design that meets industry standards is sufficient.

B. A number of commenters raised issues relating to releases of

radioactivity from surface contamination and leakage from the casks

under normal and accident conditions.

B.1. Comment. One commenter pointed out that the Certificate of

Compliance Surveillance Requirement 1.2.12 does not have a section

stating the action that is to be taken when the contamination level in

the transfer cask exceeds limits after the DSC has been transferred to

the concrete HSM.

Response. The Certificate of Compliance Surveillance Requirement in

Section 1.2.12 has been modified to clarify that decontamination of the

transfer cask is required if the surface contamination limit is

exceeded.

B.2. Comment. One commenter, who was concerned with the seismic

events at the Davis-Besse Nuclear Power Station, stated that a

displacement pulse of 60 cm, as observed in the Lander's quake in the

Mojave Desert northeast of Los Angeles, would completely destroy the

HSM and allow a substantial release of radioactivity from the fuel

within.

Response. The potential for a seismic event is not the same at

every reactor site in the United States. For Davis-Besse, the maximum

ground displacement has been calculated to be 3.33 inches (8.46 cm),

corresponding to a 0.15g maximum ground acceleration. This is

substantially less than the displacement observed in the Lander's quake

and appears to be well within the design of the Standardized NUHOMS.

Each general licensee using the Standardized NUHOMS, including Davis-

Besse is required to document their evaluations to determine that the

reactor site parameters, including seismic events, envelope the cask

design basis, as specified in its SAR and SER.

B.3. Comment. One commenter, citing a Wisconsin Public Service

Commission draft environmental impact statement (EIS) for Point Beach,

asked for an explanation of why NUHOMS and metal casks have a greater

potential to spread contamination than the Pacific Sierra Nuclear

Associates ventilated storage cask (VSC) system, VSC-24 cask.

Response. The specific rationale that forms the basis of the

statement in the Wisconsin Public Service Commission's draft EIS for

Point Beach was not documented. The decontamination requirements for

the two designs are comparable. The VSC-24 DSC is loaded into the

ventilated concrete cask (VCC) forming the VSC. The VSC is then

transported from inside the reactor auxiliary building to the storage

pad. During moving and storage of the VCC, the exterior surface remains

clean because it has not been exposed to contamination in the spent

fuel pool. The NUHOMS DSC is moved in the transfer cask from the

reactor building to the horizontal storage module in the field. Because

the transfer cask has been in the spent fuel pool, it may have small

amounts of external contamination that have the potential to spread

during transit. However, any potential contamination of this type could

not be significant. The NRC requires that the limits for surface

contamination, workers' dose, and environmental dose must all be met

for the operation of the ISFSI, including during any transfer

operations. Each 10 CFR Part 50 licensee must have a radiation

protection program to monitor operations to ensure that surface

contamination and worker and public exposure to radiation are below

acceptable levels and as low as is reasonably achievable (ALARA). Past

operation of the NUHOMS shows that the doses are well below all NRC

limits.

B.4. Comment. One commenter ``is concerned that heat generated by

fission product decay may provide the driving force, the presence of

free moisture in water-logged fuel may, in a non-mechanistic way,

provide a transport mechanism for fission product release and the

ambient air circulating through the cask concrete structure may provide

(an unmonitored) pathway to the biosphere.'' One commenter remained

concerned about the possibility of insufficient drying of the fuel

before placement in the DSC. Another commenter, citing the Battelle

Pacific Northwest Laboratory Report PNL-5987 on the removal of moisture

from degraded fuel during vacuum drying, contends that the mechanism

for free moisture and radionuclide release that pertain in normal or

upset conditions, such as conditions caused by sabotage, have not been

simulated adequately.

Response. The DSC is a closed vessel. There is no path available

for release of fission products from inside the DSC to the atmosphere.

During normal operation, the circulating air, as it passes through the

HSM and around the outside of the DSC to remove the heat, never comes

in contact with fission products and therefore, could not remove these

products from the cavity of the DSC. Moreover, design basis accidents

under upset conditions were postulated and analyzed in the SAR and SER.

These analyses show that the heat generated from fission product decay

is not capable of breaching the DSC and could not provide the driving

force for a release of radioactivity. Further, it is not expected that

any significant amount of moisture will remain in the fuel after it is

loaded into the DSC. The fuel is dried after it has been loaded into

the DSC and the topcover plate seal welded to the DSC shell. The

Certificate of Compliance requires two pump-downs to a vacuum pressure

of less than 3 mm Hg each with a holding time of greater than 30

minutes. A stable vacuum pressure of less than 3 mm Hg further assures

that all liquid water has evaporated in the DSC cavity.

The safeguards issue of radiological sabotage of storage casks has

been reviewed previously and assessed in the 1989 proposed rule to add

Subparts K and L to 10 CFR Part 72 (54 FR 19379). The NRC has

determined that the Standardized NUHOMS is sufficiently robust such

that the effects of a successful attack would have low health

consequences and are similar to the results presented in the 1989

proposed rule. (see also response to comment N.1)

C. A number of comments were received that focused on monitoring,

surveillance, and inspection activities associated with dry cask

storage of spent fuel, particularly as they relate to the Standardized

NUHOMS.

C.1. Comment. One commenter stated that there are neither active

nor passive systems in place to mitigate barrier breaches, nor are

there active radiation monitors that would indicate a breach has

occurred. There are no monitored drains and sumps nor are there

retention basins. The commenter stated that the cask is insufficient to

be relied upon for the health and safety of Ohioans.

Response. The Certificate of Compliance (Section 1.3) for the

Standardized NUHOMS includes surveillance and monitoring requirements

that are more than sufficient to detect cask degradation in time to

ensure that adequate corrective actions can and will be taken. In

addition, radiation monitoring and environmental monitoring programs

would detect any radiation leak in excess of NRC limits from an NRC-

approved cask.

In some instances, the NRC has required continuous monitoring where

it is needed to determine when corrective action needs to be taken.

Under a general license, to date, the NRC has accepted continuous

pressure monitoring of the inert helium atmosphere as an indicator of

acceptable performance of mechanical closure seals for dry spent fuel

storage casks.

However, the NRC does not consider continuous monitoring for the

Standardized NUHOMS double-weld seals to be necessary because:

(1) There are no known long-term degradation mechanisms which would

cause the seal to fail within the design life of the DSC; and

(2) The possibility of corrosion has been included in the design

(see SER Section 3.2.2.5).

These conditions ensure that the internal helium atmosphere will

remain stable. Therefore, an individual continuous monitoring device

for each HSM is not necessary. However, the NRC considers that other

forms of monitoring, including periodic surveillance, inspection and

survey requirements, and application of preexisting radiological

environmental monitoring programs of 10 CFR Part 50 during the use of

the canisters with seal weld closures can adequately satisfy NRC

requirements.

With respect to the use of instrumentation and control systems to

monitor systems that are important to safety, the user of the

Standardized NUHOMS will, as provided in Chapter 14 of the SER and in

Section 1.3.2 of the Certificate of Compliance, be required to verify,

the cask thermal performance on a daily basis by a temperature

measurement, to identify conditions that threaten to approach cask

design temperature criteria. The cask user will also be required to

conduct a daily visual surveillance of the cask air inlets and outlets

as required by Chapter 12 of the SER and Section 1.3.1 of the

Certificate of Compliance.

While the HSM and DSC are considered components important to

safety, they are not considered operating systems in the same sense as

spent fuel pool cooling water systems or ventilation systems that may

require other instrumentation and control systems to ensure proper

functioning. Due to this passive design, temperature monitoring and

surveillance activities are appropriate and sufficient to assure

adequate protection of the public health and safety for this design.

Because the Standardized NUHOMS DSC is welded closed and has been

decontaminated before being placed in a HSM, there is no routine

radioactive liquid generation that would require a retention basin or

sump. Water entering the storage area has no mechanism of becoming

contaminated because the DSC is enclosed within the HSM and is expected

to be dried by the heat generated during storage.

C.2. Comment. One commenter expressed concern over the possible

external corrosion of the stainless steel DSC because of exposure to

water over decades. Another commenter expressed concern about corrosion

of stainless steel under conditions of indefinite duration, stating

that while stainless steel corrodes less rapidly than carbon steel,

even the plumbing fixture industry is finding unexpected stainless

steel pitting and corrosion under conditions far less intense than

those in a DSC. Another commenter stated that the system is not

designed for remote inspection of the DSC for corrosion while it is in

the HSM and that the only way to inspect the DSC is to return it to the

spent fuel pool. Periodic inspection of the DSC is needed to preclude

or identify gradual canister deterioration by unknown mechanisms.

Another commenter inquired about a checking system for the NUHOMS in

the future. How will corrosion be evaluated on the canister (DSC) and

the support rails inside the HSM? Is it possible for them to accumulate

moisture and corrode together over possibly many years of storage? What

check is required on the possibility that the canister couldn't be

removed at the end of cask life?

Response. The DSC is enclosed within the HSM and is not exposed to

external water. Laboratory experiments have indicated a general

corrosion rate of less than 0.00001 inches per year for similar

stainless steels. The NRC believes these experiments more accurately

bound DSC corrosion than experiences in unrelated industries. For the

50-year design life of the DSC, the expected corrosion would therefore

not result in exceeding a corrosion depth of 0.0005 inches. This will

not affect the DSC performing its intended safety functions. Because of

the low corrosion rates expected for stainless steel, periodic

inspections for deterioration of the DSC are not considered necessary.

Therefore, inspections are not required. The support rails for the DSC

have an extremely hard-alloy steel applied to the sliding surface, are

ground to a smooth finish, and are coated with a dry film lubricant to

prevent corrosion and to reduce the coefficient of friction.

Furthermore, the environment inside the HSM is protected from rain and

it is kept dry by the heat load from the DSC. Therefore, it is highly

unlikely that corrosion between the stainless steel and the hard alloy

steel surface of the support rail will occur to any significant extent.

These conclusions and analyses regarding the very small likelihood of

corrosion indicate that there is reasonable assurance that the DSC can

be removed from the HSM when required.

C.3. Comment. One commenter questioned whether the screens between

the casks, which are essential to cooling, will remain clear of debris

and how they can be cleaned if they become partially clogged. Another

commenter was concerned about how the roof screen was inspected,

stating that it seems likely that insects, animals, and birds will be

attracted to the warm air coming from the outlet vents. Several

commenters remained concerned about vent blockage that can completely

cover and block screening and vents particularly from insects such as

paper wasps, that build huge nests, and swarms of midges that are

common to the Great Lakes. How are the screens attached?

Response. As stated in the Certificate of Compliance, a licensee

using the Standardized NUHOMS must conduct a daily visual surveillance

of the exterior of air inlets and outlets (front wall and roof bird

screen). In addition, the licensee must perform a daily close-up

inspection to ensure that no material accumulates between the modules

to block the air flow. If the surveillance shows blockage of air vents,

the licensee is required to clear the vent blockage by following

procedures developed by each user of the Standardized NUHOMS. If the

screen is damaged, the licensee must replace the screen. The required

daily surveillance and temperature measurements should readily detect

blockage of the vents or screens by insects, animals, or birds in a

timely manner, leading to the removal of the obstruction before damage

occurs from high temperatures. The bird screen is made of stainless

steel wire cloth tack-welded to stainless steel strips, which are

attached to the HSM with stainless steel wedge anchors.

C.4. Comment. One commenter expressed concern about the presence of

burrowing and other nuisance animals that have posed problems at other

waste sites.

Response. Burrowing and other nuisance animals are not expected to

pose problems for the Standardized NUHOMS. Because of the robust system

design, animals will not be able to get to the radioactive material or

cause damage such that water could cause movement of the radioactive

material. Burrowing under the concrete pad would not cause damage to

safety-related components. Further, large-scale burrowing would likely

be detected by the daily surveillance or other activities related to

the operation of the storage area.

C.5. Comment. One commenter wanted additional radiation monitoring

because of the calculated higher dose rates over previous NUHOMS

designs. The commenter stated that these higher dose rates are not

consistent with the objective of maintaining occupational exposures

ALARA, and that site-specific applications should provide detailed

procedures and plans to meet ALARA guidelines and 10 CFR Part 20

requirements with respect to operation and maintenance.

Response. No additional radiation monitoring has been specifically

identified or required for the Standardized NUHOMS. However, 10 CFR

Parts 20, 50, and 72 require that licensees comply with ALARA. In

addition, 10 CFR 72.212(b)(6) requires each licensee to review its

radiation protection program to determine that their effectiveness is

not decreased by use of the Standardized NUHOMS. Further, 10 CFR 72.212

(b)(9) requires each licensee to conduct storage activities in

accordance with appropriate written procedures. If the results of these

licensee activities indicate that additional procedures are required

then the licensee is required to implement the procedures.

C.6. Comment. One commenter was concerned about the optical survey

equipment used to align the transfer cask with the HSM before transfer.

What checks are made on this optical equipment and what regulations

apply?

Response. The optical equipment used to align the transfer cask

with the HSM is optional and is an operational convenience. However,

the licensee must meet Technical Specifications 1.2.9 in the

Certificate of Compliance. Therefore, only appropriate calibrations or

checks to assure compliance with this technical specification are

appropriate.

C.7. Comment. One commenter wants to know who evaluates the

insertion or retrieval of the DSC for excessive vibration and what is

the result of excessive vibration. Would this allow crud to be

released?

Response. The NRC Certificate of Compliance, Section 1.2.9 provides

that the cask user observe the transfer system during DSC insertion or

retrieval to ensure that motion or excessive vibration does not occur.

It also prescribes certain follow-up actions to be taken by the cask

user in the event that alignment tolerances are exceeded and excessive

vibration occurs. It is possible that excessive vibration could

dislodge crud. However, the crud would be contained within the DSC and

would not be released to the atmosphere because the DSC is a sealed

vessel. Any opening of the DSC will be under controlled conditions that

should safely contain the crud and prevent its release to the

environment.

C.8. Comment. Several commenters wanted the NRC to set definite

methods for the required surveillance and monitoring of NUHOMS,

including the daily temperature measurements, so that data are uniform

and standardized for future reference on different modules at different

reactor locations.

Response. The NRC Certificate of Compliance for the Standardized

NUHOMS has required temperature measurements. However, the licensee or

vendor has latitude in determining how the performance-based

temperature requirements will be met. The NRC is not convinced that the

possible benefits of a uniform, but prescriptive, surveillance and

monitoring system or technique would outweigh the costs of curtailing

the freedom of cask users to design an implementation scheme suited to

their individual needs. The collection of uniform data for possible

future use, but without a specific regulatory need could lead to

additional exposure to workers, or adversely affect safety without any

offsetting benefit.

C.9. Comment. One commenter asked about the design life of this

NUHOMS module and on how this is documented. Will the canister be

removed from the concrete module at a specific time and be opened?

Response. The design life of the Standardized NUHOMS is 50 years as

described in the SAR. The Certificate of Compliance has a 20-year

approval period that can be renewed by NRC for another 20 years

following a safety reevaluation. It is expected, that at the end of

operation, the canister will be removed from the concrete module and

will be opened in the spent fuel pool facility or an adequate dry

environment alternative. The fuel will be transferred to an NRC-

approved shipping cask for off-site transportation and ultimate

disposal by the DOE.

C.10. Comment. One commenter believed it prudent to monitor

temperature and air flow to ensure that temperature excursions are not

experienced.

Response. NRC believes the required temperature measurement stated

in Specification 1.3.2 of the Certificate of Compliance, plus the daily

visual inspection of HSM air inlets and outlets, are adequate to ensure

that temperature excursions exceeding the design basis are not

experienced and to determine when corrective action needs to be taken

to maintain safe storage conditions. Therefore, air flow measurements

are not required to assure safety.

D. A number of commenters raised technical issues related to the

thermal analysis of the Standardized NUHOMS and thermal performance of

the system under normal, off-normal, and accident conditions.

D.1. Comment. Several commenters wanted, in the interest of ALARA

principles, the capacity for approximately 24 kW heat removal to be

verified by using an artificial heat load. One commenter suggested that

the NUHOMS be tested with a full heat load at a testing site such as

Idaho National Engineering Laboratory (INEL), and not at each reactor

site that may load it with a higher heat generation rate fuel. Another

commenter cited the ALARA philosophy of loading the oldest fuel first

even though design basis fuel is on site. Several commenters wanted

deletion of the requirement (a literal interpretation of draft

Certificate of Compliance) to calculate the temperature rise for each

HSM loaded with canisters producing less than the design limit of 24 kW

for the following reasons:

(1) Users are not normally provided the vendor's analytical models

for this calculation,

(2) The 100 deg.F rise calculated for the design basis maximum

heat load ensures that all safety limits are met for concrete and fuel,

(3) Because 24 kW is the limit, virtually all the HSMs will be

affected, which places an undue burden on the user to ``baseline'' the

predicted delta-T by calculation considering the inherent safety

margins of the system, and

(4) Technical Specification 1.3.1 ensures that air flow is not

blocked so a false measurement of low temperature rise cannot occur.

Response. A licensee is not required by NRC to load the oldest fuel

first but, in the interest of ALARA, it may do so. However, each time

hotter fuel is loaded up to the maximum allowed in a DSC, the licensee

would need to verify the heat removal performance of the system. For

fuel producing less heat than the design limits of the system, the heat

removal capacity of the system determined by calculation must be

verified by temperature measurements. This process must be repeated

each time a DSC is loaded with hotter fuel until the maximum-system

designed heat load is reached. When loaded with spent fuel producing 24

kW heat, the system may not have an ambient and vent outlet temperature

difference of more than 100 deg.F for fuel cooled equal to or more

than 5 years. This verification process is required to confirm that the

as-built system of each licensee is performing as designed. A licensee

could use an artificial heat source to test an initial cask at a

bounding heat load of 24 kW before loading fuel. However, this test

would only verify the spent fuel heat removal capacity of the system.

It would not verify as-built performance. Experience has shown that

adequate verification testing can be performed at the reactor site.

Therefore, performing the verification at a testing site like INEL

would not provide additional safety margins.

D.2. Comment. Several commenters pointed out possible conflicting

statements about temperature measurements in the surveillance

requirements. In discussions about the heat removal capacity test,

temperatures are determined only during the test period. Daily

temperature measurements on each HSM are required to verify thermal

performance.

Response. These two temperature measurement programs have different

objectives. Temperature measurements by licensees to verify the heat

capacity calculations need only be done until equilibrium is reached.

The daily temperature measurements by licensees are intended to

demonstrate continued safe operation within specified limits over the

life of the HSM and may not be the same type of measurement done in the

initial period to verify heat removal capacity.

D.3. Comment. One commenter was concerned about the adequacy of

cooling under all atmospheric conditions in the country. The commenter

cited conditions such as humidity over 90 percent, temperature over 100

deg.F, and no wind.

Response. Regulatory requirements for general licensee users of dry

storage casks are contained in 10 CFR 72.212(b). Each user must verify

that the following conditions are not exceeded at their reactor site

for the Standardized NUHOMS: the maximum average yearly temperature

with solar incidence is 70 deg.F; the average daily temperature is 100

deg.F; and the maximum temperature is 125 deg.F with incident solar

radiation. If the power reactor site high temperature parameters fall

within these criteria, the Standardized NUHOMS can be safely used at

the site.

D.4. Comment. One commenter wants the NRC to establish procedures

to measure temperature performance, especially the thermal performance

of an individual module and not the combined performance of adjacent

modules as stated on page A-23 of the draft Certificate of Compliance.

Response. As required by the regulations, the licensees are

required to develop detailed procedures. NRC in its regulatory oversite

role has the opportunity to review the adequacy of the procedures. The

requirement cited by the commenter is a requirement for the licensee to

verify a temperature measurement of the thermal performance for each

HSM, not the combined performance of adjacent modules. A cautionary

statement is included in the basis of the specification to ensure that

licensee measurements of air temperatures reflect only the thermal

performance of an individual module and not the combined performance of

adjacent modules.

D.5. Comment. One commenter wanted to know how the temperature

differences in the roof, side wall, and floor areas are incorporated

into the daily temperature measurement.

Response. For the first HSM to be emplaced, the user is required to

measure the air inlet and air outlet temperature difference of the

system at equilibrium. This measurement is to ensure that the heat

capacity of the system will not be exceeded and that the concrete

temperature criteria will not be exceeded. For the Standardized NUHOMS,

this maximum heat capacity is 24 kW. The 24 kW heat load is the design

maximum and is the basis for the thermal hydraulic calculations for the

cask. The temperature distribution for various parts of the HSM have

been calculated (i.e., the roof, walls, and floor) by the cask vendor.

Temperature differences causing thermal stresses in the concrete were

evaluated and are duly reported in both the SAR and SER. These

calculations were reviewed by NRC as a part of the overall process for

this design approval.

D.6. Comment. One commenter stated that daily temperature

measurements are not necessary to ensure convective air flow, given the

requirement to verify that the inlets and outlets are not obstructed.

Site-specific NUHOMS require temperature measurements when the DSC is

placed into the HSM, 24 hours later, and again at 1 week after loading

to ensure adequate thermal performance.

Response. The NRC disagrees with this comment. The HSM and DSC are

considered components important to safety in the Standardized NUHOMS.

Daily temperature measurements of the thermal performance by the

licensee are required to provide additional assurance that thermal

limits are not exceeded under the general license. This requirement was

imposed on the first cask of this type approved by the NRC and listed

in 10 CFR 72.214 for use by a general licensee, the VSC-24 cask (58 FR

17967; April 7, 1994) and is now applied to the Standardized NUHOMS.

E. A number of commenters expressed concern about emergency

planning and response contingencies.

E.1. Comment. Several commenters expressed concern that in the

event of problems and the need to off-load fuel (as in the recent

situation at Palisades), a transfer cask may not be available in a

timely manner because of inclement weather or because the transfer cask

itself has experienced problems or is being used elsewhere. One

commenter expressed concern at having to have a transfer cask on site

within 40 hours of vent blockage to prevent concrete damage. If the

transfer cask is leased from VECTRA and is not at the licensee's site,

who is liable if something happens that would require the use of a

transfer cask?

Response. The NRC has analyzed all design basis accidents from the

operation of an ISFSI and concluded that there will be no release of

radioactive material to the environment. The 40-hour limit on vent

blockage is intended to prevent concrete degradation that might occur

over a long period of storage. A vent blockage accident would not

result in the release of radioactive material because the DSC would not

be breached. Therefore, the NRC believes that the potential risk to the

public health and safety is extremely small during the time needed to

obtain the use of a transfer cask. Thus, there is no requirement that a

transfer cask be at an ISFSI site all the time.

E.2. Comment. One commenter expressed concern that the effects of

tornado winds and missiles during movement of the fuel in a transfer

cask or in a storage cask on a transporter were not analyzed.

Response. Both the vendor's SAR and NRC staff's SER address the

effects of tornado winds and missiles during movement of the transfer

cask with a loaded canister. These analyses show that, for tornado

winds, there is a safety factor of 1.5 against overturning when

subjected to Design Basis Tornado winds (a safety factor greater than 1

will generally be adequate for public protection). The transfer cask

stability, tornado missile penetration resistance, and shell and end

plate stresses were calculated and shown to be below the allowable

stresses for ASME BPVC Service Level D (accident) stresses.

E.3. Comment. One commenter described an October 1972 storm that

flooded the entire Davis-Besse plant site, including the (pre-

operational) reactor building. There has been subsequent flooding of

the site, particularly during spring thaws.

Response. Safety analyses by NRC and the cask vendor show the

Standardized NUHOMS can withstand floods and will continue to perform

acceptably. With regard to the Davis-Besse site, the licensee changed

site topography during plant construction. Specifically, the area was

built up and some dikes were added. The plant structure's ground floor

elevation is 585 feet International Great Lakes Datum (IGLD), which is

also the elevation of the pad. The licensing design basis for maximum

probable static water level on the site is 583.7 feet IGLD. As noted,

the HSM and DSC were evaluated for flood conditions as required by 10

CFR 72.122(b). The HSM can withstand a maximum water velocity of 15

feet per second and a static head of 50 feet of water. The DSC can

withstand a static head of 50 feet of water. Any site that intends to

use a Standardized NUHOMS design must evaluate the conditions at their

site to verify compatibility with the design specifications of the

system.

F. A number of commenters raised issues relating to the design,

evaluation, and operation of the Standardized NUHOMS.

F.1 Comment. Several comments related to the fuel to be stored in

the Standardized NUHOMS. One commenter wanted control components

contained in assemblies addressed in the SAR and SER citing DOE

acceptance criteria. One commenter questioned how 55,000 MWD/MTU burnup

fuel now being used in pressurized water reactors will be handled since

the Standardized NUHOMS-24 is rated to handle only 40,000 MWD/MTU

burnup fuel. Another commenter, citing provisions of current site-

specific licensees for other NUHOMS designs, stated that higher burnup

should be allowed if the decay heat and radiological source terms are

within limits. Another commenter asserted that increased fission

products from higher enriched fuel may potentially increase

embrittlement of the fuel cladding and that this needs to be evaluated

in the SER. This commenter further alleged that this would increase the

probability of more defective fuel being loaded into dry casks.

Response. The vendor designed the cask system for storage of

pressurized water or boiling water reactor fuel assemblies meeting

certain specifications. By limiting the use of the cask system to

assemblies meeting these specifications, the vendor made a decision

that may partially restrict the use of the cask. However, the NRC does

not require that a cask be universal for all types of fuel or be usable

at every reactor site. For example, none of the casks previously listed

in 10 CFR 72.214 is usable for boiling water reactor spent fuel.

Currently, the 55,000 MWD/MTU burnup fuel and fuel with initial

enrichments of greater than 4% will have to remain in the spent fuel

pool because dry spent fuel cask designs to store fuel with this higher

burnup and initial enrichment or related to DOE acceptance criteria

have not yet been reviewed and evaluated by the NRC.

F.2 Comment. Several comments were related to criticality safety

analysis. One commenter questioned the conservatism of using 7.5-year

cooled spent fuel when 5-year-cooled fuel is the minimum specified and

when older fuel may also be stored in the cask. Another inquired about

criticality safety if the original basket geometry were compromised, as

might be the case for brittle failure of a spacer disk. In the

compromised basket geometry case, the commenter also asked about the

difference in criticality safety for a helium atmosphere rather than a

borated water medium. The commenter, referring to July 24, 1992,

meeting minutes, inquired why all parties agreed not to spend any

resources to make these criticality safety calculations.

Response. The Standardized NUHOMS nuclear criticality safety

analysis is based on the following: (1) Babcock and Wilcox 15 x 15/

208 pin fuel assemblies with initial enrichments up to 4.0 wt% of U-235

and (2) General Electric 7 x 7 fuel assemblies with initial

enrichments up to 4.0 wt% of U-235, for the Standardized NUHOMS-24P and

NUHOMS-52B designs respectively. The age of the fuel that will actually

be stored is not relevant in criticality safety analysis because the

analysis assumes storage of unirradiated fresh fuel that is more

reactive than cooled spent fuel. The Standardized NUHOMS-24P system has

administrative controls that limit the irradiated fuel reactivity to

less than or equal to 1.45 wt% of U-235 equivalent unirradiated fuel

(Certificate of Compliance Section 1.2.1).

The possibility of a criticality accident caused by the brittle

failure of the basket should not be a significant concern. No lifting

or handling of the DSC outside the spent fuel pool building is

permitted if the basket temperature is lower than 0 deg.F. If the user

does not determine the actual basket temperature, the ambient

temperature must be used conservatively. Under these temperature

restrictions, the basket materials will not behave in a brittle

fashion. Consequently, the basket geometry would not be compromised by

brittle failure. As for the criticality safety consideration related to

a helium atmosphere versus a borated water medium, the keff of the

fuel in a helium atmosphere is much less than the keff in borated

water. Therefore, criticality calculations for the borated water are

sufficient because they are more conservative and therefore would bound

calculations using a helium atmosphere.

F.3 Comment. Two commenters were concerned with shielding and dose

assessments for the Standardized NUHOMS. One commenter believed that

using 10-year-cooled fuel for the dose assessment was nonconservative

when 5-year-cooled fuel is needed to load the DSC to produce 24 kW of

heat. Another, referring to an NRC meeting with Pacific Nuclear Fuel

Services, Inc. (PNFSI), wanted clarification of an NRC request to

delete a clause allowing the utility to perform site-specific shielding

calculations.

Response. The cask vendor presented dose assessment results in the

SAR for both 5- and 10-year-cooled fuel. However, for this rulemaking,

NRC used the dose assessment for 5-year-cooled fuel for the shielding

analysis radiation source term and for accidental releases of

radionuclide material. NRC's use of the 5-year-cooled fuel assessment

is conservative and bounding.

To ensure safe storage of spent nuclear fuel in NRC-approved casks,

the NRC specifies, in Section 1.2.1 of the Certificate of Compliance a

number of fuel acceptance parameters. These parameters, which may

include burnup, initial enrichment, heat load, cooling time, and

radiological source term, define the properties of those assemblies

that can be stored in a cask. One such parameter of interest for the

Standardized NUHOMS is the radiological source term that forms the

basis of the shielding analyses. For this parameter, the vendor

proposed an alternative approach. Specifically, for fuel assemblies

that fall outside the specified source term parameters but satisfy all

other parameters, the vendor proposed to allow licensees to do

individual cask shielding calculations to show compliance with the

design basis dose rates. This could result in more assemblies in a

licensee's inventory that would be eligible for dry storage. In the

instance noted in the comment, the NRC did not agree with the vendor

proposal. The Certificate of Compliance dose rate specifications

provide a simple check to ensure that DSCs are not inadvertently loaded

with the wrong fuel. The dose rate specifications are based on the

shielding analyses provided by the vendor in its SAR. Because of

differences in non-fuel components in the ends of some assemblies, dose

rates higher than those evaluated by NRC in the SER may occur at the

ends of casks than were assumed in the shielding analysis. The

Certificate of Compliance specifications allow for this possibility and

permit the licensee to store such fuel provided the licensee verifies

proper cask fabrication, conformance with all other fuel parameters,

and compliance with radiation protection requirements. The site-

specific calculations referred to in the comment are not shielding

calculations, but rather are the licensee's written evaluations (or

dose assessments) required by 10 CFR 72.212(b)(2)(iii) to establish

that the radiation criteria for ISFSI in 10 CFR 72.104 have been met.

The Certificate of Compliance also requires that the licensee submit a

letter report to the NRC summarizing its actions in this type of case.

F.4. Comment. Several commenters were concerned with fuel clad

integrity issues. Particularly, they were concerned with potential

problems that may arise because of differences between vertical and

horizontal storage. One commenter noted that it was essential to

inspect the cladding carefully for the minute hairline cracks which

would allow the radioactivity inside to escape. Another commenter

wanted it made clear that for fuel to be eligible for storage it

doesn't need to be specifically inspected nor require special handling

or storage provisions within the spent fuel pool. The commenter also

asserted that pinhole leaks in fuel rod cladding do not constitute

gross breaches. The commenter wanted fuel cladding integrity clarified.

Another commenter claimed that horizontal storage of fuel rods will

lead to cladding deterioration that would challenge the technical

specifications of the NUHOMS cask. Another commenter was concerned

about the possibility of fuel rod bowing that could result in weighted

contact between the fuel cladding/crud and the DSC guide sleeve, with

the potential for eventual bonding of the materials over the duration

of the storage period. One commenter, noting that some of the fuel in

the spent fuel pools could be nearly 20 years old, was concerned that

the fuel will not be tested for leaks using specific techniques such as

penetrating dyes, eddy current, sipping, or ultrasound before canister

loading. A commenter wanted all fuel with known defects and all water-

logged fuel retained in the spent fuel pool until the cask integrity

under operating conditions is fully demonstrated. Another wanted to

know how ``grossly breached'' fuel will be ultimately handled and

shipped off site.

Response. In the Standardized NUHOMS, PWR fuel rods are stored in a

horizontal orientation and do not normally deflect in the middle of any

span so that the rods contact the DSC guide sleeve. However, the

possibility exists that a bowed rod may come in contact with the guide

sleeve.

With respect to storage of BWR fuel, the fuel channel that

surrounds the fuel bundle (rods) provides a barrier to separate coolant

flow paths, to guide the control rod, and to provide rigidity and

protection for the fuel bundle during handling. Therefore, the BWR fuel

rods inside the channel do not come in contact with the guide sleeves.

Even if there were contact with either PWR or BWR fuel rods, the

interaction would not present a significant concern because the guide

sleeve material is stainless steel, which has a very low rate of

corrosion, and the DSC cavity is evacuated and back-filled with inert

helium, which further reduces the likelihood of any corrosion or

bonding involving the guide sleeve and fuel rods.

The Certificate of Compliance requires that the fuel have no known

or suspected gross cladding breaches to ensure the structural integrity

of the fuel. Known or suspected failed fuel assemblies (rods) and fuel

with cladding defects greater than pin holes and hairline cracks are

not authorized in the Standardized NUHOMS. Fuel meeting this

specification will be safely stored and will remain intact in storage

because the dry inert atmosphere and relatively low temperature will

prevent deterioration of the cladding. Grossly breached fuel will be

handled in site-specific license applications.

F.5. Comment. Quite a few comments related to the structural

stability of the HSM, particularly its response to earthquakes.

Commenters questioned the possibility of vertical storage of the

Standardized NUHOMS and suggested that it would be very difficult to

restrain the HSM if the DSC were in a vertical position. One commenter

wanted dry storage casks constructed to Building Officials Code

Administrators (BOCA) National Building Code (and Ohio Administrative

Code) for structures in use group H-4, high hazard use, which includes

radioactive materials. Commenters questioned whether ground

acceleration as used by the NRC in its evaluation could adequately

describe all potential earthquakes east of the Rocky Mountain Front and

suggested that a ground acceleration of 2.5g would not be realistic for

all sites, despite proximity to fault lines. Another commenter alleged

a number of seismic events in the midwest which had some effect in the

Ohio area could cause a complete failure of the cask and requested that

the NRC insist that the cask, containment structure, and foundation pad

be designed to substantially exceed all earthquakes with a potential

for 0.60g. One commenter wanted to know if the module had been analyzed

for earthquake events at all United States reactor sites, according to

Laurand Findmun Seismic Hazard Curves. Other commenters expressed

various concerns about the integrity and reaction of the Standardized

NUHOMS components under earthquake conditions and asked the following

questions:

Could the casks crash against each other as the ground moves

beneath them?

Could the module shift, crack, or move off the pad?

How are the rail support holdings evaluated?

Could the DSC be knocked off the rails? and

Could the module roof crack and fall on the canister?

Response. The Standardized NUHOMS design described in the vendor's

applications for approval and the SAR does not address vertical

storage. Consequently, NRC neither evaluated nor approved vertical

storage for the system. Therefore, it may not be stored vertically.

The NRC reviewed the Standardized NUHOMS for compliance with design

criteria that are more stringent than those of the BOCA National

Building Code (NBC) (see response to Comment A.5). These more stringent

criteria are included in national standards that more closely represent

the use of the Standardized NUHOMS.

Part 72 specifies a design basis maximum ground acceleration of

0.25g for areas east of the Rocky Mountain Front that are not in areas

of known seismic activity. All HSMs and DSCs are designed to withstand

a 0.25g earthquake. Any reactor licensee who intends to use the

Standardized NUHOMS must verify that the maximum displacements at the

cask's location on the reactor site are within the design criteria for

the system. The Standardized NUHOMS is free standing and not dependent

on the pad for safety. Failure of the pad caused by seismic events will

not cause the Standardized NUHOMS to fail. Therefore, cask safety does

not require the pad to be designed to withstand a seismic event.

F.6. Comment. One commenter stated that the SAR did not include

consideration of the accident events such as: aircraft crashes, turbine

missiles, external fires, explosions, and sabotage.

Response. Before using the Standardized NUHOMS, the general

licensee must evaluate them to ensure the site is encompassed by the

design bases of the approved cask. The events listed in the comment are

among the site-specific considerations that must be evaluated.

The site evaluation for a nuclear plant considers the effects of

nearby transportation and military activities. It is incumbent upon the

user of the cask to determine if the SER for the facility encompasses

the design basis analysis performed for the Standardized NUHOMS or any

certified cask. The great majority of the aircraft are single-engine

propeller airplanes which typically weigh on the order of 1,500 to

2,000 pounds. The cask's inherent design will withstand tornado

missiles and other design loads and also provides protection from the

collision forces imposed by these light general aviation aircraft

without adverse consequences. NUREG-800, Section 3.5.1.6 ``Standard

Review Plan for Light Water Reactors,'' contains methods and acceptance

criteria for determining if the probability of an accident involving

larger aircraft (both Military and civilian) exceeds the acceptable

criterion. It is incumbent upon the licensee to determine whether or

not the reactor site parameters are enveloped by the cask design basis

as required by 10 CFR 72.212(b)(3). These would include an evaluation

demonstrating that the requirements of 10 CFR 72.106 have been met.

Turbine missile analyses typically show a very low probability of a

turbine missile breaking the turbine casing. The site's turbine missile

analyses must be considered as part of the facility's analysis of the

suitability of the storage location. External fires are handled by

established fire control programs. Explosions are prevented by control

of combustibles under the licensee's fire protection program. Sabotage

is considered under the criteria for security programs that each

licensee must implement. (See also response to comment N.1).

F.7. Comment. Several commenters raised issues about the pad and

foundation for the Standardized NUHOMS. One commenter referred to a

previous rulemaking that stated that the NUHOMS casks required site-

specific approvals because they are constructed in place. Other

commenters, concerned with seismic events at the Davis-Besse Nuclear

Power Station and soil stability issues similar to cask use at the

Palisades Plant, asserted that there was a necessary relationship of

the Standardized NUHOMS cask or module to the pad at a specific site

and that evaluation of it could not be based on the reactor site

seismic analysis. Each site required singular seismic and soil analysis

for dynamic loads and not just static loads.

Response. The NUHOMS design referred to in the July 18, 1990, 55 FR

29181, rulemaking includes the site-specific pad as an integral part of

the concrete HSM and therefore it is important to safety. The

Standardized NUHOMS considered in this rulemaking have the HSMs as

free-standing units; that is, they have no structural connections to

the pad. The Standardized NUHOMS does not rely on the pad to perform a

safety function to protect public health and safety. The vendor

analyzed the HSM containing the DSC for peak ground accelerations of

0.25g caused by earthquakes and found that it would neither slide nor

overturn. NRC evaluated the Standardized NUHOMS under a wide range of

site conditions that could diminish cask safety. Further, under the NRC

general license, before using the Standardized NUHOMS a licensee must

verify that reactor site parameters are within the envelope of

conditions reviewed by NRC for the cask approval. If potential

conditions exist at the reactor site (including potential erosion, soil

instability, or earthquakes) that could unacceptably diminish cask

safety by any credible means, the licensee's analysis must include an

evaluation of the potential conditions to verify that impairment of

cask safety is highly unlikely.

The NRC's regulations do not explicitly require a licensee using a

cask under a general license to evaluate the cask storage pad and

foundation under such site conditions for erosion or earthquakes. If

conditions at the reactor site could unacceptably diminish cask safety

by affecting the stability of the supporting foundation so as to put

the cask in an unsafe condition, the cask may not be used unless the

foundation is appropriately modified or a suitable location at the

reactor site is found. Implicitly, therefore, the pad and the

underlying foundation materials must be analyzed under site conditions

that include erosion, soil instability, and earthquakes, even though

the pad has no direct safety function and the cask is designed to

retain its integrity even assuming the occurrence of a range of site

conditions.

The licensee has the responsibility under the general license to

evaluate the match between reactor site parameters and the range of

site conditions (i.e., the envelope) reviewed by NRC for an approved

cask. Typically, the licensee will have a substantial amount of

information already assembled in the Final Safety Analysis Report

(FSAR) for the nuclear reactor. In addition, the envelope for the

approved cask is identified in the NRC SER and Certificate of

Compliance and in the cask vendor's SAR for the cask. Of course, the

licensee should consider whether the envelope evaluated by NRC

adequately encompasses the actual location of the cask at the reactor

site. The licensee should also consider whether there are any site

conditions associated with the actual cask location that could affect

cask design and that were not evaluated in the NRC safety evaluation

for the cask.

The vendor analyzed the DSC and the HSM for rigid body response

(i.e., sliding and overturning) to seismic accelerations. The resultant

peak horizontal ground acceleration is 0.37g and the peak vertical

acceleration is 0.17g. The margin of safety against sliding is 1.35.

Similarly, the design seismic force will not cause the HSM to tip over

because the stabilizing moment of the HSM is greater than the seismic

overturning moment. The margin of safety against overturning is 1.26.

Thus, no sliding or overturning of the HSM or DSC will occur from the

design earthquake.

Because the pad is not considered a safety-related item, a specific

pad design is not being approved in this rulemaking for the

Standardized NUHOMS.

F.8. Comment. A few commenters had questions pertaining to the

operation of and procedures for the Standardized NUHOMS. One commenter

inquired whether just one module of the Standardized NUHOMS could be

purchased by a utility, or whatever number of modules desired could be

procured and easily added like singular casks. One commenter expressed

concern about snow removal procedures to prevent blockage of the bottom

vents by drifting snow. Another commenter wanted NRC to establish a

procedure and criteria for dose rates discussed on pages A-15 and A-16

in the draft Certificate of Compliance. Several commenters noted that a

procedure for opening a storage cask and removing the fuel has not been

tried before nor documented in the rulemaking. They were also concerned

that unloading of a cask would place workers at higher risk.

Response. The NRC Certificate of Compliance does not permit or

limit the number of NUHOMS modules that may be purchased by a general

license. The NRC does not regulate the commercial arrangements between

the cask vendor and the users including any provisions on the number of

casks that can be purchased or added to the Standardized NUHOMS.

Under the Certificate of Compliance, Section 1.3, the user of the

Standardized NUHOMS (general licensee) is required to conduct a visual

surveillance of the exterior of air inlets and outlets. If the

surveillance shows blockage of air vents, they must be cleaned in

accordance with proper procedures. These procedures will minimize the

potential impact to the health and safety of workers. The daily

temperature measurements indicate proper thermal performance.

The Certificate of Compliance requires each licensee to develop

procedures to implement the dose criteria prescribed on pages A-15 and

A-16. On page A-15 of the Certificate of Compliance, Section 1.26, the

dose rate criteria to be met is equal to or less than: (a) 200 mrem/hr.

at the top shield plug surface at centerline with water in the cavity;

and (b) 400 mrem/hr. at the top cover plate surface at centerline

without water in the cavity. On page A-16 of the Certificate of

Compliance the dose rate criteria is less than or equal to: (a) 400

mrem/hr. at 3 feet from the HSM surface; (b) 100 mrem/hr. outside of

the HSM door on center line of the DSC; and (c) 20 mrem/hr. at the end

shield wall exterior. Each licensee is required to develop its own

procedures to implement these criteria. In addition, each licensee must

develop operational procedures for the ISFSI for workers' radiation

exposure to be ALARA.

For the Standardized NUHOMS, removal of spent fuel from the DSC is

addressed in Chapter 5 of the SAR and in Chapter 11 of the SER. The

process is essentially the reverse of loading operations and would be

performed under the reactor license radiation protection program. The

Certificate of Compliance requires each user to develop written

procedures for these operations and includes precautions to be

considered for unloading. ALARA is required to be addressed by 10 CFR

Part 20. Specification 1.1.6 of the Certificate of Compliance requires

that pre-operational testing and training exercises include the opening

of a DSC and returning the DSC and transfer cask to the spent fuel

pool. The Certificate of Compliance also requires the training program

to include off-normal events.

F.9. Comment. One commenter, citing the May 1993 study prepared for

the NRC by the Center for Nuclear Waste Regulatory Analyses of San

Antonio, Texas, questioned the relatively higher temperature

consequences of dry storage on fuel cladding. The report states that,

``the dry environment has the potential of producing such problems as

further fuel cladding oxidation, increased cladding stresses, and creep

deformation as a result of rod internal pressure * * *. These possible

spent fuel and cladding alteration modes could be quite accelerated

under dry storage conditions, since temperatures are much higher than

in wet storage.'' The commenter does not believe that NRC is fulfilling

its obligation in 10 CFR 72.122(h) to see that ``spent fuel cladding

must be protected during storage against degradation that leads to

gross rupture.''

Response. The May 1993 study addresses the long-term geological

disposal of high-level waste (spent fuel) and is not directed to the

short-term interim storage of spent fuel at nuclear power plants. The

report evaluates processes over 10,000 years of repository performance

for geological disposal. The conclusions of the report are not

applicable for the interim storage period of a 20-year cask certificate

during which spent fuels stored in the DSC have to meet the NRC's

criteria to ensure that cladding is protected. Under normal operation

of the ISFSI, leakage of radionuclides is not expected to occur. The

design and the double-seal welding of the DSC covers are checked and

tested to provide structural integrity throughout the approved storage

period. During normal storage conditions, the licensee is required to

conduct a radiation monitoring program to ensure protection of workers

and the safety of the general public.

G. A number of comments were related to broad policy and program

issues in connection with the storage and disposal of high-level

radioactive waste, including the DOE repository program. Some

commenters questioned the use of dry cask storage and technology in

general. Some commenters stated that only dry storage casks that would

be compatible with DOE interim or final repository operations,

including transportation, should be approved for use under a general

license.

G.1. Comment. One commenter does not want any more casks approved

until a permanent Federal repository is opened. The wet fuel pool is a

proven technology that has been successful in containing radioactivity.

Another commenter stated that dry storage is dangerous.

Response. The NRC, in implementing the Nuclear Waste Policy Act of

1982, has an obligation to review dry storage technologies and to

determine whether to approve the use of these technologies for the

storage of spent fuel if they meet applicable safety requirements. The

July 18, 1990, 55 FR 29181, rulemaking found that spent fuel stored in

dry storage casks designed to meet the NRC regulatory requirements can

safely contain radioactivity. This rulemaking adds one cask design that

meets the safety requirements previously developed to the list of

approved casks. The previous responses to comments, as well as the

detailed safety and environmental analyses underlying this rulemaking

(and described elsewhere in this notice), all reveal that the

Standardized NUHOMS will conform to the NRC requirements and that its

use should not pose the potential for significant environmental

impacts.

DOE is required by the Nuclear Waste Policy Act of 1982 to accept

spent fuel for ultimate disposal. Moreover, the Commission made a

generic determination in its waste Compliance Decisions (September 18,

1990, 55 FR 38474 and August 31, 1994, 49 FR 34658) that safe disposal

is technically feasible and will be available within the first quarter

of the 21st century.

Dry cask storage has significant advantages over wet storage in

that the system is passive and requires minimal human intervention. No

pumps, filters, or water quality monitoring are needed to maintain the

conditions necessary for wet storage. The only monitoring required for

the Standardized NUHOMS is daily temperature monitoring and visually

checking inlet and outlet vents.

G.2. Comment. A number of commenters wanted a full formal trial-

type public hearing on the use of the NUHOMS cask.

Response. Consistent with the applicable procedure, the NRC does

not intend to hold formal trial-type public hearings on the

Standardized NUHOMS rule or separate hearings at each reactor site

before the use of the dry cask technology approved by the Commission in

this rulemaking. Rulemaking procedures, used by the NRC for generic

approval of the Standardized NUHOMS, including the underlying NRC staff

technical reviews and the opportunity for public input, are more than

adequate to obtain public input and assure protection of the public

health and safety and the environment. In this rulemaking, the NRC has

taken additional steps to elicit and fully consider public comments on

the Standardized NUHOMS technology. These steps included NRC

participation in public meetings near Davis-Besse and extension of the

public comment period by 45 days in response to public requests. This

extension provided a total public comment period of almost 4 months.

Section 133 of the Nuclear Waste Policy Act of 1982 authorizes the

NRC to approve spent fuel storage technologies by rulemaking. When it

adopted the generic process in 1990 for the review and approval of dry

cask storage technologies, the Commission stated that ``casks * * *

[are to] be approved by rulemaking and any safety issues that are

connected with the casks are properly addressed in that rulemaking

rather than in a hearing procedure'' (July 18, 1990; 55 FR 29181).

Rulemaking under NRC rules of practice, described in 10 CFR 2.804 and

2.805, provides full opportunity for expression of public views but

does not use formal trial-type hearings of the kind requested by

commenters.

In this proceeding, rulemaking clearly provided adequate avenues

for members of the public to provide their views regarding NRC's

proposed approval of the Standardized NUHOMS, including the opportunity

to participate through the submission of statements, information, data,

opinions and arguments. In this connection, technical evaluations for

Standardized NUHOMS and detailed documented findings of compliance with

NRC safety, security, and environmental requirements were prepared by

the NRC staff for public examination. In November 1993, the NRC staff

reviewed the Standardized NUHOMS and approved the design for the

purpose of initiating this rulemaking to grant a generic approval of

the design. In addition, the NRC staff conducted a second review in

response to the public comments on the Standardized NUHOMS in this

rulemaking, again finding compliance with NRC requirements as discussed

in this document.

In addition to reviewing systematically and in depth the technical

issues important to protecting public health and safety, and the

environment, the NRC has taken extra steps to obtain and fully consider

public views on the Standardized NUHOMS technology and has made every

effort to respond to public concerns and questions about the

Standardized NUHOMS compliance with NRC safety, security, and

environmental requirements. The initial public comment period opened on

June 2, 1994, and was scheduled to close on August 16, 1994. On August

29, 1994, the public comment period was extended to September 30, 1994.

The NRC also participated in an earlier meeting near the Davis-Besse

site.

Under these circumstances, formal hearings would not appreciably

add to NRC's efforts to ensure adequate protection of public health,

safety, and the environment and they are unnecessary to NRC's full

understanding and consideration of public views on the Standardized

NUHOMS.

G.3. Comment. One commenter stated that because there is not now

and there may not be a permanent high-level radioactive waste (HLWR)

repository for commercial reactor fuel, and since the NUHOMS 24P and

52B casks are non- transportable, any distinction between so called

``temporary storage'' and ``permanent disposal'' of this waste is moot.

Because of the lack of a permanent repository or Monitored Retrievable

Storage (MRS) in the foreseeable future, a case of a serious spill and

the resultant contamination at an environmentally unsuitable site like

Davis-Besse where ``short and long-term adverse impacts associated with

the occupancy and modification of (a) floodplain * * * potential

release of radioactive material during the lifetime of the ISFSI * * *

(and location) over an aquifer which is a major water resource'' have

been inadequately dealt with.

Response. This rulemaking to certify the Standardized NUHOMS is for

interim storage of spent fuel in an approved cask for 20 years. It does

not authorize or approve the ultimate disposal in a permanent HLRW

repository, which is under the responsibility of the DOE. During

interim storage, the user (holder of a Part 50 license) must protect

the spent fuel against design basis threats, and against environmental

conditions and natural phenomena such as tornadoes, tornado missiles,

earthquakes, and floods. In regard to flooding, the Certificate of

Compliance has a provision (see A-2 of Certificate of Compliance) for

flood condition analysis to ensure that there is no release of

radioactive material from flooding.

G.4. Comment. One commenter stated that projected future uses of

land and water within the region are impossible to make given the

unknown length of time this waste may remain on site and the options

for both cask and reactor license renewal beyond 20 and 40 years,

respectively, and the fact that no known man-made structure can last

for the length of time that this waste must be isolated from humans and

the environment. If an MRS or repository ever become available, this

waste may have to be repacked. Each handling of this waste increases

the likelihood of an accident, spill, contamination, and worker and

public exposures.

Response. Projected future land and water use can be made based on

the continued safe operation of a reactor and its associated dry cask

storage facility. The continued operation of these facilities should

have no greater impact on land and water use in the future than they do

today. As previously noted, the NRC Waste Confidence decisions

concluded there is reasonable assurance that safe disposal of spent

fuel by the Federal Government will be available by the year 2025.

Therefore, the spent fuel will not remain at a reactor site for the

length of time it must be isolated from humans and the environment.

It should be noted that the absence of significant environmental

impacts from dry cask storage at a reactor site is the conclusion of

NRC's environmental assessment for the Standardized NUHOMS and for

previously approved dry casks analyzed in earlier rulemakings

addressing 10 CFR Part 72, as well as in the Commission's Waste

Confidence decisions in 1984 (August 31, 1984; 49 FR 34658) and 1989

(September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence

decision, the Commission concluded there was reasonable assurance that

spent fuel can be safely stored at reactor sites, without significant

environmental impacts, for at least 30 years beyond expiration of NRC

reactor operating licenses. The 1989 Waste Confidence decision review

reaffirmed earlier Commission conclusions on the absence of significant

environmental impacts.

G.5. Comment. One commenter questioned whether the NUHOMS canister

will fit the conceptual design for the DOE multi-purpose canister

(MPC). If DOE chooses to use vertical casks (like the VSC) at the MRS,

will the NUHOMS inner canister fit into the vertical outer concrete

shell in the MPC design? If local reactors choose the VSC-24 or the

NUHOMS, will either inner metal canister fit into the overpacks for

DOE, or will they have to be opened after storage, returned to the

pool, the fuel put in a new canister, and the old one discarded as

radioactive waste?

Response. The Certificate of Compliance for the Standardized NUHOMS

is intended for the interim storage of spent fuels and is not required

to conform to, and has not been evaluated by NRC for conformance with,

the conceptual design for the DOE MPC. DOE has not yet made final

decisions regarding design or deployment of the MPC. Therefore, it is

not possible to speculate on conformance of the Standardized NUHOMS to

the MPC.

G.6. Comment. One commenter asked what are the criteria for 20-year

renewal of this cask design? How will this be checked? If the design is

not renewed, what is the plan?

Response. The 1989 proposed rule (May 5, 1989; 54 FR 19379) to add

Subparts K and L to Part 72 indicated that the 20-year period

represents what the Commission believes to be an appropriate increment

for cask design approvals. The application for design reapproval would

have to demonstrate the cask's ability to perform the necessary safety

functions for the reapproval period. The application would be evaluated

by NRC against the Commission's regulatory requirements. If a cask

design is not reapproved, the licensee would have to remove casks from

service as the 20-year approved storage life expired. This could mean

removal of the spent fuel and storing it elsewhere.

G.7. Comment. One commenter wanted to discuss the need for an

additional cask design, including how it would better meet the need of

the interim dry cask storage of high-level waste.

Response. Section 218(a) of the Nuclear Waste Policy Act of 1982

(NWPA) provides the following directive: ``The Secretary [of DOE] shall

establish a demonstration program in cooperation with the private

sector, for the dry storage of spent nuclear fuel at civilian nuclear

reactor power sites, with the objective of establishing one or more

technologies that the [Nuclear Regulatory] Commission may, by rule,

approve for use at the sites of civilian nuclear power reactors

without, to the maximum extent practicable, the need for additional

site-specific approvals by the Commission.'' After subsequent DOE

technical evaluations and based on a full review of all available data,

the Commission approved dry storage of spent nuclear fuel in a final

rule published in the Federal Register on July 18, 1990 (55 FR 29181).

The final rule established a new Subpart K within 10 CFR Part 72,

entitled ``General License for Storage of Spent Fuel at Power Reactor

Sites.'' Therefore, there is a need for casks to be approved by NRC to

implement the NWPA to meet the demand of the interim dry cask storage

of spent fuels in the nuclear power plants. However, the variety of

cask designs submitted by vendors for NRC review and approval is mostly

dictated by economic reasons that do not involve NRC.

H. A number of commenters wanted site-specific analyses done for

each use of the Standardized NUHOMS despite the fact that each licensee

must determine that the site parameters are enveloped by the cask

design specified in the SAR, SER, and Certificate of Compliance. The

intent of Subpart K of 10 CFR Part 72 was to grant a general license to

licensees of power reactors to use NRC-approved dry storage casks

listed in 10 CFR 72.214 without additional licensing review by NRC.

H.1. Comment. A number of commenters wanted site-specific

Environmental Impact Statements (EIS). Several commenters stated that

an EIS should be required on any waste facility that may be permanent

along the Great Lakes fresh water system. To say that this will have no

adverse effect on public health and safety is a prediction most of the

public does not accept. The commenter believes that the generic ruling

to use a dry cask storage design at any reactor site is impossible and

should be discarded. By relying on environmental evaluations done in

the 1970s before Davis-Besse construction, the NRC was remiss in its

responsibility to protect the people of Ohio from harm by its licensee.

Another commenter wants the NRC to prepare, at a minimum, an

Environmental Assessment (EA) for each site, including information on

sensitive ecosystems, wildlife, demography, meteorology, and geology.

The EA should discuss the cask's capability to withstand weather

conditions and potential catastrophic events.

Response. The potential environmental impacts of utilities using

the Standardized NUHOMS (or any of the other spent fuel casks approved

by NRC (10 CFR 72.214)) have been fully considered and are documented

in a published Environmental Assessment (EA) covering this rulemaking.

Further, as described below, the EA indicates that use of the casks

would not have significant environmental impacts. Specifically, the EA

notes the 30-plus years of experience with dry storage of spent fuel

have shown that the previous extensive NRC analyses and findings that

the environmental impacts of dry storage are small and succinctly

describes the impacts, including the non-radiological impacts of cask

fabrication (the impacts associated with the relatively small amounts

of steel, concrete, and plastic used in the casks are expected to be

insignificant), the radiological impacts of cask operations (the

incremental offsite doses are expected to be a small fraction of and

well within the 25 mrem/yr limits in NRC regulations), the potential

impacts of a possible dry cask accident (the impacts are expected to be

no greater than the impacts of an accident involving the spent fuel

storage basin), and the potential impacts from possible sabotage (the

offsite dose is calculated to be about one rem). All of the NRC

analyses collectively yield the singular conclusion that the

environmental impacts and risks are expected to be extremely small.

NRC EA's for previously approved dry casks also concluded there was

an absence of significant environmental impacts from dry cask storage

at a reactor site when they were analyzed in earlier rulemakings

addressing 10 CFR Part 72 as well as in the Commission's Waste

Confidence decisions in 1984 (August 31, 1984; 49 FR 34658) and 1989

(September 29, 1989; 54 FR 39765). In the 1984 Waste Confidence

decision, the Commission concluded there was reasonable assurance spent

fuel can be safely stored at reactor sites, without significant

environmental impacts for at least 30 years beyond expiration of NRC

reactor operating licenses. The 1989 Waste Confidence decision review

reaffirmed earlier Commission conclusions on the absence of significant

environmental impacts.

Given the Commission's specific consideration of environmental

impacts of dry storage and the absence of any new information casting

doubt on the conclusion that these impacts are expected to be extremely

small and not environmentally significant, the NRC is not convinced

that meaningful new environmental insights would be gained from either

a new site-specific EIS or EA for each site using dry storage methods.

The EA covering the proposed rule, as well as the finding of no

significant impact (FONSI) prepared and published for this rulemaking,

fully comply with the NRC environmental regulations in 10 CFR Part 51.

The Commission's environmental regulations in Part 51 implement the

National Environmental Policy Act (NEPA) and give proper consideration

to the guidelines of the Council of Environmental Quality (CEQ). The EA

and FONSI prepared as required by 10 CFR Part 51 conform to NEPA

procedural requirements. Further analyses are not legally required.

The regulation 10 CFR Part 72, Subpart K, already authorizes dry

cask storage and approves dry casks for use by utilities to store spent

fuel at reactor sites. See 10 CFR 72.214 for a listing of information

on Cask Certificate Nos. 1000 through 1003, 1005, and 1007. The purpose

of this final rule is to add one more cask to the list of casks already

approved by NRC. The cask added to the list in Sec. 72.214 by this

final rule complies with all applicable NRC safety requirements.

Finally, this final rulemaking applies to the use of this cask by

any power reactor within the United States.

H.2. Comment. One commenter stated that the January 30, 1994, reply

from NRC's Robert Bernero to Mr. Adamkus, EPA, is completely

inadequate, as is the March 1994 ``Draft Environment Assessment and

Finding of No Significant Impact'' because no consideration is given to

the site's unsuitability even for LLRW per NRC's own admission, and

``new information which could alter the original site evaluation

findings'' is ignored.

Response. This final rule does not provide any site-specific NRC

approval or address site-specific parameters that are peculiar to a

particular reactor site. The rule only adds one cask design, the

Standardized NUHOMS, to the list of approved casks available for use by

a power plant licensee in accordance with the conditions of the general

license in Part 72. Pursuant to those conditions, each licensee must

determine whether or not the reactor site parameters (including

earthquake intensity and tornado missiles) are encompassed by the cask

design bases considered in the cask SAR and SER. The EA and FONSI for

this rule are limited in scope to the Standardized NUHOMS in a generic

setting.

Unlike interim storage prescribed in 10 CFR Part 72, the in-ground

disposal of radioactive material, whether high-level or low-level waste

(HLW or LLW), must take into account the geologic, hydrologic, and

geochemical characteristics of the site or region to isolate the

radioactive waste from the accessible environment. Site criteria for

in-ground disposal of radioactive wastes enable an applicant to choose

an appropriate site, one with a combination of favorable conditions

that will be a natural barrier to retard or attenuate the migration of

any leaked radioactive material over a long period to control releases

within acceptable limits. The disposal period for LLW is on the order

of 500 years, and for HLW it is greater than 10,000 years. Therefore,

site characteristics are investigated and assessed for interim spent

fuel storage under Part 72, not to determine their suitability as a

barrier to release of radioactive material, but rather to determine the

frequency and the severity of external natural and artificial events

that could affect the safety of an ISFSI. Unlikely, but credible,

severe events are considered to determine the safety of the storage

cask design.

H.3. Comment. One commenter stated that the NRC has not approved

technologies for the use of spent fuel at the sites of * * * without

the need for additional site reviews. If that were so, no additional

site review would have been necessary at Palisades, nor would an SAR

revision or a Certificate of Compliance amendment be called for right

after the VSC-24 was certified.

Response. The approval and use of dry storage technologies under

the provisions of the general license are relatively new. Questions

were raised by members of the public about the possible effects of

earthquakes and erosion at the Palisades site on the safe storage of

spent fuel in the VSC-24 dry casks. As the agency which is responsible

for questions about compliance with regulatory requirements, which

oversees such matters as the ``cop on the beat,'' the NRC began an

independent assessment to more closely examine the behavior of the pad

at Palisades under normal conditions, under the long-term effects of

erosion, and under conditions of a postulated earthquake that might

cause the sand below or around the pad to move. The results of NRC's

assessment were documented in the NRC Final Safety Assessment of

Independent Spent Fuel Storage Installation (ISFSI) Support Pad (TAC

No. M88875). As is the case at all sites, NRC requires the cask user to

determine if the design basis for the storage technology being

considered encompasses the site parameters at the location where the

fuel is to be stored. The review at Palisades confirmed this to be the

case. As experience with use of this new design is gained,

modifications to the design described in the SAR are expected and

allowed under the provisions of 10 CFR 72.48.

H.4. Comment. One commenter wanted the environmental impacts of

alternatives, such as: renewable energy sources, conservation of

energy, shutting down the nuclear power plants, and wind and solar

power evaluated.

Response. Energy production is not the subject of this rulemaking

and alternative sources of energy are, therefore, not reasonable

alternatives requiring evaluation. This rulemaking is limited to the

addition of the Standardized NUHOMS to the list of approved casks in 10

CFR 72.214.

H.5. Comment. One commenter stated that the NRC is ignoring the

regulatory requirements of a site-specific license as to the

feasibility of using the cask or of modifying its design.

Response. This rulemaking does not cover site-specific NRC

licensees; however, the NRC is not ignoring them. Under NRC

regulations, the utility has two options in using dry cask storage of

spent fuel: (1) The licensee may apply for a site-specific license from

NRC; or (2) the licensee may use an NRC-approved cask under the general

license provisions of Subpart K of 10 CFR Part 72. However, not all

licensees may be able to use the general license provisions, either

because the fuel type they possess is not storable in any cask listed

in 10 CFR 72.214 or because none of the cask designs envelope the

reactor site parameters. The NRC is also not ignoring site-specific

license considerations relating to modifying cask designs. Quite the

contrary, the criteria that apply to modifications of an NRC-approved

cask such as the Standardized NUHOMS are the same as the criteria that

apply to modifications of site-specific ISFSIs.

H.6. Comment. Because the populations of several states and

provinces, including two-thirds of the population of Quebec, are based

along the St. Lawrence Seaway, one commenter wanted an Economic Impact

Statement conducted with a cost/benefit analysis citing possible

adverse impact on tourism and sport fishing.

Response. A regulatory analysis, which considers both benefits and

impacts of adding the Standardized NUHOMS to the list of NRC-approved

casks under Subpart K of 10 CFR Part 72, was prepared in support of

this rulemaking action. It was included as a part of the notice of

proposed rulemaking and is also included in this final rulemaking

notice. However, this regulatory analysis reflects the limited scope of

this rulemaking. Because the rulemaking does not provide any site-

specific NRC approvals, NRC did not evaluate site-specific economic

impacts.

H.7. Comment. One commenter wanted to restrict the use of the cask

to reactor sites that have responded on schedule to NRC Generic Letter

88-20, Supplement 4, ``Individual Plant Examination of External Events

(IPEEE).''

Response. IPEEE response submittals will not address dry cask

storage and are not necessary for Standardized NUHOMS use.

H.8 Comment. One commenter stated that NUHOMS must not receive

generic approval because site-specific characteristics must be

considered. The commenter stated that placing this cask on the shores

of Lake Erie is potential ecocide and the cask is not terrorist-proof.

Another commenter stated that the potential engineering problems of

storing high-level nuclear waste in a variety of climatic and geologic

regions of the United States are not considered.

Response. A utility's use of the Standardized NUHOMS, for the

storage of spent fuel in casks at a reactor site, would not have a

significant impact on the environment. This finding is supported by the

NRC safety and environmental evaluations for the Standardized NUHOMS,

including the applicant's demonstration of compliance of the cask with

NRC requirements, as well as by the 1990 rulemaking on dry cask storage

and the 1984 and 1989 waste confidence proceedings. Because the

Standardized NUHOMS can only be used by a licensee if the site

parameters are enveloped by the cask design basis, as specified in the

SAR and SER, cask storage of spent fuel near the shore of Lake Erie

within the specified parameters would not have a significant impact on

the environment.

I. The following comments relate to the transportability of dry

storage casks to an off-site location.

I.1. Comment. One commenter questioned how the cask transport

methods used at both on-site and off-site locations are related.

Response. In this rulemaking, the NRC reviewed the cask vendor's

proposed means for transporting the Standardized NUHOMS canister and

transfer cask outside the reactor buildings to the on-site storage pad

under the storage requirements of 10 CFR Part 72. This on-site movement

occurs within an owner-controlled area where access can be limited and

where operations would be safely managed by the general licensee. The

NRC did not review the Standardized NUHOMS for transport off-site, for

example to a DOE MRS or repository. Generally, off-site transport of

spent fuel occurs in public places where the shipper has fewer access

restrictions and limited control of the surroundings. Off-site spent

nuclear fuel shipments must be made in a transportation cask approved

by the NRC pursuant to NRC's regulations found in 10 CFR Part 71,

``Packaging and Transportation of Radioactive Material,'' and must also

comply with pertinent Department of Transportation (DOT) regulations.

At this time, the NRC is approving the Standardized NUHOMS for storage

only.

I.2. Comment. One commenter, citing a Wisconsin Public Service

Commission EIS for Point Beach, questioned the statement, ``The

baskets' heavier weight and larger diameter make the transportability

of an intact NUHOMS canister to an MRS site or repository

questionable.''

Response. The NRC has not reviewed the Standardized NUHOMS in this

rulemaking for off-site transportation.

I.3. Comment. One commenter wanted to know the relationship between

the Standardized NUHOMS and the NUHOMS MP187 now applying for a

Certificate of Compliance. Is the MP187 transportable? Will the

canister of all models fit into the transport overpack? Wouldn't a

utility be better off waiting for the transportable cask rather than

choosing a storage only cask that may have compatibility problems with

an MPC system?

Response. The MP-187 transportation overpack uses a canister

similar to the Standardized NUHOMS. However, it is the subject of a

separate NRC review as part of a site-specific licensing application.

Both the Standardized NUHOMS and the MP-187 share many common design

features. However, they are separate applications, and the NRC has not

been asked by the cask vendor to review whether the Standardized NUHOMS

can be transported in the NUHOMS MP187 transportation overpack.

The issue of whether a utility should consider the transportability

of dry storage casks is beyond the scope of this rulemaking.

I.4. Comment. One commenter cited a report given at the HLW

Conference at Las Vegas, in 1990, ``Integrated Spent Fuel Storage and

Transportation Systems using NUHOMS,'' by PNFSI (page 671): ``While

subsequent transfer of an intact DSC from a NUHOMS on-site transfer

cask directly to an OCRWM rail/barge is feasible, this method of

transfer is not preferred since the assemblies would be oriented top

down and the DSC bottom shield plug and grapple ring assembly would be

orientated top up, thus complicating the canister opening and fuel

handling process at the MRS or geologic repository following

shipment.'' Has NRC evaluated this situation? Has it been rectified?

Response. Because the cask vendor applied for certification of the

Standardized NUHOMS only as a storage cask under 10 CFR Part 72,

transportation of this cask is not a subject of this rulemaking.

Therefore, the NRC review of the standardized NUHOMS did not consider

the particular transportation problem described in the comment.

J. Several commenters supported the rule stating that it is

beneficial to the NRC and licensees, and it is consistent with NRC's

direction to avoid unnecessary site-specific licensing reviews. Others

disagreed and asked specific questions about NRC's approval and

oversight process.

J.1. Comment. One commenter stated that the NRC statement, ``The

proposed rule will not have adverse effect on public health and

safety,'' cannot be guaranteed and, therefore, even though it may be

convenient for the nuclear industry and the NRC to avoid site-specific

approvals, in this case these are essential for maintaining public

safety. Another commenter following the same theme questioned how the

following determination was made: ``this cask, when used in accordance

with the conditions specified in the Certificate of Compliance and NRC

regulations, will meet the requirements of 10 CFR Part 72; thus,

adequate protection of the public health and safety would be ensured.''

Response. Dry storage casks approved by the NRC for use under the

general license are of a robust design that relies on generic cask

features to ensure protection of the public health and safety.

Additional NRC site-specific approvals are unnecessary. NRC oversight

and inspections are sufficient to ensure that general licensees

implement NRC conditions on cask use. If specific concerns are raised,

the NRC also has the authority to look into them and respond as

necessary to protect public health and safety. The NRC has established

specific requirements in 10 CFR Part 72 that must be met in order to

obtain a Certificate of Compliance for a cask. The details of the

review and the bases for the NRC concluding that the cask meets the

requirements of 10 CFR Part 72 are provided in the SER. The goal of dry

cask storage technology is to store spent fuel safely. That goal, and

the effectiveness of the technology, have been demonstrated empirically

and experimentally. Different cask designs may require different types

of analysis to demonstrate their safety. Therefore, different review

methods may be appropriate to reach that conclusion. In each case, the

level of review performed is that needed to provide assurance of

adequate protection of the public health and safety.

J.2. Comment. Several commenters expressed concern over the

exemption to 10 CFR 72.234(c) granted to VECTRA to begin transfer cask

fabrication (but not use) ``to have the necessary equipment available

for use by Davis-Besse Nuclear Power Station (DBNPS) in mid-1995, and

thus enable DBNPS to maintain complete full-core off-load capability in

its spent fuel pool following the refueling outage scheduled for early

1996.'' One commenter said that seeking public comment and providing

comments is an exercise in futility because cask approval seems to be a

fait accompli. Another commenter wants no exemptions for fabrication

before certification to be allowed, stating that problems have

developed when all these exemptions are allowed.

Response. The NRC granted VECTRA's request for an exemption to

fabricate the transfer cask before issuance of the Certificate of

Compliance under its NRC-approved quality assurance program. NRC's

exemption decision made a special effort to clarify that fabrication

was entirely at VECTRA's financial risk and did not ensure favorable

consideration of VECTRA's application. The NRC's finding, based on the

SAR for the Standardized NUHOMS and the NRC's SER, concluded that

beginning fabrication before the issuance of the Certificate of

Compliance would pose no undue risk to public health and safety. Use of

the transfer cask is dependent on satisfactory completion of NRC's

certification process.

The NRC staff carefully considers the public comments received in

rulemakings to determine whether changes are needed to the proposed

rule. As noted elsewhere in this notice, several public comments

received in this and other cask-approval rulemakings have resulted in

changes to the SER and the Certificate of Compliance. For this reason,

the public comments provide useful inputs to the NRC's safety approval

process.

J.3. Comment. One commenter wanted a Regulatory Guide outlining the

requirements of an SAR for cask certification (CSAR). Requirements for

a CSAR have not been clarified. Specific criteria for a TR (TSAR) by a

vendor for a generic Certificate of Compliance need to be set.

Response. Regulatory Guide 3.61, ``Standard Format and Content for

a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask,''

dated February 1989, provides guidance for the preparation of a TSAR.

Regulatory Guide 3.62, ``Standard Format and Content for the Safety

Analysis Report for Onsite Storage of Spent Fuel Storage Casks,'' dated

February 1989, provides guidance in preparing an SAR locating an ISFSI

at a reactor site. Both Regulatory Guides identify similar information

that can be potentially useful to prospective applicants for cask

certification.

J.4. Comment. One commenter wanted to know why Pacific Nuclear

divested itself of any ownership or relationship to the VSC design in

January 1992. How does this affect proprietary material shared in these

two closely related designs? How does it affect their relationship to

the DOE MPC system?

Response. The key individual involved in the design and development

of the VSC-24, who was also involved in the design and development of

the NUHOMS design, left Pacific Nuclear and formed a new company,

Pacific Sierra Nuclear, for the commercial manufacture and marketing of

the VSC-24 storage system. The NRC has experienced no difficulty

obtaining the required safety information, including proprietary

information or answers to its questions from either firm, either before

or after divestiture. The NRC is not aware of any relationship between

the vendors. In addition, the NRC fully reviewed the health and safety

aspects of each vendor's cask design independently. The NRC did not

rely on any assumed relationship between the two vendors. Concerning

their relationship to the DOE MPC system, each vendor has to establish

its own relationship with DOE.

J.5. Comment. One commenter wanted to know how long any model of

NUHOMS has been used and if fuel has been taken out and evaluated. Has

the 24P or 52B ever been used anywhere and for how long? If not, this

is a test of a new cask at a reactor site.

Response. The NUHOMS-24P is being used at Duke Power Company,

Oconee Nuclear Station, under a site-specific license issued January

29, 1990, and at Baltimore Gas and Electric Company, Calvert Cliffs

Nuclear Station, under a site-specific license issued November 25,

1992. Monitoring and surveillance of the system is being performed

under the conditions of the site-specific license. However, there has

been no need for fuel to be removed for evaluation.

The NUHOMS-52B has not been used yet. Pre-operational testing of

the first cask system put in place under the general license is to be

performed in accordance with Certificate of Compliance, Attachment A,

``Conditions for Systems Use.'' Monitoring and surveillance of the

system will be performed under the conditions of the Certificate of

Compliance.

The first use of the Standardized NUHOMS-52B will not place plant

workers, the public, or the environment at risk. Conditions of use for

the Standardized NUHOMS-52B ensure adequate safety of the workers, the

public, and the environment. The Standardized NUHOMS-52B has been

designed and will be fabricated to well established criteria of the

ASME B&PV and ACI codes. It uses construction materials that have well

known and documented properties to provide the necessary structural

strength and radiation shielding to meet regulatory requirements. While

the Standardized NUHOMS-52B is not identical to the NUHOMS-24P, many

parallels in design and function can be drawn to demonstrate that the

Standardized NUHOMS-52B will perform as intended.

J.6. Comment. One commenter stated that even though dry cask

storage passes all NRC rules and is one of the least expensive methods,

it would seem that a different location or more expensive storage

method is worth lives, resources, and property.

Response. Based on numerous NRC reviews and growing experience with

dry cask storage technologies, the NRC has concluded that spent fuel

can be safely stored in dry casks without significant risk to the

public health and safety. More expensive storage techniques or

alternative storage locations would not provide any significant

additional public protection. Further, the storage location is a matter

of Congressional policy as reflected in Section 218(a) of the Nuclear

Waste Policy Act of 1982, which includes the following directive: ``The

Secretary [of DOE] shall establish a demonstration program in

cooperation with the private sector, for the dry storage of spent

nuclear fuel at civilian nuclear power reactor sites, with the

objective of establishing one or more technologies that the [Nuclear

Regulatory] Commission may, by rule, approve for use at the sites of

civilian nuclear power reactors without, to the maximum extent

practicable, the need for additional site-specific approvals by the

Commission.'' Section III(a) also finds that the generators of the

spent fuel have the primary responsibility to provide for the interim

storage of the spent fuel until it is accepted by the DOE.

The type of spent fuel stored in the dry cask storage systems is

one factor that allows the cost of the systems to be lower. Because the

fuel has cooled a number of years, passive cooling can be used rather

than active cooling as is required for fuel just removed from the

reactor. Passive cooling reduces the cost by not having active

components such as pumps, heat exchanger, water filters, and the

maintenance required for these components.

J.7. Comment. One commenter opposed licensing any dry cask storage

system other than the DOE multi-purpose canister (MPC) because it

minimizes handling individual fuel assemblies, standardizes

compatibility between storage sites and DOE, and reduces cost. Multiple

cask designs lead to less expertise in production, operation, and

accident management. Federal regulations need to be amended to mandate

only the use of the MPC.

Response. The DOE MPC system will not be available for general use

until well after 1997. In the meantime, additional storage capacity is

needed now at several reactor sites. Once the MPC is available for

general use, most utilities might use it. However, given the

demonstrated and immediate need of some reactors for an additional

storage capacity, and given NRC's responsibility to implement dry cask

storage under a general license pursuant to NWPA of 1982, it would not

be prudent for NRC now to require use of MPC designs that not even DOE

has yet approved.

The NRC does not agree that the number of cask designs has a

significant effect on the level of expertise available because standard

engineering and scientific skills such as mechanical and civil

engineers and health safety specialists can be hired as needed.

K. Several commenters had concerns about decommissioning issues.

K.1. Comment. One commenter, citing the draft SER, stated that

decommissioning and decontamination of reactors and reactor sites

remain uncertain at best. ``At this time, it is not known whether

demolition and removal of the HSM can be performed by conventional

methods * * *. The reinforced structure of the HSM, for example, will

require considerable effort to demolish.'' The commenter continues by

indicating that in its typical fashion of putting off until tomorrow

what it cannot deal with today, the NRC considers ``ease of

decommissioning (a) secondary consideration.''

Response. The demolition of the HSM will be more difficult than a

typical building because of the large amount of reinforced steel it

contains. However, it is technically feasible and represents a likely

level of effort similar to that required to demolish a bank vault. Bank

vaults are routinely demolished without extraordinary effort. The HSM

may become slightly radioactive from being exposed to a neutron

radiation field during the spent fuel storage period, which would

require some containment during demolition to prevent the spread of

contamination. Recognizing this, the NRC considers decommissioning a

secondary consideration compared to the safety afforded by storage of

spent fuel in dry casks.

K.2. Comment. One commenter questioned how, where to, and when the

spent fuel and casks will go? How does the decommissioning of NUHOMS

affect the reactor decommissioning plan if no repository is sited and

the pool must remain open? Another commenter expressed concern that

after the operating facility has been decommissioned, the spent fuel

pool may not be available for use in recovery of a breached DSC.

Response. The Commission determined in the Waste Confidence

decisions that sufficient repository capacity will be available, in the

first quarter of the 21st century, to accept spent fuel that is already

in storage or that will be generated during the lifetime of the reactor

licensed by NRC. In addition, the Commission determined that spent fuel

can be safely stored at reactors until it is disposed. The bases for

these determinations are extensively discussed in the Waste Confidence

decisions (54 FR 39765; September 28, 1989 and 49 FR 34658; August 31,

1984) and remain applicable today.

To operate the dry spent fuel storage area under the provisions of

the general license, a license to possess or operate a nuclear power

reactor under 10 CFR Part 50 is required. If the reactors were

decommissioned and the license terminated, and if the spent fuel were

to remain on site, a specific license issued under 10 CFR 72.40 would

be required. At the time of application for a specific license and

before the Part 50 license was terminated, the licensee would have to

address the subject of how the fuel will be repackaged for shipment to

an MRS or repository. (None of the casks now listed in 10 CFR 72.214

are approved for transportation). Decommissioning and termination of a

Part 50 license for a given reactor site must take into account the

proper disposal of any spent fuel.

L. A number of positive and negative comments were received about

the application of 10 CFR 72.48 or Item 9 of the Certificate of

Compliance to general licensees.

L.1. Comment. Several commenters questioned the application of 10

CFR 72.48 to Certificate of Compliance holders for use by a general

licensee. Some commenters believe that this regulation is being

inappropriately applied to general licensees and cask vendors. These

commenters believe that the regulation was intended to apply to site-

specific licenses issued under 10 CFR 72.40 only. One commenter cited

the parallel application of 10 CFR 50.59 to 10 CFR Part 50 licensees.

Any changes to the Certificate of Compliance and the supporting SAR and

SER need public input using the rulemaking process. Who would make the

decisions in using the terms ``unreviewed safety questions,''

``significant increase,'' and ``significant environmental impact''?

Other commenters liked this addition, stating that non-safety-

significant changes can be made in a timely and cost effective manner.

Several commenters supported the incorporation of item number 9 (in

72.48 type language) in the draft Certificate of Compliance. One

commenter wanted similar provisions made for general license holders

with recordkeeping requirements applicable to the general license

rather than the certificate holder. Changes requiring an amendment to

the certificate should be initiated by the certificate holder only.

Response. The NRC will not allow changes in the Certificate of

Compliance under 10 CFR 72.48. However, the general licensee may make

changes in the SAR under 10 CFR 72.48, unless it involves an unreviewed

safety question, a significant increase in occupational exposure, or a

significant unreviewed environmental impact. The general licensee must

make the determinations, in the first instance, that are necessary for

application of 10 CFR 72.48. The licensee must also retain its

evaluations on its records (which are subject to NRC review).

Supporting this application of 10 CFR 72.48 to the general license

are the words of 10 CFR 72.48(a)(1) which provides as follows: ``The

holder of a license issued under this part may: (i) Make changes in the

ISFSI * * * described in the Safety Analysis Report, * * * (iii) * * *

without prior Commission approval, unless the proposed change, test or

experiment involves a change in the license conditions incorporated in

the license, an unreviewed safety question, a significant increase in

occupational exposure, or a significant unreviewed environmental

impact.'' Also supporting the interpretation is 10 CFR 72.210 which

provides as follows: ``A general license is hereby issued for the

storage of spent fuel in an independent spent fuel storage installation

at power reactor sites to persons authorized to possess or operate

nuclear power reactors under Part 50 of this chapter.'' The NRC staff

is considering a rulemaking to amend NRC regulations to explicitly

state that 10 CFR 72.48 applies to general licensees.

L.2. Comment. One commenter stated that the CFR is silent on how a

vendor can change a cask SAR and certificate after the final rule. It

should be made clear for the vendor that this cask SAR (CSAR) is

generic for all United States sites. All seismic, control component,

distance, changes in length and weight, changes in transfer devices,

etc., need to be clearly defined in the proposed rulemaking for the

cask and the CSAR before public comment. Who would be liable if a

utility requested the vendor to change a certified cask design?

Response. The cask vendor can apply to the NRC for a change to the

cask certificate and SAR after the final rule is published in the

Federal Register. The vendor must propose the generic revisions to the

certificate and SAR and request NRC review of the proposed revision.

The NRC will evaluate the proposed revision in an SER, and if

appropriate, prepare a draft revised Certificate of Compliance. These

documents would then be placed in the NRC Public Document Room and a

proposed rule would be published requesting public comments on the

proposed revised Certificate of Compliance. After consideration of

public comments (and assuming an appropriate basis exists), a final

rule would be published incorporating the revision in the revised

Certificate of Compliance.

The SAR (CSAR) is not necessarily generic for all United States

operating reactor sites as the comment appears to suggest. The SAR is

pertinent for those sites that have parameters that are incorporated by

the cask design bases analyzed in the SAR. From a practical standpoint,

it is difficult for a cask vendor to foresee all possible combinations

of seismic, control component, distance, changes in length and weight,

changes in transfer devices, etc. Revisions are expected when the

vendor submits its initial application for approval. The vendor is

responsible for the certified cask design.

L.3. Comment. One commenter wanted an explanation for not allowing

buyer substitution of material for a Certificate of Compliance and that

these references should be deleted from fabrication specifications and

drawings. Does this mean that no changes in any materials are allowed

once the design is certified? If so, explain this in reference to new

models of the VSC-24 as far as materials, coatings, etc.?

Response. Under 10 CFR Part 72, the licensee is permitted to make

changes in the ISFSI as described in the SAR provided the changes do

not involve an unreviewed safety question. The licensee and cask

certificate holder must have a quality assurance (QA) program that

provides control over activities affecting quality of the identified

structures, systems, and components to an extent commensurate with the

importance to safety and to ensure conformance with the approved

design. The NRC does not want buyers (who may not be the licensee or

certificate holder) of cask materials to automatically be able to

substitute material without the necessary safety evaluations. Rather,

the licensee, through the cask certificate holder, has the ultimate

responsibility for approving any changes to ensure conformance with the

approved design. For structures, systems, and components identified as

important to safety, if alternative materials are desired to be used

and those specific materials form the basis of the safety evaluation,

it would be appropriate to identify those materials in the cask

application. Alternatively, the certificate holder may seek an

amendment to the SAR and, if necessary, a change to the Certificate of

Compliance. For other structures, systems, or components that are

needed for the design to be used or are otherwise prudent, but do not

perform a safety function and were not relied upon in the basis for

design approval, appropriate changes may be permitted provided the

licensee and the Certificate of Compliance holder document the

appropriate evaluations and use their quality assurance programs to

implement the change. New models of the VSC-24 casks are not the

subject of this rulemaking.

L.4. Comment. One commenter questioned how the draft Environmental

Assessment and Finding of No Significant Impact would remain valid if

changes to cask design and procedures can be made. Tests or experiments

could be conducted under draft Certificate of Compliance Item No. 9

(see also 10 CFR 72.48) leading to the use of a cask that does not meet

the conditions specified in the Certificate of Compliance. These

changes may adversely impact site-specific public health, safety, and

the environment.

Response. Given the limiting criteria of 10 CFR 72.48, it is

unlikely that any change would materially change the environmental

analysis. The licensee's authority under 10 CFR 72.48 does not permit

any changes that involve unresolved safety issues, changes to the

conditions for cask use in the Certificate of Compliance, significant

increase in occupational exposure, or significant environmental impact.

In the Environmental Assessment supporting this rulemaking to approve

the Standardized NUHOMS, the NRC staff evaluated various types of

accidents that could happen to the ISFSI facility. The NRC staff's

evaluation encompassed design basis accidents and concluded that no

radioactive material will be released to the environment. The NRC staff

also evaluated a worst-case accident and found that the environmental

impact is insignificant. Therefore, it is unlikely that the potential

impact from changes to cask design or tests or experiments under the

control of the licensee would introduce new environmental

considerations or impacts that differ from or exceed those as analyzed

in the Environmental Assessment. Changes in environmental impacts, as a

result of changes to the cask design or procedures, must be evaluated

by the licensee. The licensee's evaluations are available for

inspection by the NRC.

M. A number of technical clarifications and editorial issues were

raised.

M.1. Comment. One commenter stated that both the SAR and SER on

which the Certificate of Compliance is based should be dated, as was

the case for the VSC-24 Certificate of Compliance. If not, the public

will be commenting on an unfinished document that can be endlessly

revised.

Response. Both the draft SER and the SAR are dated November 1993.

These documents were revised based on public comments.

M.2. Comment. One commenter wanted page one of the Certificate of

Compliance revised to change the name ``Pacific Nuclear'' to

``VECTRA''.

Response. The Certificate of Compliance has been revised to reflect

this.

M.3. Comment. One commenter pointed out a typographical error on

page A-19 of the draft Certificate of Compliance. In the Basis

paragraph, the sentence starting, ``Acceptable damage may occur * * *''

should read ``Unacceptable damage may occur * * *''

Response. The Certificate of Compliance has been revised to correct

this.

M.4. Comment. One commenter requested clarification of Technical

Specification 1.2.16 on page A-25 of the draft Certificate of

Compliance, as to whether the Yearly Average Ambient Temperature is a

surveillance requirement or an action statement. It is unclear what

action should be taken if either of the two specified limits (Yearly

average temperature eff remains below 0.95. The SAR for

this rulemaking does not request, nor does the SER grant, exemption

from the requirement of keff = 0.95 for all accident conditions,

including misloading of 24 unirradiated fuel assemblies and optimum

moderation density.

The NRC has not yet approved the use of burnup credit in

criticality analyses for spent fuel storage and transportation casks.

The applicant did, however, analyze credit for burnup as an alternative

design acceptance basis for the NUHOMS-24P DSC, pending future

acceptance of burnup credit by NRC. As discussed in the SER, the

NUHOMS-24P DSC criticality safety is approved based on, among other

assumptions, the key assumptions of loading with irradiated fuel

assemblies with equivalent enrichment <1.45 wt% U-235, misloading

unirradiated fuel with maximum enrichment of 4.0 wt% U-235, and soluble

boron in water for wet loading and unloading. The NRC still considers

the use of the burnup-enrichment curve, Certificate of Compliance

Figure 1-1, as a fuel selection criteria, to be prudent. Its use adds

additional unanalyzed conservatism in the criticality safety margin. It

is comparable to previous NUHOMS-24P approvals. Its use would also be

consistent with the requirement that storage cask designs be, to the

extent practicable, compatible with removal of the stored spent fuel

from the reactor site, transportation, and ultimate disposition by DOE.

Therefore, the NRC disagrees with the commenters request to allow

Standardized NUHOMS-24P users the option of using the burnup-enrichment

curve.

M.12. Comment. Several commenters stated that the listing of

specific fuel types in the draft Certificate of Compliance is overly

restrictive. Allowance should be made for very similar fuel types or a

``fuel qualification table'' as proposed by the vendor should replace

the listing.

Response. The NRC agrees that allowance should be made for very

similar types of fuel to be stored. The Certificate of Compliance

provides this flexibility. The ``fuel qualification table''

consideration at this time is not subject to this rulemaking.

M.13. Comment. One commenter citing the first paragraph of page A-

27 of the draft Certificate of Compliance states that the postulated

adiabatic heatup would result in concrete temperatures being exceeded

in approximately 40 hours. As a result, it is appropriate and

conservative to perform the visual surveillance to verify no vent

blockage on a daily basis to ensure that a blockage existed for less

than 40 hours. The last sentence in the first paragraph should reflect

that the module needs to be removed from service if it cannot be

established that the blockage is less than 40 hours, not 24 hours. A

24-hour surveillance interval will adequately verify this. One

commenter cited an inconsistency in Section 3 of the draft Certificate

of Compliance. Section 3.1 indicates that a module must be removed from

service if a vent blockage is in existence for longer than 24 hours.

Surveillance Section 1.3.2 indicates that a module must be removed from

service if the concrete accident temperature criterion has been

exceeded for more than 24 hours. A vent blockage of less than 24 hours

would not cause the temperature limit to be exceeded, as explained in

Section 1.3 and the objective for the 24-hour frequency required by

surveillance 1.3.1. The apparent conflict between Section 1.3 and the

action for Surveillance Requirement 1.3.2 should be resolved. It

appears that Surveillance Requirement 1.3.2 actions are appropriate.

Response. The Certificate of Compliance has been clarified to

reflect the comment.

M.14. Comment. One commenter stated that Section 1.2.14 to

Attachment A of the draft Certificate of Compliance is unnecessary

because the time to transfer the DSC from the transfer cask to the HSM

would normally require less than 8 hours. During this time, even with

temperatures above 100 deg.F without the solar shield, any increase in

fuel clad temperature and neutron shield temperature would be small and

therefore not detrimental. Additionally, the transfer cask is open to

the atmosphere and would not pressurize.

Response. The vendor, VECTRA, has proposed this limiting condition

of operation in lieu of showing what detrimental effect might occur on

the cladding or neutron shield, should the ambient conditions involve

temperatures above 100 deg.F. The NRC concurs with this condition as

cited in Attachment A, Section 1.2.14 of the Certificate of Compliance.

N. Several commenters raised safeguards/sabotage issues.

N.1. Comment. One commenter cited the World Trade Center bombing

and the ease with which a disturbed individual recently breached

security and remained undetected at a U.S. reactor. Explosive

technology has become very sophisticated in the last 15 years since the

NRC and Sandia Laboratories studied the effect of sabotage on shipping

casks in the March 1979, NUREG-0459, ``Generic Adversary

Characteristics Summary Report.'' Another commenter made reference to

an experiment with balloons which failed. Yet another commenter

questioned the degree of protection in the spent fuel pool versus dry

cask storage. Will the cask be in a vital area? Will safeguards be

reviewed as part of the security plan? What is the effect on the

security of these casks?

Response. The NRC reviewed potential issues related to possible

radiological sabotage of storage casks at reactor site ISFSIs in the

1990 rulemaking that added Subparts K and L to 10 CFR Part 72 (55 FR

29181; July 18, 1990). NRC regulations in 10 CFR Part 72 establish

physical protection and security requirements for an ISFSI located

within the owner controlled area of a licensed power reactor site.

Spent fuel in the ISFSI is required by 10 CFR 72.212(b)(5) to be

protected against the design basis threat for radiological sabotage

using provisions and requirements as specified in 72.212(b)(5). Each

utility licensed to have an ISFSI at its reactor site is required to

develop security plans and install a security system that provides high

assurance against unauthorized activities that could constitute an

unreasonable risk to the public health and safety. The security systems

at an ISFSI and its associated reactor are similar in design features

to ensure the detection and assessment of unauthorized activities.

Alarm annunciations at the ISFSI are monitored by the security alarm

stations at the reactor site. Response to intrusion is required. Each

ISFSI is periodically inspected by NRC and annually audited by the

licensee to ensure that the security systems are operating within their

design limits. The validity of the threat is continually reviewed, with

a formal evaluation every six months by the NRC.

The NRC is currently conducting a study into the consequences of a

vehicle bomb detonated in the vicinity of an ISFSI. Following

completion of this study the NRC will make a determination as to

whether additional physical protection is warranted. In the interim,

the NRC staff believes that the inherent nature of the fuel, along with

the degree of protection provided by the approved storage means for

spent fuel, provides adequate protection against a vehicle bomb.

N.2. Comment. One commenter wanted the emergency plan updated to

include initiating events caused by unnatural occurrences, such as

sabotage, particularly for this fuel storage option. The commenter

believes that the NRC should determine if upgraded or new security

barriers are necessary for the David-Besse site.

Response. Under 10 CFR 72.212 requirements, each general licensee

must protect the spent fuel against the design basis threat of

radiological sabotage. Also, 10 CFR 72.212 requires each general

licensee to review the reactor emergency plan to determine whether its

effectiveness is decreased, and if so, to prepare the necessary changes

and obtain the necessary approvals. Therefore, the comment is already

essentially incorporated into NRC regulations.

O. Several commenters had fabrication, quality assurance, and

inspection concerns.

O.1. Comment. One commenter raised questions about NRC oversight

and requirements for proper cask fabrication by licensees. This is

based on tests of the faulty welds at the Palisades plant conducted in

July 1994 just before the cask was filled, but the test was not

reviewed.

Response. The ultimate responsibility to ensure proper cask

fabrication belongs to the user of the cask. Each Part 50 licensee

(general licensee) must have its own quality assurance (QA) program in

place to oversee vendor activities. The QA requirements apply to

design, purchase, fabrication, handling, inspection, testing,

operation, maintenance, repair, modifications of structures, systems

and components, and decommissioning that are important to safety. In

addition, certified cask vendors have NRC-approved QA programs that

control the implementation of these quality activities in a manner

appropriate to the safety significance of these activities. In turn,

the general licensee reviews, approves, and oversees its vendor's QA

programs and activities. The NRC inspects both the general licensee and

the subtiered vendors for compliance with the respective QA program

requirements and for the adequacy of the activities performed.

The faulty welds at Palisades in a loaded cask happened because the

radiographs were not read initially. If the radiographs were read in a

timely manner, the cask should not have been loaded without corrective

action first being taken. NRC oversight and involvement in the process

contributed to timely detection of the defective cask weld.

O.2. Comment. One commenter wants clarification of the quality

assurance program. NRC should have a regulatory guide for vendors with

strong criteria for audits and subcontractors, and NRC inspection

reports of fabricating facilities need to be put in the PDR. How will a

subcontractor of NUHOMS vendor be checked by NRC in the future? If a

vendor is going to continuously change subcontractors, the NRC should

inspect each cask and carefully inspect the vendor QA manual.

Response. Chapters 11 or 13 of Regulatory Guides 3.62 and 3.61,

respectively, provide guidance on acceptable quality assurance

programs. These chapters state that a QA program meeting the

requirements of Appendix B of 10 CFR Part 50 or Subpart G of 10 CFR

Part 72 will be accepted by NRC. Both Parts 50 and 72 require an audit

program. An NRC Branch Technical Position titled ``Quality Assurance

Programs for an Independent Spent Fuel Storage Installation (ISFSI) 10

CFR 72,'' implements the NRC review of quality assurance programs

submitted by applicants. NRC inspection reports are routinely placed in

the PDR except for reports containing sensitive information. Inspection

reports of NUHOMS fabrication are available in the PDR.

O.3. Comment. One commenter wanted to know if any nonconformances

have been discovered in inspection reports of any fabrication of the

NUHOMS canister. If so, what? How was this resolved? How has the QA

program for NUHOMS been reviewed? Is there a manual? How will

contractors and subcontractors be checked?

Response. A notice of nonconformance is documented in NRC

Inspection Report No. 721004/93-07 dated August 23, 1993. The NRC staff

conducted inspections in three phases at Duke Power Company, its

contractor (Pacific Nuclear Fuel Services, Inc.) and subcontractor

(Rancor, Inc.), concerning the QA activities with regard to the NUHOMS-

24P dry spent fuel storage canisters. The NRC staff found that

implementation of Duke Power Company QA Program was satisfactory, in

general. However, certain NRC requirements under Subpart G of 10 CFR

Part 72 were not met. QA activities cited in the inspection report were

documentation of nonconforming materials, parts, or components; quality

assurance records; control of purchased material, equipment, and

services; control of measuring and test equipment; instructions,

procedures, and drawings; licensee inspection; and audits.

Nonconformance corrective actions were taken and documented by Duke

Power Company. The NRC staff found these corrective actions acceptable

and so stated in letters dated January 13, 1994, and April 4, 1994. The

corrective actions taken and the implementation of the QA Program are

reviewed in periodic inspections by the NRC staff.

The latest version of the QA manual is ``VECTRA Technologies, Inc.,

Quality Assurance Manual,'' Revision 1, transmitted July 25, 1994,

which reflects the corporation's new name and organization and includes

additional changes to update the manual and clarify QA recordkeeping

commitments. The NRC staff found Revision 1 acceptable and so stated in

its letter dated August 23, 1994. In its review, the NRC staff compared

Revision 1 of the VECTRA QA Manual with Revision 3, Edition 2, of the

PNSI QA manual, which the NRC staff found acceptable by letter dated

January 28, 1993.

Contractors and subcontractors of cask vendors (or licensees) are

subject to periodic QA inspections performed by the NRC staff.

O.4. Comment. One commenter wanted to know if there is a possible

problem, and if there was, how it was resolved, with a material defect

in Swagelok tube fittings for NUHOMS?

Response. The NRC is not aware of any material defect problem with

Swagelok tube fittings on NUHOMS designs. There is no reliance on the

Swagelok fittings as part of the confinement boundary for the NUHOMS

canister. The fittings are covered by a metal plate that is welded on

after the canister is vacuum dried. Therefore, if there is a failure in

the fitting it would be the responsibility of the licensee to repair or

replace it so that the DSC can be loaded properly, but its failure

would not cause a public health and safety concern.

Finding of No Significant Environmental Impact: Availability

The Commission has determined under the National Environmental

Policy Act of 1969, as amended, and the Commission's regulations in

Subpart A of 10 CFR Part 51, that this rule is not a major Federal

action significantly affecting the quality of the human environment and

therefore an environmental impact statement is not required. This final

rule adds an additional cask to the list of approved spent fuel storage

casks that power reactor licensees can use to store spent fuel at

reactor sites without additional site-specific approvals from the

Commission. The environmental assessment and finding of no significant

impact on which this determination is based are available for

inspection at the NRC Public Document Room, 2120 L Street NW. (Lower

Level), Washington, DC. Single copies of the environmental assessment

and finding of no significant impact are available from Mr. Gordon E.

Gundersen, Office of Nuclear Regulatory Research, U.S. Nuclear

Regulatory Commission, Washington DC, 20555, telephone (301) 415-6195.

Paperwork Reduction Act Statement

This final rule does not contain a new or amended information

collection requirement subject to the Paperwork Reduction Act of 1980

(44 U.S.C. 3501 et seq.). Existing requirements were approved by the

Office of Management and Budget approval number 3150-0132.

Regulatory Analysis

The Commission has prepared a regulatory analysis on this

regulation. The analysis examines the costs and benefits of the

alternatives considered by the Commission. Interested persons may

examine a copy of the regulatory analysis at the NRC Public Document

Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of

the analysis may be obtained from Mr. Gordon E. Gundersen, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington DC, 20555, telephone (301) 415-6195.

Regulatory Flexibility Certification

As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.

605(b), the Commission certifies that this rule does not have a

significant economic impact on a substantial number of small entities.

This rule affects only licensees owning and operating nuclear power

reactors and cask vendors. The owners of nuclear power plants do not

fall within the scope of the definition of ``small entities'' set forth

in Section 601(3) of the Regulatory Flexibility Act, 15 U.S.C. 632, or

the Small Business Size Standards set out in regulations issued by the

Small Business Administration at 13 CFR Part 121.

Backfit Analysis

The NRC has determined that the backfit rules 10 CFR 50.109 and 10

CFR 72.62 do not apply to this final rule. A backfit analysis is not

required for this final rule because this amendment does not involve

any provisions that would impose backfits as defined in 10 CFR

50.109(a)(1) or 72.62(a).

List of Subjects in 10 CFR Part 72

Manpower training programs, Nuclear materials, Occupational safety

and health, Reporting and recordkeeping requirements, Security

measures, Spent fuel.

For the reasons set out in the preamble and under the authority of

the Atomic Energy Act of 1954, as amended, the Energy Reorganization

Act of 1974, as amended and 5 U.S.C. 552 and 553, the NRC is adopting

the following amendments to 10 CFR Part 72.

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF

SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

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

Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,

184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,

954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.

2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,

2234, 2236, 2237, 2238, 2282); sec. 274 Pub. L. 86-373, 73 Stat.

688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88

Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846);

Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-

486, sec. 2902, 106 Stat. 3123, (42 U.S.C. 5851); sec. 102, Pub. L.

91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135,

137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,

Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153,

10155, 10157, 10161, 10168).

Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),

Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),

10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat.

955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42

U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.

L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also

issued under secs. 2(2), 2(15), 2(19) 117(a), 141(h), Pub. L. 97-

425, 96 Stat. 2202, 2203, 2204, 2222, 2244 (42 U.S.C. 10101,

10137(a), 10161(h). Subparts K and L are also issued under sec. 133,

98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 (42

U.S.C. 10198).

2. In Sec. 72.214, Certificate of Compliance 1004 is added in

numerical order to read as follows:

Sec. 72.214 List of approved spent fuel storage casks.

* * * * *

Certificate Number: 1004

SAR Submitted by: VECTRA Technologies, Inc.

SAR Title: Safety Analysis Report for the Standardized NUHOMS

Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision

2

Docket Number: 72-1004

Certification Expiration Date: (20 years after final rule effective

date)

Model Numbers: NUHOMS-24P for Pressurized Water Reactor fuel; NUHOMS-

52B for Boiling Water Reactor fuel.

* * * * *

Dated at Rockville, Maryland this 15th day of December, 1994.

For the Nuclear Regulatory Commission.

James M. Taylor,

Executive Director for Operations.

[FR Doc. 94-31307 Filed 12-21-94; 8:45 am]

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