[Federal Register Volume 65, Number 47 (Thursday, March 9, 2000)]
[Rules and Regulations]
[Pages 12444-12460]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-5588]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG 37


List of Approved Spent Fuel Storage Casks: NAC-MPC 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 NAC International Multi-Purpose Canister cask 
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 system under a general license.

EFFECTIVE DATE: This final rule is effective on April 10, 2000.

FOR FURTHER INFORMATION CONTACT: Merri Horn, telephone (301) 415-8126, 
e-mail [email protected] of the Office of Nuclear Material Safety and 
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.

SUPPLEMENTARY INFORMATION:

Background

    Section 218(a) of the Nuclear Waste Policy Act of 1982, as amended 
(NWPA), requires that ``[t]he Secretary [of Energy] 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

[[Page 12445]]

maximum extent practicable, the need for additional site-specific 
approvals by the Commission.'' Section 133 of the NWPA states, in part, 
``[t]he Commission shall, by rule, establish procedures for the 
licensing of any technology approved by the Commission under Section 
218(a) for use at the site of any civilian nuclear power reactor.''
    To implement this mandate, the NRC approved dry storage of spent 
nuclear fuel in NRC-approved casks under a general license, publishing 
a final rule in 10 CFR Part 72 entitled, ``General License for Storage 
of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). 
This rule also established a new Subpart L within 10 CFR Part 72 
entitled, ``Approval of Spent Fuel Storage Casks,'' containing 
procedures and criteria for obtaining NRC approval of dry storage cask 
designs.

Discussion

    This rule will add the NAC International Multi-Purpose Canister 
cask system to the list of NRC-approved casks for spent fuel storage in 
10 CFR 72.214. Following the procedures specified in 10 CFR 72.230 of 
Subpart L, NAC International (NAC) submitted an application for NRC 
approval with the Safety Analysis Report (SAR) entitled ``Safety 
Analysis Report for the NAC Multi-Purpose Canister System (NAC-MPC).'' 
The NRC evaluated the NAC submittal and issued a preliminary Safety 
Evaluation Report (SER) and a proposed Certificate of Compliance (CoC) 
for the NAC Multi-Purpose Canister (NAC-MPC) cask system. The NRC 
published a proposed rule in the Federal Register (64 FR 45918; August 
23, 1999) to add the NAC-MPC cask system to the listing in 10 CFR 
72.214. The comment period ended on November 8, 1999. Five comment 
letters were received on the proposed rule.
    Based on NRC review and analysis of public comments, the NRC staff 
has modified, as appropriate, its proposed CoC and the Technical 
Specifications (TSs) for the NAC-MPC cask system. The NRC staff has 
also updated the CoC and removed the bases section from the TSs 
attached to the CoC to ensure consistency with NRC's format and 
content. The NRC staff has also modified its SER in response to some of 
the comments.
    The title of the SAR has been revised to delete the revision number 
so that in the final rule the title of the SAR is ``Final Safety 
Analysis Report for the NAC Multi-Purpose Canister (NAC-MPC) System.'' 
This revision conforms the title to the requirements of new 10 CFR 
72.248, recently approved by the Commission. The NRC staff has also 
modified the rule language by changing the word ``certification'' to 
``certificate'' to clarify that it is the Certificate that expires.
    The proposed CoC has been revised to clarify the requirements for 
making changes to the CoC by specifying that the CoC holder must submit 
an application for an amendment to the certificate if a change to the 
CoC, including its appendices, is desired. This revision conforms the 
change process to that specified in 10 CFR 72.48, as recently approved 
by the Commission. In addition, other minor, nontechnical changes have 
been made to the CoC 1025 to ensure consistency with NRC's new standard 
format and content for CoCs.
    The NRC finds that the NAC-MPC cask system, as designed and when 
fabricated and used in accordance with the conditions specified in its 
CoC, meets the requirements of 10 CFR Part 72, Subpart L. Thus, use of 
the NAC-MPC cask system, as approved by the NRC, will provide adequate 
protection of public health and safety and the environment. With this 
final rule, the NRC is approving the use of the NAC-MPC cask system 
under the general license in 10 CFR Part 72, Subpart K, by holders of 
power reactor operating licenses under 10 CFR Part 50. Simultaneously, 
the NRC is issuing a final SER and CoC that will be effective on April 
10, 2000. Single copies of the CoC and SER are available for public 
inspection and/or copying for a fee at the NRC Public Document Room, 
2120 L Street, NW. (Lower Level), Washington, DC.

Summary of Public Comments on the Proposed Rule

    The NRC received five comment letters on the proposed rule. The 
commenters included two utilities, a public interest group, and two 
letters from one member of the public. Copies of the public comments 
are available for review in the NRC Public Document Room, 2120 L 
Street, NW (Lower Level), Washington, DC 20003-1527.

Comments on the NAC-MPC Cask System

    The comments and responses have been grouped into nine subject 
areas: general, radiation protection, accident analysis, welds, design, 
thermal, structural, technical specifications, and miscellaneous 
issues. Several of the commenters provided specific comments on the 
draft CoC, the NRC staff's preliminary SER, and the TSs. To the extent 
possible, all of the comments on a particular subject are grouped 
together. The listing of the NAC-MPC cask system within 10 CFR 72.214, 
``List of approved spent fuel storage casks,'' has not been changed as 
a result of the public comments. A review of the comments and the NRC 
staff's responses follow:

A. General

    Comment A.1: One commenter stated that each cask review should be 
site specific with an Environmental Impact Statement (EIS) and a public 
hearing. The commenter further stated that the NRC should not be 
certifying numerous generic cask designs because the waste system in 
the country lacks standardization and integration.
    Response: These comments are beyond the scope of this rulemaking 
that is focused solely on whether to add a particular cask design, the 
NAC-MPC cask system, to the list of approved casks. Pursuant to the 
general license, each licensee must determine whether or not the 
reactor site parameters are encompassed by the cask design bases 
considered in the cask SAR and SER. Further, each general licensee must 
document this determination in accordance with 10 CFR 72.212. The 
rulemaking process, used by the NRC for generic cask approval, is the 
regulatory vehicle that provides opportunity for public input.
    Comment A.2: One Commenter stated that tiering on past EISs for dry 
storage is invalid for modern dry cask storage.
    Response: The NRC disagrees with the comment. The environmental 
assessment (EA) and finding of no significant impact (FONSI) prepared 
as required by 10 CFR Part 51 conform to National Environmental Policy 
Act (NEPA) procedural requirements. Tiering on past EISs and EAs is a 
standard process under NEPA. As stated in the Council on Environmental 
Quality's 40 Frequently Asked Questions, the tiering process makes each 
EIS/EA of greater use and meaning to the public as the plan or program 
develops without duplication of the analysis prepared for the previous 
impact statement.
    Comment A.3: One commenter stated that the cask should be built and 
tested before use at reactors, including the loading and unloading 
procedures.
    Response: The NRC disagrees with the comment. The NAC-MPC storage 
cask system design has been reviewed by the NRC. The basis of the 
safety review and findings are identified in the SER and CoC. Testing 
is normally required when the analytic methods have not been validated 
or assured to be appropriate and/or conservative. In place of testing, 
the NRC finds acceptable analytic

[[Page 12446]]

conclusions that are based on sound engineering methods and practices. 
The NRC staff has reviewed the analyses performed by NAC and found them 
acceptable.
    Comment A.4: One commenter objected to the use of the term 
``transportable'' in the SER, SAR and CoC and recommended that the term 
be removed because the certification is only for storage. In addition 
the cask cannot be considered a multipurpose cask until both storage 
and transportation are certified. The commenter further stated that the 
NRC should review a cask for storage and transport, and issue both 
certificates at the same time so any changes necessitated in the design 
can be accounted for in the initial approval. The commenter expressed 
concern that utilities may end up with loaded casks that cannot be 
transported.
    Response: The NRC disagrees with the comment. The use of the term 
``transportable'' in the SER, SAR, or CoC is descriptive of the 
intended functionality of the canister. The use of such terminology in 
a dry storage cask application or an NRC SER/CoC does not represent a 
certification under 10 CFR Part 71 for the transport of radioactive 
materials. Further, separate certifications are required for approval 
of a cask design (or individual components such as a canister) under 
the provisions of use for 10 CFR Parts 71 and 72. There is no 
regulatory requirement that the certification be simultaneous. The NRC 
staff's review schedule depends on applicant submittals and workload 
considerations. The NRC staff notes that the NRC, on March 25, 1999, 
approved the NAC-MPC's transportable storage canister and its contents 
for transport in the NAC-STC cask design (Docket No. 71-9235).
    Comment A.5: One commenter stated that the only multi-purpose casks 
acceptable for the high level repository would be a design that the 
Office of Civilian Radioactive Waste Management develops and therefore, 
the cask should not be called a multi-purpose.
    Response: The NRC disagrees with the comment. The name or model 
number given to the cask design is developed by the applicant. The CoC 
for the NAC-MPC is intended for the interim storage of spent fuel. In 
the case of the NAC-MPC, the same contents within the Transportable 
Storage Canister have been approved for transportation. The use of the 
NAC-MPC cask design for disposal at a high-level waste repository is 
beyond the scope of this rule. The U.S. Department of Energy (DOE) has 
not yet made final decisions regarding design or deployment for the 
cask design to be used in the high-level waste repository.
    Comment A.6: One commenter asked TSC to be defined.
    Response: TSC stands for Transportable Storage Canister.
    Comment A.7: One commenter stated that taking the cask to the pad 
should not be referred to as transport.
    Response: The term ``TRANSPORT OPERATIONS'' and its associated use 
is defined in the DEFINITIONS section of the TSs and refers to the on-
site movement of a loaded Vertical Concrete Cask (VCC) to the pad. The 
term is used consistently throughout the TSs and is pertinent only to 
activities carried out in accordance with the 10 CFR Part 72 CoC. The 
term is not associated with the offsite transport of spent fuel in 
accordance with 10 CFR Part 71.
    Comment A.8: One commenter stated that dates should be added to 
some of the references for which the date is missing.
    Response: The NRC agrees with the comment. Dates have been added, 
as appropriate, to the list of references in the SER.
    Comment A.9: One commenter asked if the lids, containments, and 
VCCs were interchangeable, the commenter felt that they should be 
interchangeable and built to specific criteria.
    Response: The NRC agrees with the comment. The specifications to 
which the storage cask design components are built, including the 
canister, lids, and vertical concrete casks, are listed in the license 
drawings contained in Section 1.5 of the SAR. Because all of the 
components for each cask are built to the same specifications, they are 
considered interchangeable to that extent.
    Comment A.10: One commenter objected to the use of the term 
``Final'' in the title of the SAR because changes will be made. The 
commenter also objected to the use of ``or'' instead of ``and'' in 
Condition 2 of the CoC because the TSs are part of the CoC.
    Response: The use of the term ``Final'' in the title of the SAR 
does not imply that changes can not be made. It is indicative that the 
NRC has approved the design and is consistent with the added regulatory 
requirement in 10 CFR 72.248 (effective February 1, 2000) to submit a 
``Final'' SAR. The use of the term ``or'' in Condition 2 is appropriate 
because it is possible to change the CoC without necessitating a change 
to the technical specifications.
    Comment A.11: One commenter stated that the TSs should be easy to 
understand (simple directions) and that there should be clear definite 
criteria.
    Response: The NRC agrees that the TSs should be understandable to a 
knowledgeable user such as licensee staff and should contain clear, 
definite criteria. An NRC goal in the development of the NAC-MPC TSs 
was to make them easy to understand and to contain clear, definite 
criteria.
    Comment A.12: One commenter asked what kind of communication 
devices are mandatory for workers and how the devices were checked 
(during movement of casks on the pads and in other high noise and low 
visibility activities) because the workers need to be in constant 
communication.
    Response: Communication devices utilized during the performance of 
cask operations are beyond the scope of this rule that certifies the 
cask design. Effective communications are an aspect of site-specific 
operating procedures to be developed by the cask users.
    Comment A.13: One commenter expressed concern that the copy of the 
SER they received was missing some pages. The commenter was concerned 
that the SER was not complete when the CoC was proposed for rulemaking.
    Response: The SER and CoC were complete at the time of the proposed 
rulemaking. The copy in the PDR is complete. During the copying process 
of copies to be dispatched for comment, apparently some pages were 
skipped by the copy machine. Subsequently, a complete copy was provided 
to the commenter. The NRC apologizes for any inconvenience that was 
caused by the missing pages.
    Comment A.14: One commenter stated that references from the 1970s 
should not be used for modern dry casks. Specifically, the commenter 
referred to a 1974 reference on tornadoes and a 1978 ALARA reference.
    Response: The NRC disagrees with the comment. The references cited 
are considered appropriate for the approval of dry cask storage system 
designs and were also utilized in the recent development of the 
standard review plan for dry cask storage systems. The NRC staff is not 
aware of technical inaccuracies in these documents that would render 
their use inappropriate. The commenter did not identify any specific 
technical inaccuracies.

B. Radiation Protection

    Comment B.1: One commenter questioned the use of a maximum value 
for contamination of the outside surface of the canister. The commenter 
felt that the contamination should be at a minimum to protect worker 
and public exposure. The commenter also questioned the use of a small 
accessible area of the canister as being

[[Page 12447]]

representative of other areas in checking for contamination and how the 
interior surface contamination of the transfer cask was verified.
    Response: Technical Specification 3.2.2 specifies the maximum 
permissible levels of removable surface contamination on the exterior 
surface of the canister. These limits are taken from guidance in NRC IE 
Circular 81-07. Experience has shown that these limits are low enough 
to prevent the spread of contamination to clean areas and are 
consistent with accepted ALARA practices.
    By circulating demineralized water through the annulus between the 
canister and transfer cask to keep the pool water out of this region 
during loading operations in the spent fuel pool, the chance of 
contaminating the canister is reduced. The highest levels of canister 
contamination are expected to be on the accessible surfaces exposed to 
spent fuel pool water. By ensuring that this area meets the Technical 
Specification limit for contamination, it is expected that the exterior 
surface of the canister will also meet the same limits. The cask user 
is also required to verify the interior surface of the transfer cask is 
not contaminated. The interior walls of the transfer cask are made of 
the same material as the canister and will be exposed to the same water 
environment as the canister during loading in the fuel pool. Therefore, 
if the transfer cask walls are not contaminated as determined by a 
survey after VCC loading, then the exterior walls of the canister 
should also be free of contamination.
    Comment B.2: One commenter asked about the surface contamination 
levels of a transfer cask after frequent use. The commenter also asked 
where the transfer cask is stored when it is not in use.
    Response: After each use of the transfer cask, the surface 
contamination levels must be verified to be less than or equal to the 
limit specified in Technical Specification 3.2.2. Additionally, in 
accordance with 10 CFR Part 20, the end-user of the cask is required to 
have a radiation protection program in place that is commensurate with 
the activities of the facility. This program is designed to ensure 
levels are maintained ALARA.
    The question on where the transfer cask is stored when not in use 
is beyond the scope of this rule. The transfer cask must be handled and 
stored in accordance with the cask user's radiation program procedures.
    Comment B.3: One commenter was concerned about the dose to a worker 
checking the top outlets or welding near the inlets and outlets or 
conducting other maintenance or surveillance activities (including for 
the casks in the future) and asked if there was gap streaming at the 
top end. The commenter further questioned where the dosimeters for 
workers were located (on the feet, shoulder height, etc) to make sure 
the readings were accurate. The commenter further stated that shielding 
must be confirmed in areas of high dose.
    Response: The occupational doses from maintenance and surveillance 
from MPC casks loaded with design basis fuel is described in Chapter 10 
of the SAR. The calculated occupational doses have been reviewed and 
have been found to be acceptable. Additionally, if the dose rates 
measured on the loaded concrete cask are equal to or less than the 
limits specified in Technical Specification 3.2.1, then there is 
adequate assurance that the shielding is in place.
    The specifics on doses received by workers performing maintenance 
and surveillance will be managed under the cask user's radiation 
protection program required by 10 CFR Part 20. This program will 
include radiation surveys of the casks so maintenance workers will know 
where the areas of high radiation occur, instruction to workers on how 
long they can stay in the area of the casks to perform maintenance and 
surveillance, and instructions for proper dosimeter placement.
    Comment B.4: One commenter questioned why a Kansas University 
Skyshine experiment was used as a benchmark and whether this had been 
rechecked by the NRC. The commenter further questioned why a skyshine 
input manual was considered proprietary.
    Response: The NRC finds conclusions based on sound engineering 
methods and practices to be acceptable. The previous version of the 
code, Skyshine II, was sponsored by the NRC. The current version, 
Skyshine III, extended the program's capabilities and was sponsored by 
Los Alamos National Laboratories. The changes to NAC's PC version of 
the Skyshine code were benchmarked against the results of experiments 
conducted by Kansas State University (KSU). These benchmark 
computations have been published in technical journals, textbooks on 
radiation shielding, and in a Sandia National Laboratory report. The 
KSU Skyshine experiment results are accepted industry wide for the 
methodology and were conducted by experts in the field of radiation 
shielding. Therefore, the NRC finds the Skyshine code to be acceptable.
    By a letter dated October 8, 1998, NAC requested that the skyshine 
manual and calculations be considered as proprietary under the 
provisions of 10 CFR 2.790. By a letter dated May 3, 1999, NRC informed 
NAC that their request to keep the skyshine manual and calculations 
proprietary was approved for the following reasons:
    a. The information has been held in confidence and is the result of 
design calculations and computer code development performed by NAC. The 
information is customarily held in confidence by NAC based on the 
significant commercial investment expended in its development;
    b. The information is not available in public sources, and NAC is 
transmitting it to the Nuclear Regulatory Commission (NRC) in 
confidence; and
    c. The public disclosure of the information would cause substantial 
harm to the competitive position of NAC. Competitors seeking to develop 
similar computer code information and calculations would have to expend 
similar amounts of time, engineering labor, and money in its 
development.
    Comment B.5: One Commenter stated that the dose consequences from a 
failure of all fuel rods with a subsequent canister breach, including 
the source term, should be evaluated because the canister can not be 
assured to be leaktight.
    Response: The NRC disagrees with the comment. Interim Staff 
Guidance (ISG) No. 3, ``Post Accident Recovery and Compliance with 10 
CFR 72.122(l)'', specifies that only credible accidents, and the 
associated consequences, be evaluated against the requirements of 10 
CFR Part 72. The hypothetical accident of a ground level breach, with 
100% fuel rod failure, is considered to be a non-mechanistic, non-
credible accident. Therefore, the applicant is not required to analyze 
the consequence of this type of accident. As indicated in SAR Section 
7.1, the confinement boundary is completely welded and inspected in 
accordance with both the ASME Code and ISG No. 4, ``Cask Closure Weld 
Inspections,'' and is leak tested to American National Standards 
Institute leaktight standards. Further, the analyses presented in the 
SAR demonstrated that the stresses, temperatures, and pressures of the 
TSC are within the design basis limits under the accident conditions 
identified by the applicant and that the confinement boundary of the 
TSC remains intact from all credible accidents. The NRC concurs with 
the evaluation in the SAR and believes that the design of the 
confinement boundary, which includes the inspection of welds, is 
adequately rigorous and meets the applicable regulations.

[[Page 12448]]

    Comment B.6: One commenter questioned how the effluent from 
flushing the radioactive gases with nitrogen would be managed, how it 
would impact workers and the public from ALARA considerations, and how 
time factored in for the release.
    Response: The canister to be unloaded will be flushed with nitrogen 
gas to remove any accumulated radioactive gases prior to initiating 
fuel cooldown. The amount of radioactive gases displaced by the 
nitrogen gas is first assessed by sampling to determine the appropriate 
radiological controls. Any gaseous effluent released from the cask 
would be processed through HEPA filters and any additional filtration 
systems a facility may have in order to filter the air from a fuel 
handling building or reactor building. All effluents released from this 
building system would have to be in compliance with the 10 CFR Part 50 
license.

C. Accident Analysis

    Comment C.1: One commenter questioned the adequacy of 
administrative controls to exclude explosions (such as a truck bomb) in 
the vicinity of an Independent Spent Fuel Storage Installation (ISFSI). 
The commenter recommended that the evaluation of a sabotage event for 
an ISFSI be updated.
    Response: These comments are outside the scope of this rulemaking. 
Spent fuel in the ISFSI is required to be protected against 
radiological sabotage under the provisions of 10 CFR 72.212(b)(5). Each 
utility licensed to have an ISFSI at its reactor site is required to 
develop physical protection plans and install a physical protection 
system that provides high assurance against unauthorized activities 
that could constitute an unreasonable risk to the public health and 
safety. The physical protection systems at an ISFSI and its associated 
reactor are similar in design to ensure the detection and assessment of 
unauthorized activities. Response to intrusion alarms is required. Each 
ISFSI is periodically inspected by NRC. Also, the licensee conducts 
periodic patrols and surveillances to ensure that security systems are 
operating within their design limits. The NRC believes that the 
inherent nature of the spent fuel storage cask also provides 
significant protection against malevolent acts.
    Comment C.2: One commenter recommended that a multi-missile 
(natural or man-made) scenario be considered in the accident analysis.
    Response: The NRC disagrees with the comment. The NRC staff, in 
Section 3.4.2 of the SER, agreed with the SAR conclusion that the 
design basis tornado-driven missiles are not capable of overturning the 
cask or penetrating the VCC. Multiple tornado-driven high-energy or 
penetrating missiles impinging simultaneously at the same cask location 
is beyond the design bases and is not considered to be credible.
    NRC regulations in 10 CFR Part 72 establish physical protection 
requirements for an ISFSI located within the owner-controlled area of a 
licensed power reactor site. Spent fuel in the ISFSI is required to be 
protected against radiological sabotage using provisions and 
requirements as specified in 10 CFR 72.212(b)(5). Further, specific 
performance criteria are specified in 10 CFR Part 73. Each utility 
licensed to have an ISFSI at its reactor site is required to develop 
physical protection plans and install systems that provide high 
assurance against unauthorized activities that could constitute an 
unreasonable risk to the public health and safety.
    Comment C.3: One commenter questioned the bounding fire analysis (8 
minute, 638 deg.F fire) and recommended that a fire initiated from an 
airplane crash or a different type of vehicle be used. The commenter 
further questioned the location of the fire at the base because flaming 
debris could land on top of the cask. The commenter also questioned 
whether lightning was considered to start a fire.
    Response: The basis for the 8-minute fire is associated with the 
time it would take to burn 50 gallons of fuel, presumably carried by 
the transporter. Other modes of transport causing the fire (such as 
airplanes, trains, delivery trucks) are not considered plausible. 
However, before using the NAC-MPC cask, the general licensee must 
evaluate the site to determine whether or not the chosen site 
parameters are enveloped by the design bases of the approved cask as 
required by 10 CFR 72.212(b)(3). Included in this evaluation is the 
verification that the cask handling equipment used to move the VCC to 
the pad is limited to 50 gallons of fuel (as detailed in Technical 
Specification 4.4.5-Site Specific Parameters and Analyses). The fire is 
assumed to burn at 1475 deg.F and is assumed to be at ground level 
since that produces the worse case scenario of fire/heated air entering 
the inlet vents of the VCC and coming into direct contact with the 
outside of the canister. Exposure of the VCC to fire of this duration 
would have little effect on the canister or its contents. Lightning 
causing a fire in the vicinity of the VCC is not considered plausible 
because of the absence of combustible material.
    Comment C.4: One commenter questioned why a seismic event or a 
landside that buries a cask is not considered credible.
    Response: Burying a cask due to seismic event, landside, or tornado 
is considered a very unlikely event. Considering the unlikeliness of 
the event and the capability of cask components and contents to be 
within their thermal limits after blockage of the air passages for 45 
hours, adverse consequences from cask burial are not considered to be 
credible. For example, casks are designed to withstand tipover 
loadings, yet tipover is designed not to happen for a certain size 
earthquake. Further, casks are analyzed to be within their thermal 
limits for up to 45 hours that would allow ample time for restoring the 
cask's cooling system to an operable status.
    Comment C.5: One commenter questioned whether the pad had been 
evaluated for an earthquake because the pad could crack and cause the 
cask to tipover. The commenter further questioned what happens to the 
pad footer and steel reinforcement during an earthquake.
    Response: The storage pad, which is beyond the scope of this cask 
design rulemaking, has not been evaluated for natural phenomena, 
including earthquakes. In accordance with 10 CFR 72.212, the cask 
operators are required to perform written evaluations to ensure that 
storage pads have been designed to adequately support the stored casks. 
The earthquake motions defined for the top surface of the pad are the 
site parameters for which the SAR has satisfactorily demonstrated that 
the cask will not overturn or slide.
    Comment C.6: One commenter questioned what happens to the berm or 
wall used as a shield during a tornado, hurricane, or earthquake and 
questioned the composition of the berm.
    Response: The use and composition of berms or walls are beyond the 
scope of this rulemaking for the cask design. If an engineered feature 
is needed to satisfy the requirements of 10 CFR 72.104(a), then these 
features are to be considered important to safety and must be evaluated 
to determine applicable quality assessment category on a site specific 
basis as required by Section 4.4.7 of the TSs. The cask design does not 
rely on engineered features to meet the Section 72.106 post-accident 
dose rate requirement.
    Comment C.7: One commenter questioned what would happen if a 
seismic event occurred while the transfer cask was attached to the top 
of the concrete shield.

[[Page 12449]]

    Response: This is not a design basis event for approval of the cask 
design's capability to safely store spent fuel. Section 72.212(b)(4) 
requires the general licensee to determine whether activities related 
to the storage of spent fuel involve any unreviewed safety question or 
change in the facility TSs, as provided under 10 CFR 50.59.
    Comment C.8: One commenter questioned if the drop test considered 
the condition of materials at the end of cask life.
    Response: As noted in SAR Section 11.2.11 and SER Section 3.3.9, 
the 6-inch end drop will exert a maximum axial deceleration of less 
than 20 g to the TSC components and the spent fuel assemblies. This g-
load is much smaller than the design basis impact load of 56.1 g for 
which the cask system structural integrity has satisfactorily been 
demonstrated. Because the margin of safety is large and the material's 
strength is not expected to degrade, the NRC believes that the cask 
system will remain capable of withstanding a 6-inch cask drop accident 
throughout the 20 year storage period.
    Comment C.9: One commenter asked a number of questions related to 
the Boral panels concerning whether the Boral poison remains in place 
under accident conditions, including cask tipover; the necessity of the 
Boral panels; how the Boral is manufactured and tested; the content of 
the Boral; the continued efficiency over time; and whether the panels 
can structurally deform.
    Response: The Boral panels are necessary for ensuring that the NAC-
MPC system meets 10 CFR Part 72 requirements for criticality safety. 
Each Boral poison panel is held in place by a stainless steel cover 
plate that is welded around its perimeter to the outer wall of the fuel 
tube. The applicant has shown that impact loads greater than those 
expected in storage accidents, including a postulated cask tipover, 
produce maximum stresses in the seal weld that are a small fraction of 
the weld material's ultimate strength. The NRC staff has found no 
credible mechanisms for deforming the poison panels in a way that would 
lead to loss or reduced effectiveness of the panels. Warping of the 
panels in relation to the tube walls to which they are attached is 
prevented by the welded stainless steel cover plates.
    Boral will be manufactured and tested under the control and 
surveillance of a quality assurance and quality control program that 
conforms to the requirements of 10 CFR Part 72, Subpart G. A 
statistical sample of each manufactured lot of Boral is tested by the 
manufacturer using wet chemistry procedures and/or neutron attenuation 
techniques. The specified minimum content of the neutron poison in the 
Boral panels (i.e., 0.01 grams of 10B per cm2) is 
ensured by the acceptance testing procedures described in SAR Section 
9.1.6.
    Boral has been used in the nuclear industry since the 1950's and 
used in baskets since the 1960's. Several utilities have also used 
Boral in spent fuel storage racks. Industry experience has revealed no 
credible mechanisms for a loss of Boral efficacy in the cask.
    Comment C.10: One commenter asked how important the minimum flux 
trap width is to criticality safety and whether it can be altered in an 
accident.
    Response: The minimum flux trap width is an important design 
parameter in limiting the system's maximum neutron multiplication 
factor (keff) under normal and accident conditions. Bounding 
structural analysis performed by the applicant indicate that flux trap 
widths may be slightly reduced as a result of side-impact loads from a 
postulated cask tipover accident. The NRC staff has analyzed the 
reactivity effects from hypothetical flux trap deformations well beyond 
those expected from tipover accidents and concludes that the resulting 
increases in keff are minuscule in relation to the large 
overestimates of keff arising from the conservatisms used in 
the applicant's criticality calculations. These conservatisms include 
modeling the spent fuel as though it were fresh, assuming flooding of 
the cask interior with unborated water, crediting only 75 percent of 
the minimum neutron poison content of Boral panels, assuming all major 
dimensions and parameters of the basket components and fuel contents 
are at their most reactive tolerances limits, and assuming the most 
reactive lateral shifting of all basket components and contents.
    Comment C.11: One commenter questioned why lateral shifting of 
tubes in disk holes was not a concern and stated that it should not be 
allowed because you can not be sure what happens in all cases.
    Response: Lateral shifting of the fuel tubes within their disk 
holes is not a concern because the criticality analysis presented in 
SAR Section 6.4.3.2 has adequately accounted for tube shifting 
variations in identifying and analyzing the most reactive 
configurations of the basket and contents.
    Comment C.12: One commenter asked for clarification on what is 
meant by pure water, whether this meant unborated water. The commenter 
further questioned whether uneven flooding was a concern and if the 
analysis had been checked.
    Response: Pure water is unborated water. Uneven flooding is not a 
concern because the basket components are designed to allow the free 
flow of water between the interior and exterior of the fuel tubes. 
Prevention of uneven flooding within and outside the fuel tubes ensures 
that the flux traps function as analyzed in limiting the maximum 
keff of the system. The NRC staff has checked and confirmed 
the applicant's analysis and conclusions regarding the design's ability 
to prevent uneven flooding of the basket.

D. Design

    Comment D.1: One commenter recommended that canister identification 
be added on the top of the structural lid per the requirements of 10 
CFR 72.236(k).
    Response: The NRC agrees with the comment. SAR Drawings 455-871 and 
-872 have been revised to show that the structural lid of the 
transportable storage canister is steel stamped with its model number, 
unique identification number, and empty weight.
    Comment D.2: One commenter recommended that the number of hose 
connections be increased to 8 around the transfer cask near the bottom 
to improve the forced air cooling capability.
    Response: The NRC agrees with the comment. Although the original 
design with two hose connections remains acceptable, increasing the 
number of hose connections to eight will more evenly distribute the 
cooling service air supply around the bottom of the transfer cask. The 
changes have been made to the SAR.
    Comment D.3: One commenter recommended that an alternative slip-on 
flange detail be permitted at the top of the fuel tube versus the butt 
welded flange detail indicated on the drawing. The commenter further 
stated that the flange should be attached with continuous full fillet 
on interior of fuel tube with intermittent weld on exterior.
    Response: The NRC agrees with the comment, as the alternative 
detail provides the same integrity as the original butt weld design. 
SAR Drawing 455-881 has been revised to show that the flange at the top 
of the fuel tube is attached to the tube using a continuous fillet weld 
on the interior of the tube.
    Comment D.4: One commenter stated that the venting of hydrogen 
should not be allowed because of the associated fire or explosion 
hazard. The commenter further stated that the design should not be 
certified if hydrogen is generated.

[[Page 12450]]

    Response: The NRC disagrees with the comment. As noted in SAR 
Section 3.4.1.2.2, the applicant anticipates that no hydrogen gas is 
expected to be detected before, or during, the loading or unloading 
operations. However, in the event that a reaction between the aluminum 
heat transfer disks and the spent fuel pool water occurs, the loading 
and unloading procedures of SAR Chapter 8, which include procedures to 
detect and remove hydrogen from the space between the shield lid and 
the top of the water during any welding or cutting operations, provide 
adequate assurance that the welders will be protected. Further, the NRC 
has licensed other storage casks that utilize aluminum heat transfer 
components, including 10 TN-32 casks and 2 NAC-I28 casks. Loading of 
these casks has not resulted in unsafe conditions for the workers.
    Comment D.5: One commenter objected to allowing the storage of 
Reconfigured Fuel Assemblies (RFA) in the same cask as intact fuel 
assemblies and believed there should be separate analysis and 
certificates. The commenter questioned whether the RFAs would remain in 
position during handling, storage, and possible unloading, or if they 
would float in a reflood and if the tubes would remain leaktight. The 
commenter asked about the composition of an RFA. The commenter asked 
how much the ``debris'' weighs, how dryness is assured, and how the 
utility can ensure that the cask is not overloaded and that the weight 
is properly distributed.
    Response: The NRC disagrees with the comment. The individual RFA 
tubes are positioned in a stainless steel container with perforated top 
and bottom end plates that retain the tubes for all conditions. The 
individual tubes have plugs at each end to retain their contents. The 
plugs are trapped in place by the top and bottom end plates. A loaded 
RFA weighs about 550 pounds and will remain in position for all 
conditions. Neither the container nor the individual tubes are closed, 
so they will drain as a canister is emptied and will refill (if 
canister reflooding is ever necessary) as the canister is filled. Thus, 
an RFA will not float. Additional description of the RFAs can be found 
in SAR Section 2.1.2.
    No actual weighing of the contents will be done. A conservative 
maximum weight of contents is analyzed in each fuel assembly location 
in the basket for each authorized loading configuration. The weight of 
an actual fuel assembly is always less than that analyzed. The debris 
that is contained in an RFA cannot exceed the weight of one fuel rod in 
each of the 64 stainless steel tubes. Because PWR fuel assemblies 
contain 179 (14x14 assembly) fuel rods or more, the weight of the RFA 
with only 64 fuel rods is much less than the conservative weight of 
contents that is analyzed.
    The intact or damaged spent fuel rods and fuel debris are loaded 
into the individual RFA tubes under water. Each tube has a drain hole 
in each end. There is a perforated plate on the top and bottom ends of 
the RFA container to permit drainage but retain gross particles and 
pieces of debris. Thus, as the transportable storage canister is 
drained, the RFA tubes and container are drained as well. The double 
vacuum drying cycle specified in the TSs and described in the canister 
loading procedures ensures the removal of any residual water for the 
canister and from the RFA.
    The slight variation in load distribution due to one or more RFAs 
has been considered and is bounded for all loading evaluations. 
Consideration of a fully loaded configuration bounds any reduced 
loading configuration. The potential for slight unevenness in loading 
does not affect canister handling because the 3-point lifting 
arrangement maintains the canister vertical for all lifts.
    Comment D.6: One commenter recommended calling the inlet a drain or 
flow tube to avoid worker confusion.
    Response: The NRC disagrees with the comment. The components are 
labeled to reflect their intended function for loading operations and 
are shown as such on the drawings. For wet unloading operations, the 
components are properly called out in the SAR procedures with respect 
to the drawings. The NRC considers it appropriate to label components 
to reflect their intended routine function under normal operations.
    Comment D.7: One commenter stated that the cask label should 
clearly identify the contents of the cask, indicating if the cask 
contains damaged fuel and the type of cladding and that the label 
should be stainless steel so it won't rust.
    Response: The NRC disagrees with the comment. Each stainless steel 
canister structural lid is stamped to identify the model number, unique 
identification number, and empty weight. Additionally, each vertical 
concrete cask has a stainless steel nameplate attached that identifies 
the model number, unique identification number and empty weight. These 
markings meet the requirements of 10 CFR 72.236(k). Space is provided 
in both instances for the addition of cask user specified information; 
however, the specific identification of cask contents is not required 
for the permanent markings affixed to the cask. The NRC notes that 
Sec. 72.212(b)(8) requires each general licensee to accurately maintain 
a record for each cask that lists the spent fuel stored in the cask. 
This record must be maintained by the cask user until decommissioning 
of the cask is complete.
    Comment D.8: One commenter questioned why only Yankee class fuel 
could be stored in the cask. The commenter further questioned whether 
burnable poison rod assemblies and TPAs would eventually be stored in 
the cask and if so, stated that the evaluation should be completed 
before the CoC is issued.
    Response: Each cask approval is specific and limited to the 
contents requested by the applicant, that in this case, is for spent 
fuels designated as ``Yankee Class'' within the application. Future 
changes to the authorized contents, if any, including different spent 
fuel assemblies and other radioactive materials associated with fuel 
assemblies, must be requested and approved in accordance with the 
regulations of 10 CFR Part 72.
    Comment D.9: One commenter stated that the documents should make it 
clear that no control components should be used in an RFA and that any 
empty position needs a dummy rod.
    Response: The NRC agrees with the comment and notes that the Fuel 
Assembly Limits (Table 2-1 of the TS) specify that intact fuel 
assemblies and RFAs shall not contain control components, and that any 
missing fuel rods in an intact fuel assembly shall be replaced with a 
dummy rod.
    Comment D.10: One commenter asked how the lifting slings were 
attached and if they had ever been tested. The commenter indicated that 
a dry run should be performed.
    Response: SAR Section 1.2.1.4.8 describes the use of the load rated 
rigging attachments and slings. All slings are designed to have 
adequate safety margin to meet the requirements of ANSI N14.6 and 
NUREG-0612 for lifting heavy loads. The administrative controls of the 
TS require the cask user to perform dry runs of certain evolutions 
prior to initial loading. These controls specify that the dry runs will 
include the heavy load activities of moving the concrete cask, moving 
the transfer cask, and lowering the canister into the concrete cask.
    Comment D.11: One commenter asked how the transfer cask is attached 
to the concrete cylinder, how high up in the air is the transfer cask, 
and what is

[[Page 12451]]

located in the vicinity that the cask could fall on.
    Response: As depicted in SAR Figure 1.1-1 and described in SAR 
Section 8.1.2, after the transfer adapter plate is bolted to the 
concrete cask top, the transfer cask, with the TSC in place, is brought 
to rest on the transfer adapter by aligning the transfer cask bottom 
door rails and connector tees with the adapter plate rails and door 
connectors. In this configuration, the bottom of the transfer cask is 
about 160 inches above the bed of a heavy-haul trailer on which the 
concrete cask is rested. The evaluation of a transfer cask drop is 
governed by NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants,'' that is subject to site-specific evaluations and is beyond 
the scope of this rulemaking.
    Comment D.12: One commenter stated that the mockup needs to clearly 
work for opening and unloading demonstration evaluations.
    Response: The NRC agrees with the comment. The administrative 
controls incorporated in the TSs require that a mockup, if used in 
place of an actual canister for dry runs, shall demonstrate the 
activities necessary to open and unload a canister.
    Comment D.13: One commenter asked whether structural lids meant the 
structural and shield lids. The commenter asked several questions about 
the shield plug concerning whether the NS-4-FR serves the same function 
as the RX-277 in the VSC-24, if the NS-4-FR was encased in the carbon 
steel, why carbon steel is used instead of stainless steel (concern 
over rusting), where the shield plug is located, and if the carbon 
steel is coated.
    Response: The transportable storage canister has a 3-inch shield 
lid and a 5-inch structural lid. After the loaded canister is placed in 
the concrete cask, a shield plug is installed over the canister. The 
shield plug is comprised of 1 inch of NS-4-FR and 4.125 inches of 
carbon steel. The NS-4-FR is encased in carbon steel. Then, a 1.5-inch 
thick carbon steel lid is used to seal the concrete cask. The carbon 
steel is coated with either Keeler and Long E-series epoxy enamel or 
Ameron PSX738 Siloxane. Therefore, rusting is not a concern. As noted 
in SER Section 5.1.2, NS-4-FR is a solid borated polymer used for 
neutron shielding. The RX-277 in the VSC-24 cask design is used as a 
neutron shielding material in the top plug assembly. The cask designer 
determines the materials to be used for the cask.
    Comment D.14: One commenter stated that ignoring full shielding on 
the bottom of the cask was a mistake and that the bottom plate needed 
to be evaluated for better shielding.
    Response: The NRC disagrees with the comment. Full shielding on the 
bottom of the cask is not necessary to provide adequate protection for 
the public. The calculated occupational doses have been reviewed and 
have been found to be acceptable. See also the response to B.3.

E. Welds

    Comment E.1: One commenter asked how much water is to be drained 
before welding and stated that the water level should be set as a 
criteria.
    Response: In SAR Chapter 8, Operating Procedures, the cask end user 
is directed to drain approximately 50 gallons of water from the 
canister.
    Comment E.2: One commenter asked how various welds are checked and 
tested, and if they were leak tight (could water seep in adding 
weight). The commenter indicated that the welding procedures were very 
important.
    Response: The examination and testing of welds is described in SAR 
Sections 9.1.1, 9.1.2 and 9.1.3. Leakage of the confinement boundary is 
not anticipated because all shop welds are volumetrically and surface 
examined in accordance with the governing ASME Code's requirements. 
Field welds (i.e. shield lid, structural lid and port cover plates) of 
the confinement boundary are liquid penetrant examined. In addition, 
the field weld on the shield lid is leak tested to ensure that it is 
leaktight. These examinations ensure that the welds will not leak.
    Comment E.3: One commenter stated that there should not be any 
exceptions on the maximum flaw size for a weld that is allowed, the 
criteria should be clear (including temperature limits). The commenter 
questioned why the postulated cracks under each liquid penetrant (PT)-
examined surface were not required to be additive for comparison to the 
critical flaw size.
    Response: The NRC accepts examination of the cask closure welds in 
accordance with Interim Staff Guidance-4, Revision 1 that allows the 
use of a multi-layer (i.e. progressive) PT examination in lieu of a 
volumetric examination. As stated in the ISG, the critical flaw size is 
determined in accordance with ASME Section XI methodology and is used 
to determine the spacing between successive PT examination layers. 
There is enough experience with the progressive PT method to conclude 
with reasonable assurance that it will detect flaws that are open to 
the surface and are of a size that would affect the serviceability of 
the weld. The probability of a flaw of this size not being detected 
because it did not break the surface is not very high because the 
liquid penetrant test is undertaken at intermediate weld pass levels. 
Thus, the concept of adding up theoretical undetected flaw sizes under 
each PT layer in a way that the sum could be greater than the 
determined critical flaw size is not considered plausible by the NRC. 
For the NAC-MPC canister, which is composed of ductile stainless steel, 
no restriction has been placed on its movement based on permissible 
flaw sizes.
    Comment E.4: One commenter asked about concerns with corner welds 
of tubes and if they could bend at the corners.
    Response: The square fuel tube is fabricated with a full-length 
longitudinal weld along the center line of one of the four sides of the 
tube. Weld examination and testing are described in SAR Sections 
9.1.1.1 and 9.1.1.2. There are no tube corner welds and, therefore, no 
concerns with bending the fuel tube at its corners, as suggested.
    Comment E.5: One commenter asked what is meant by galling of a 
weld.
    Response: Galling is excessive wear in the region of contact 
between load bearing surfaces, i.e. bolt threads during torquing, or 
trunnions on a component like a transport cask where the lifting device 
rotates in contact with the trunnions. For the vertical load test of 
the transfer cask, the loading fixture should not rotate with respect 
to the trunnions, and thus, galling of the trunnions is not expected to 
occur. The trunnion welds are inspected for permanent deformation or 
cracking, and the trunnion load bearing surfaces are inspected for 
permanent deformation and galling.
    Comment E.6: One commenter questioned whether both the structural 
and shield lids were ultrasonic tested (UT) because the SER claimed the 
lids provided redundant sealing and the commenter didn't think the 
claim should be made if they were not both UT tested. The commenter 
questioned what a progressive penetrant test was and why it could be 
used instead of the UT. The commenter further stated that the 
progressive penetrant test should not be allowed for the confinement 
boundary welds if it was not in agreement with ASME Section III, Class 
I requirements.
    Response: As stated in SAR Section 9.1.1.1 ``Nondestructive Weld 
Examination,'' the shield lid has a root and final pass liquid 
penetrant (PT) examination and the structural lid could have either 
ultrasonic examination or a progressive PT examination. For the shield 
lid weld, the liquid penetrant examinations of the root and final

[[Page 12452]]

surface, the pneumatic pressure test, and the subsequent liquid 
penetrant re-examination have been accepted by the NRC staff as 
adequate for demonstrating the weld integrity.
    The basis for the structural lid weld examination methods is 
documented in the NRC's Interim Staff Guidance-4, Revision 1 that 
allows the use of a multi-layer (i.e., progressive) PT examination in 
lieu of a volumetric examination. Because the shield lid and structural 
lid are both welded and examined, this constitutes compliance with the 
redundant sealing requirement of 10 CFR 72.236(e).
    Comment E.7: One commenter stated that a helium leak test of the 
shield lid was inadequate to provide seal reliability and that a UT 
should be completed.
    Response: The NRC disagrees with the comment. For the type of 
welding process, the environmental conditions near the weld, and the 
stainless steel weld base material, there are no known delayed cracking 
mechanisms that could cause the weld to crack after it has been 
inspected. Therefore, the liquid penetrant examinations of the root and 
final surface, the pneumatic pressure test and subsequent liquid 
penetrant examination, and the helium leak test conducted in accordance 
with the leak-tight criteria of ANSI N14.5 have been accepted by the 
NRC staff as meeting the requirements of 10 CFR 72.236(e) for redundant 
sealing of the confinement boundary.
    Comment E.8: One commenter stated that time frame for calibrating 
UT equipment was very important.
    Response: NRC agrees with the comment in that calibration of any 
equipment used in applications affecting quality needs to be assured. 
In addition, 10 CFR 72.164 ``Control of Measuring and Test Equipment'' 
and 10 CFR 50, Appendix B, XII, ``Control of Measuring and Test 
Equipment'' provide the regulatory foundation of a licensee's quality 
assurance program to ensure that these calibrations take place.
    Comment E.9: One commenter stated that the results of a PT 
examination need to be permanent and that criteria should be 
established for permanent records. The commenter requested information 
on what is required to keep records permanent.
    Response: 10 CFR Part 72, Subpart G, requires that records 
pertaining to the design, fabrication, erection, testing, maintenance, 
and use of systems, structures, and components important to safety 
shall be maintained until decommissioning of the cask is complete. This 
includes cask closure welds which are important to safety. Criteria for 
records is given in Subpart G.
    Comment E.10: One commenter questioned what was meant by 
``sufficient'' and indicated that there should be specific criteria for 
acceptability of a PT exam because ``sufficient'' is not an acceptable 
criteria. The commenter also questioned what was meant by in the field 
in the performance of welding.
    Response: The NRC accepts PT examination of field welds (meaning 
those that are not made in the fabricators shop but are made at the 
location where the spent fuel is being loaded) for the root and final 
weld passes. For the port covers the welds are relatively small (i.e. 
\1/4\ inch) fillet welds that do not lend themselves to volumetric 
examination techniques nor progressive PT examinations, and the welds 
are not subject to any significant loadings which means they basically 
perform a sealing function. Therefore, the NRC believes that PT 
examination of the port cover plate root and final welds is adequate. 
Additionally, the closure weld of the structural lid will be either 
progressively PT examined or UT'd at the option of the licensee. The 
acceptability of the progressive PT examination is documented in NRC's 
Interim Staff Guidance-4, Revision 1. The term ``sufficient'' was used 
in reference to the actual number of intermediate layers of PT 
examinations necessary to detect critical flaws. For the NAC-MPC 
``sufficient intermediate layers'' means that in addition to the root 
and final weld passes, each successive \3/8\ inch weld thickness will 
also be PT examined as shown on SAR Drawing 455-872.
    Comment E.11: One commenter questioned why the backing ring is not 
considered in analysis and how the ring affected the timing, equipment, 
and worker dose for the unloading procedures in cutting the cask.
    Response: The backing ring is utilized to aid in the welding 
process. During the welding operation, it effectively reduces fit up 
time and welding time without compromising weld integrity. The NRC does 
not believe that the inclusion of backing rings would impose any 
additional worker exposure during an unlikely unloading operation and 
when weighed against dose saved during welding, results in an overall 
reduction in dose compared to not using a backing ring.
    Comment E.12: One commenter questioned how structural and shield 
lid welds were cut open, what equipment was used, whether shims were 
used, and how the shims were removed (commenter did not think that 
shims should be used). The commenter also asked how falling debris is 
avoided.
    Response: The NAC-MPC design does not use shims for positioning the 
shield lid or structural lid on the canister. The operating procedures 
for removal of the structural lid, the vent and drain port covers, and 
the shield lid are included in Section 8.3 of the NAC-MPC SAR. Detailed 
site-specific procedures for these activities will be developed by the 
cask user. The adequacy of these specific procedures will be evaluated 
by the licensee.

F. Structural Evaluation

    Comment F.1: One commenter recommended that the certification for 
the NAC-MPC canister system be withheld because NAC has not considered 
information that questions the structural integrity of the NAC cask 
system to withstand a 30-foot drop test. The information is contained 
in Singh, K.P. and Max DeLong, ``A Structural Assessment of Candidate 
Fuel Basket Designs for Storage and Transport of Spent Nuclear Fuel'' 
(Presented at the INMM Conference, Washington, D.C., January 14-16, 
1998).
    Response: The NRC disagrees with the comment. The cask-drop test 
requirements are for transport consideration that is beyond the scope 
of this rulemaking. Certification of the NAC-MPC for listing under 10 
CFR 72.214 can only be used by the general licensee to store, not 
transport, spent fuel. However, the cask, including the fuel tube has 
been evaluated for a side impact load of 55 g, that bounds the side 
impact load associated with a cask tipover accident. The evaluation has 
satisfactorily demonstrated structural integrity of the system for its 
storage configuration. There is no basis for withholding the 
certification for the NAC-MPC storage canister system as suggested.
    Comment F.2: The same commenter objected to the NRC staff's 
discussion, in an NRC letter dated August 25, 1999, to D. Lochbaum 
regarding the Singh and DeLong paper, which the commenter interpreted 
as ``crediting'' NAC's design as conservative by considering the 
structural properties of portions of the internal basket system and 
other items. In the commenter's view, allowing design ``credit'' for 
portions of the overall structure not intended to provide gross 
structural support undermines the entire cask drop requirement. The 
commenter believed that the NAC-MPC system should not be certified if 
it does not have adequate external structure to withstand the drop test 
and protect the irradiated fuel bundles within the cask.

[[Page 12453]]

    Response: Although the 30-foot drop test is not an explicit Part 72 
requirement, the applicant referenced, in part, the NAC-STC 
transportation cask 30-foot analysis. Sections 2.7.8 and 2.7.9 of the 
SAR for the NAC-STC transportation cask, Docket 71-9235, evaluate 
structural integrity of the fuel tube under a side impact load of 55 g. 
The analysis considers the approach and information consistent with 
those discussed in the paper by Singh and DeLong. With no credit given 
to the basket structural properties other than the fuel tube and its 
interaction with the support disks, the analysis has demonstrated that 
the fuel tube is capable of withstanding a cask-drop test, thus, 
protecting the irradiated fuel bundles within. Because the load also 
bounds the side impact load associated with the cask tipover accident, 
the fuel tube is demonstrated to be capable of maintaining its 
structural integrity in a cask tipover accident. Moreover, the NRC 
staff notes that in a November 2, 1999, letter to Mr. Block to offer 
his comment on NRC's August 25, 1999, communication to Mr. Lochbaum, 
Dr. K.P. Singh, the senior author of the paper, indicated that he had 
neither reviewed NAC's design documents nor been in a position to 
comment on the nuances of NAC's design.
    Comment F.3: One commenter asked about the structural soundness of 
the inlet parts as it relates to withstanding the stress and pressure 
from the lifting jack use, and whether the inlets could be damaged or 
deformed by using the jack.
    Response: The structural design and analysis of the air cooling 
inlets when serving as bearing surfaces for lifting the storage cask 
are described in SAR Section 3.4.3.1. The stress analysis results show 
that the air inlets are structurally capable of withstanding all forces 
associated with the cask lifting operation and could not be damaged or 
deformed by using the jack. SAR Section 8.1.3 describes a procedure for 
operating the air pads and lifting jacks to transport the concrete 
cask. The jacks are installed at the bottom of the air inlet without 
the inlet screens in place. Any effects resulting from use of the air 
pads or lifting jacks is readily visible for inspection.
    Comment F.4: One commenter inquired about a Nelson stud anchor and 
the TSC support pedestal. The commenter asked if the pedestal took the 
place of the tiles used in the VSC-24 cask, why the pedestal used 2 
inches of carbon steel instead of ceramic or stainless steel because of 
a concern over rusting, how the pedestal is attached to the VCC bottom 
plate, how high is the pedestal, and if the pedestal could shift or 
deform during handling. The commenter further asked if the force had 
been calculated for possible adherence of the metal surfaces after 
rusting.
    Response: The term ``Nelson stud'' is a trade name for headed steel 
studs used for developing anchoring action between reinforced concrete 
and its steel liner plate. SAR Drawing 455-861 provides design details 
on how Nelson studs are welded to the cask bottom plate and the air 
inlet top so that the bottom plate and the concrete wall will act as an 
integral part to achieve its structural support function. As depicted 
in the same SAR drawing, the 23-inch high pedestal is a carbon steel 
weldment that consists of two major structural part, a 2-inch thick 
horizontal circular pedestal plate for providing direct bearing surface 
to the TSC and a connecting vertical ring plate assembly as a load path 
to transmit the TSC inertia load to the cask bottom and storage pad. If 
carbon steel is exposed to moist air, it may corrode. Detail B-B of SAR 
Drawing 455-862 shows that a \1/4\-inch thick stainless steel plate is 
installed between the TSC bottom and the pedestal plate, in addition to 
an \1/8\-inch thick BISCO insulation. This cover is installed on a 
sheet of fire block insulation that isolates the TSC from the VCC 
carbon steel base plate. This construction will prevent the pedestal 
plate from rusting to the canister bottom. Therefore, no adherence 
force will develop to cause any shifting, deforming, or cracking of the 
pedestal plate in handling, as suggested.
    Comment F.5: One commenter asked if there would be any deformation 
of the fuel tubes in a tipover or drop. The commenter further asked how 
the tubes and disks respond to each other when stressed and how they 
affect each other.
    Response: The support disk cutouts and the fuel tubes are sized to 
avoid binding when the cask is kept in its upright position. In a cask 
tipover accident, the support disk ligaments are in contact with fuel 
tubes and will provide support to fuel assembly inertia loads. Sections 
2.7.8 and 2.7.9 of the safety analysis report for the NAC-STC 
transportation cask, Docket 71-9235, analyzes stresses and strains of 
the fuel tubes for cask side-drop tests. SAR Section 11.2.12.3.3 
evaluates structural performance of the support disks for bounding 
impact loads. As concluded in SER Section 3.3.8, both the fuel tubes 
and support disks have been shown to behave satisfactorily for a cask 
tipover accident.
    Comment F.6: One commenter asked about the energy balance method 
used for estimating impact loads and whether it considered elastic-
plastic deformation.
    Response: The energy balance method, as used in SAR Section 
11.2.11, assumes that the potential energy associated with a 6-inch 
vertical drop of the TSC is dissipated by plastic deformation of the 
steel support pedestal. By considering the maximum force associated 
with the crushed area of and the corresponding flow stress in the 
pedestal support ring assembly, the method provides a conservative 
estimate of a height reduction of the air inlet region by 0.35 inches 
that has been evaluated to be acceptable.
    Comment F.7: One commenter questioned why the NRC did not consider 
a cask tipover off the air pad in movement or from a transporter 
tipover. The commenter asked what kind of deformation (from a tipover) 
is acceptable. The commenter further asked if the cask could roll after 
it is tipped over and what would happen if it rolled into a ditch. The 
commenter indicated that the transport path should be evaluated 
(potholes, snow, ice, gravel, etc.).
    Response: The tipover and bottom end drop analyses form part of the 
structural design basis for the NAC-MPC system design. NAC described 
the VCC drop and tipover analyses in SAR Sections 11.2.11 and 11.2.12. 
The NRC's evaluation of the vendor's analyses is described in the 
corresponding SER Sections 3.3, 11.2.11 and 11.2.12. The NRC found the 
results of these analyses to be satisfactory in that the calculated 
stresses were within the design requirements. Before using the NAC-MPC 
system, the general licensee must evaluate the foundation materials to 
ensure that the site characteristics are 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 by the licensee 
using the cask.
    Comment F.8: One commenter asked if dry unloading is evaluated for 
this cask as implied by finding F3.10 and if it is it should be 
discussed more fully in the SER and TSs.
    Response: The SAR procedures only address wet loading and unloading 
fuel from the NAC-MPC storage cask. Dry loading or unloading procedures 
are not included with this application and were not a part of the NRC 
staff's review. The SER was modified to indicate that the materials are 
compatible with wet loading and unloading operations and facilities.

[[Page 12454]]

G. Thermal Evaluation

    Comment G.1: One commenter questioned whether the EPRI Report could 
be used for stainless steel clad fuel. The commenter further stated 
that 430 deg. C must be the limit.
    Response: The NRC disagrees with the implication that improper 
cladding temperature limits were established. Because the NAC-MPC is 
designed to store both stainless steel clad and zircaloy clad fuel, the 
most restrictive temperature limit was used for both the short term and 
long term storage. These temperature limits bound both types of 
cladding and therefore, segregating the fuel is not necessary. For 
general information, the short term temperature limit of 806 deg. F and 
430 deg. C are the identical temperature except they are on different 
temperature scales.
    Comment G.2: One commenter asked about the gas in the fuel rods 
contained in RFAs concerning what it is and whether it will come out 
over time.
    Response: The fuel contained in the RFAs is by definition failed 
fuel or fuel that has cladding defects. Therefore, it is reasonable to 
assume that any fission product gases have been released from the rods 
before to placement into the MPC and that any residual gases have been 
further reduced to negligible amounts after vacuum drying the canister 
and purging it with helium.
    Comment G.3: One commenter questioned whether 200 deg.F is 
conservative enough for the water temperature during loading operations 
because of possible defects in measuring devices.
    Response: Defects in temperature measuring devices would not result 
in an operational safety problem. As a result, Technical Specification 
3.1.1 has been deleted (see response to comment H.6). The operating 
procedures now impose a 20-hour time limit supported by analysis to 
prevent the water in the canister from approaching boiling during 
welding operations and through draining.
    Comment G.4: One commenter questioned whether 24 hours for the 
helium filled canister to be in the pool is adequate to cool the 
canister before restarting loading operations. The commenter asked how 
a helium filled canister reacts in the pool and if an analysis has been 
conducted. The commenter also asked if the term ``drying'' meant the 
same thing as ``cooling''.
    Response: In Limiting Condition for Operation (LCO) 3.1.5, the term 
``drying'' means vacuum drying where the spent fuel cladding 
temperature rises due to the lack of a surrounding medium to remove 
heat. The term ``cooling'' refers to either in pool cooling or external 
forced air cooling supplied through the eight connections at the bottom 
of the transfer cask where the air is forced inside the transfer cask 
and directed up the outside of the canister, cooling the outside of the 
canister. As stated in the bases section of the TSs, the temperature of 
the fuel cladding, based on analysis, will be below 466 deg.F after 24 
hours of either in pool cooling or forced air cooling considering an 
assumed maximum decay heat loading. Therefore, after in pool cooling or 
forced air cooling, the maximum time to place the canister in the 
concrete cask is 25 hours (refer to revised LCO 3.1.6.2) that will 
result in a cladding temperature less than the limit of 806 deg.F, 
based on analysis. LCO 3.1.6.2 was revised to correct an editorial 
error on the time duration. As discussed above, the thermal analysis 
and TS bases support a 25-hour time duration instead of the 15-hour 
duration previously specified.
    Also, under LCO 3.1.5, if the LCO time limits are not met, the 
transfer cask with the helium filled canister can be placed in the 
spent fuel pool for cooling. No reaction is anticipated between the 
helium-filled canister and the pool because the canister is made of 
corrosion resistant material. Water is prevented from entering the 
canister since the shield lid welding operations have been completed 
and by the quick disconnect fittings. Therefore, the helium filled 
canister placed in the pool is bounded by the standard loading 
configuration when pool water is in direct contact with the basket 
internals.
    Comment G.5: One commenter asked for clarification of the required 
actions for LCO 3.1.6 and for forced air cooling.
    Response: If the time limits stated in LCO 3.1.6 are not met, one 
required action is to begin air cooling of the canister by supplying 
cooling air through the eight connections at the bottom of the transfer 
cask. This supplies forced air cooling to the outside surface of the 
canister before exiting out the top of the transfer cask. This action 
is allowed at the licensee's option in lieu of in pool cooling. As 
stated in the bases for the subject LCO, this forced air cooling (250 
CFM of air at 75 deg.F maximum) is sufficient to maintain the fuel 
cladding below 644 deg.F (i.e., the long term temperature limit) when 
cooled in this manner for at least 24 hours. However, because this is a 
short term event, the short term temperature limit for the fuel 
cladding (i.e., 806 deg.F) is applicable. Therefore, the time limit of 
25 hours that is applicable after the forced air cooling is stopped 
until the canister is placed in the concrete cask does not result in a 
temperature rise that would cause the short term cladding temperature 
limit to be exceeded. No temperature measurements are required to be 
taken during this action because analysis provides the justification 
for this approach. If something went wrong (e.g., air supply lost) 
during cask loading evolutions, the licensee would have the option of 
placing the helium filled canister in the spent fuel pool. TS 3.1.10 
has been added to address time limitations for canister removal from a 
concrete cask to another concrete cask or the NAC-STC transport cask.
    Comment G.6: One commenter questioned if there was an outlet air 
temperature for air cooling. The commenter further questioned whether 
forced air cooling works, if it had ever been tested and checked, and 
what happens if it does not work. The commenter stated that the short-
term temperature limits must be maintained.
    Response: For forced air cooling of the canister with air supplied 
at the transfer cask's eight lower connections at a rate of 250 cfm and 
maximum temperature of 75 deg. F, no monitoring of the outlet air 
temperature is required. Cooling in this configuration has been 
evaluated by analysis. See also the response to comment G.5 for a 
discussion of meeting short term temperature limits.
    Comment G.7: One commenter questioned whether cooling water 
recirculation flow had ever been tried and tested, how long it takes to 
connect and disconnect the system, and if the flow was through the in 
pool condenser unit. The commenter asked if there were emergency plans 
if the system does not work adequately.
    Response: The section of the technical specifications associated 
with monitoring the temperature of the water in the canister during 
loading operations has been deleted (see response to comment H.6). 
However, monitoring of the water temperature is a part of the operating 
procedures. Based on analysis, cooling of the water will not be needed 
based on analysis because 20 hours is a reasonable amount of time to 
complete the associated operations of shield lid welding, pressure 
test, and draining. However, if it appears that there is not enough 
time to complete these operations, contingencies like recirculating 
cooling water through an in pool heat exchanger or placing the transfer 
cask back in the fuel pool will be available through planning, 
procedures and rehearsal before actually loading fuel. The cooling of 
the water is not critical to this loading operation or to maintaining 
the cladding temperature limit. However, the

[[Page 12455]]

presence of water is necessary for shielding. Therefore, as long as the 
water level is maintained, it will perform its shielding function.
    Comment G.8: One commenter asked where and how the external 
temperature is measured.
    Response: The external temperature refers to an outside ambient 
temperature representative of the environment in which the transfer 
cask might be used. The method of measuring ambient temperatures is a 
site-specific consideration for the NAC-MPC system user and should be 
employed using good engineering practice.

H. Technical Specifications

    Comment H.1: One commenter indicated that the concrete and soil 
specifications do not meet the inclusion criteria of 10 CFR 72.44 and 
should not be included in the Technical Specifications.
    Response: The NRC disagrees that the specifications can be removed 
at this time. The NRC staff determined that the concrete and soil 
specifications proposed by the applicant were acceptable for ensuring 
that the cask remains within the design envelope. In order to remove 
this specification, technical justification is necessary and may be 
accomplished through the amendment process. Concrete and soil 
specifications are useful for establishing the site parameter 
conditions to ensure that once they are met, the impact force 
associated with a cask tipover accident is bounded by the design basis 
load considered in evaluating the storage cask. By complying with these 
specifications, a user is relieved of the burden of calculating the 
cask impact force for a tipover accident.
    Comment H.2: One commenter requested that the TS for the ISFSI pad 
concrete compressive strength be changed to less than 4,000 psi at 28 
days.
    Response: The NRC agrees with the comment. SAR Section 11.2.12 has 
considered a concrete compressive strength of 4,000 psi for the ISFSI 
pad bounding this revision. The staff also considered a concrete 
compressive strength of 4000 psi in its SER. SAR Section 4.4, Appendix 
A of Chapter 12, ``Site Specific Parameters and Analysis,'' Item 6c, 
has been revised to read: `` 4,000 psi at 28 days.''
    Comment H.3: One commenter requested that TS for the ISFSI pad 
concrete density be changed to 125    
150 lbs/ft 3.
    Response: The NRC agrees with the comment. NAC's additional 
calculations with a concrete density up to 150 lbs/ft3 have 
shown the maximum impact force of  45 g, the bounding impact loading 
considered in SAR Section 11.2.12. SAR Section 4.4, Appendix A of 
Chapter 12, has been revised as suggested.
    Comment H.4: One commenter requested that the soil density upper 
limit TS be modified to read ``85    130 
lbs/ft 3.''
    Response: The NRC agrees with the comment. NAC's additional 
calculations with a soil density up to 130 lbs/ft 3 have 
shown a maximum impact force of  45 g, which is bounding. SAR Section 
4.4, Appendix A of Chapter 12, has been revised as suggested to provide 
flexibility in the selection of available material.
    Comment H.5: One commenter requested that a tolerance of 
 50 be included with this site specific parameter for soil 
stiffness in order to accommodate soil variability. The commenter 
recommends that the soil stiffness be expressed as 200  k 
 300 psi/inch, where k is the sub-grade modulus.
    Response: The NRC agrees with the comment. NAC's additional 
calculations with a soil stiffness up to 300 psi/in have shown a 
maximum impact force of  45 g, which is bounding. Because the lower 
limit soil stiffness is not meaningful for determining the maximum cask 
tipover impact force, it need not be considered a soil site parameter. 
SAR Section 4.4, Appendix A of Chapter 12, ``Site Specific Parameters 
and Analyses,'' Item 6f, has been revised to read: ``k  300 
psi/in.''
    Comment H.6: One commenter recommended that LCO Section 3.1.1, 
``Canister Water Temperature'' and its basis be removed from the TSs 
because this process variable does not represent a significant risk to 
the public health and safety and is not consistent with the inclusion 
criteria of 10 CFR 72.44. The commenter recommends that the TS be 
modified to add air cooling of the canister as an alternative.
    Response: The NRC agrees with the comment that the canister water 
temperature technical specification can be removed from Chapter 12 
because defects in temperature measuring devices would not result in an 
operational safety problem. The operating procedures of Chapter 8 of 
the SAR have been modified to remove the reference to the subject LCO 
and to include the 20-hour time limit associated with the rise in 
canister water temperature after its removal from the spent fuel pool 
to the completion of draining operations. This limit is necessary to 
ensure that water remains in the canister for shielding purposes but is 
not critical to ensuring adequate cooling of the fuel cladding. 
However, vacuum drying and transfer operations are both controlled by 
time limits through the TSs because they contribute significantly to 
the temperature rise of the fuel cladding during these loading 
operations.
    Comment H.7: One commenter noted that the time for vacuum drying is 
not defined consistently in the TSs and recommended the use of 
``completion of canister draining operations'' as the definition. The 
commenter also recommended revising the bases section of the TS to 
address forced air cooling.
    Response: The NRC agrees with the comment. The associated 
surveillances in Surveillance Requirement (SR) 3.1.5.1 and SR 3.1.5.2 
have been changed to monitor elapsed time from the completion of 
canister draining operations until the start of helium backfill. Also, 
the NRC agrees that forced air cooling (at 250 CFM with 75 deg.F 
maximum air temperature for 24 hours minimum) be permitted as an 
alternative cooling method under the required actions section of LCO 
3.1.5.
    Comment H.8: One commenter recommended that LCO 3.1.5.2 be revised 
to clarify the term ``in pool cooling'' and to revise the Required 
Action to allow air cooling.
    Response: The NRC disagrees with the comment and believes the LCO, 
including the term ``in pool cooling,'' is adequate. The comment lacks 
specifics as to what is being proposed and if some other cooling 
configuration is planned then details regarding that cooling 
arrangement need to be presented.
    Comment H.9: One commenter recommended that the Technical 
Specifications contain a consistent definition of the time duration in 
LCO 3.1.6.1 and SR 3.1.6.1.
    Response: The NRC agrees with the comment. However, the initiation 
of the time duration has been modified to ``from the introduction of 
helium backfill'' to be consistent with the previous LCO 3.1.5 and not 
``from the completion of backfilling'' as requested in the comment. The 
consistency between LCO's 3.1.5 and 3.1.6 is necessary to avoid any 
unaccounted time for heatup of the canister and contents during loading 
operations.
    Comment H.10: One commenter requested that the 1,000 cfm value in 
Required Action A.2.1 of LCO 3.1.5 and the supporting bases be changed 
to 250 cfm.
    Response: The NRC agrees with the comment. Air at 250 cfm with 
75 deg.F maximum temperature for 24 hours minimum is an adequate 
cooling rate. The Required Action and the bases have been changed. 
Required Action A.1.2

[[Page 12456]]

was also changed to add eight connections to supply cooling air instead 
of the current two connections to ensure even air distribution around 
the canister.
    Comment H.11: One commenter recommended that the Bases for SR 
3.1.6.2 be revised to allow forced air cooling.
    Response: The NRC agrees with the comment and has added the words 
``or forced air cooling'' to the last sentence in the Bases Section SR 
3.1.6.2, because forced air cooling is a permissible cooling option.
    Comment H.12: One commenter recommended that the TS for fuel 
cooldown requirements addressing wet unloading be clarified to only be 
applicable for licensees maintaining spent fuel pools beyond dry fuel 
storage or be deleted.
    Response: The NRC agrees with the comment. The intent of the first 
note in LCO 3.1.7 was that this technical specification only applies to 
wet unloading operations using a spent fuel pool. Interim Staff 
Guidance No. 2, ``Fuel Retrievability'' and No. 3, ``Post Accident 
Recovery and Compliance with 10 CFR 72.122(i)'' state that spent fuel 
pools are not required to be maintained for dual purpose designs.
    Comment H.13: One commenter noted an inconsistency between the SAR 
and TSs concerning the canister pressure test value and stated that the 
correct value is 50 psig.
    Response: The NRC agrees with the comment. However, because the 
test pressure is not invoked by other parts of the technical 
specification, it has been removed from the table for canister limits. 
The operational procedures remain unchanged and still specify a 50 psig 
pressure test.
    Comment H.14: One commenter recommended that the TSs be revised to 
reflect the latest NRC-accepted format, i.e., the UMS TSs.
    Response: Large-scale changes to re-format the NAC-MPC TS similar 
to those of the NAC-UMS or other cask rulemakings should be 
incorporated through the amendment process. Focused comments modeled 
after the NAC-UMS regarding the implementation of individual technical 
specifications have been addressed separately and incorporated in this 
rulemaking action.
    Comment H.15: One commenter stated that the note for LCO 3.1.7 
concerning applicability should be located at the top of the page 
because it was confusing where it is currently located.
    Response: The NRC disagrees with the comment. The note is directly 
below the APPLICABILITY statement and is intended to clarify the 
operations for which the technical specification is applicable. The 
APPLICABILITY statement and its location are in accordance with the 
standard format for technical specifications.
    Comment H.16: One commenter stated that TS 3.1.7 should be 
clarified to make it clear that the transport operations mentioned are 
limited to onsite transport to and from the pad.
    Response: The NRC disagrees with the comment. The term TRANSPORT 
OPERATIONS is clearly defined in the technical specification 
DEFINITIONS and includes all activities involved in moving a loaded 
NAC-MPC concrete cask and canister to and from the ISFSI pad. Further 
clarification of the term is not warranted.
    Comment H.17: One commenter asked what is meant by the terms 
``outside of the fuel handling facility'' and ``external to the 
facility'' in LCO 3.1.9. The commenter further questioned whether this 
TS could be used for dry transfer at the pad.
    Response: The terms ``outside of the fuel handling facility'' and 
``external to the facility'' refers to handling operations of a 
transfer cask outside of a covered or heated facility as described in 
the Bases for the TS. The intent of the specification is to ensure that 
the structural integrity of the transfer cask and its capability to 
handle and shield a loaded canister is maintained for the temperatures 
experienced by the ferrous materials of the transfer cask.
    A dry unloading operation of spent fuel in the canister was not 
requested or explicitly described in the SAR and thus is not currently 
allowed for the NAC-MPC system and is beyond the scope of this 
rulemaking. The NAC-MPC system is designed to facilitate, using the 
transfer cask, the dry transfer of a closed canister to the NAC-STC 
transport cask without the need to unload the canister in a pool. This 
dry transfer from a vertical concrete cask used for storage to the NAC-
STC transport cask would be carried out at a facility that meets both 
the heavy-loads and overall regulatory requirements for licensed 
operation, and could be located at or adjacent to the ISFSI pad. Site-
specific evaluations and procedures for these operations, consistent 
with the technical basis established in the storage and transport cask 
SARs, are required to be developed by the cask user.
    Comment H.18: One commenter stated that utilities should not be 
allowed to use the provisions of Surveillance Requirement (SR) 3.0.2 
repeatedly and that allowance for operational convenience should not be 
provided.
    Response: The NRC disagrees with this comment. As stated in the 
Bases for this specification, the 25 percent extension facilitates 
surveillance scheduling and considers facility conditions that may not 
be suitable for conducting the surveillance. The 25% extension does not 
significantly degrade the reliability that results from performing the 
surveillance at its specified frequency because the most probable 
result of any particular surveillance being performed is a verification 
of conformance. This provision is consistent with the standard format 
for TSs.
    Comment H.19: One commenter stated that the Bases for TS 3.1.1 
should describe what is meant by ``transfer cask and canister in 
position'', what is meant by on top of the concrete shell and the 
actual height, and the doors that open at the base and how they work in 
loading and unloading. The commenter further asked if the procedures 
had been evaluated for the reverse in unloading and if a dry run had 
been conducted. The commenter also thought that sampling for water 
temperature should begin at 12 hours instead of 18 hours.
    Response: The background section of the Bases for TS 3.1.1 contains 
an appropriate amount of detail for an overview of canister and 
transfer cask operations pertinent to the specification of maximum 
canister water temperature. Further descriptions of transfer operations 
are located in Chapters 1 and 8 of the SAR and the NRC staff's SER, 
including a description of the transfer cask relative to the concrete 
cask during transfer operations, component dimensions that detail the 
height of the concrete cask and transfer cask designs, and operation of 
the shield doors during transfer operations. The start time for 
monitoring water temperatures was determined based on a bounding 
conservative analysis found to be adequate by the NRC staff. Detailed 
site-specific loading and unloading procedures are to be developed by 
the cask user based on the technical basis established in the SAR . The 
performance of site-specific dry runs including a canister unloading 
procedure before the initial system loading is specified in the TS as 
Administrative Control 5.2.
    In response to comment G.3, TS 3.1.1 and its associated Bases have 
been removed. The NRC staff agrees that the monitoring of canister 
water temperatures is more appropriately controlled in the detailed 
site-specific operating and welding procedures. Because the welds are 
ultimately

[[Page 12457]]

examined for acceptance, there would be an insignificant benefit to 
health and safety of the public by controlling the canister water 
temperatures in the TS.

I. Miscellaneous

    Comment I.1: One commenter asked what kind of deformation of the 
cask was acceptable in the 30-foot drop test.
    Response: The 30-foot drop test is a hypothetical accident 
condition in 10 CFR Part 71 and is not evaluated for storage. The 
comment is beyond the scope of this rule.
    Comment I.2: One commenter questioned the use of a heavy haul 
trailer instead of a transporter.
    Response: A heavy-haul trailer is described in the application as 
the method for moving the loaded vertical concrete cask from the fuel 
handling facility to the ISFSI pad. The method of transport is a site-
specific consideration and is subject to the required evaluations under 
10 CFR 72.212 to be performed by the cask user to ensure that the NAC-
MPC system is used within its analyzed design basis.
    Comment I.3: One commenter asked the definition for a post-shutdown 
decommissioning activities report (PSDAR).
    Response: A PSDAR is required to be submitted by reactor licensees 
no later than 2 years after the permanent cessation of operations. The 
PSDAR describes planned decommissioning activities, a schedule for 
accomplishment of significant milestones, an estimate of expected cost, 
and documents that environmental impacts associated with site-specific 
decommissioning activities have been considered in previously approved 
environmental impact statements. The licensee must submit a license 
amendment request if all of the environmental impacts of 
decommissioning have not been considered in existing environmental 
assessments.
    Comment I.4: One commenter asked how a lift limit of 3 inches for 
air pad use could be enforced and whether an air pad has ever failed. 
The commenter further questioned what happens if an air pad deflates or 
bursts while in use. The commenter also asked how smooth the pad needs 
to be for air pads to work, if they can work over ice, and how they are 
removed.
    Response: The maximum lifting height of 6-inches maintains the NAC-
MPC system within the design and analysis basis during transport 
operations of the loaded concrete cask to the ISFSI pad. The NAC-MPC 
system has been evaluated and found acceptable for a 6-inch VCC drop 
that bounds the failure of the air pad. An air pad creates an air 
``filler'' between the inflated air cushion and the supporting surface. 
A reasonably smooth supporting surface, such as an ISFSI pad, 
facilitates optimum performance of an air pad. From a performance 
standpoint, an air pad would be able to work over a supporting surface 
coated with ice, although this is obviously not a desired condition for 
cask movement operations. It is the general licensee's responsibility 
to limit the VCC lifting height to allowable values. The lift height 
requirements are specified in TS LCO 3.1.8. Surveillance requirements 
require verification that VCC lifting requirements are met after the 
VCC is lifted to install or remove the air pad, and prior to moving the 
VCC to and within the ISFSI.
    Comment I.5: One commenter stated that the inlet and outlet vents 
(and screens) need to be checked for blockage due to snow and ice, bird 
nests, leaves, sand, etc., and that the screens should be cleaned. The 
commenter asked how the outlets are visually inspected each day and 
asked if the inlets and outlets were non-planar.
    Response: The TSs require the cask user to establish a thermal 
monitoring program for each cask. The program entails daily 
measurements of inlet and outlet air temperatures and visual inspection 
of the inlets and outlets or other appropriate actions for any 
unexplained reading. As a result of the daily surveillances, 
appropriate actions are to be taken in response to abnormal indications 
that would include the clearing of any blockages associated with the 
air passages. The cooling air pathways are non-planar and designed to 
minimize radiation streaming at the inlets and outlets.
    Comment I.6: One commenter asked that the acceptably low amount of 
water and potentially oxidizing material remaining in the TSC be 
specified.
    Response: The term ``acceptably low amount of water and potentially 
oxidizing material remaining in the TSC'' refers to the 1 gram-mole 
limit for oxidizing species recommended in PNL-6365, ``Evaluation of 
Cover Gas Impurities and Their Effects on the Dry Storage of LWR Spent 
Fuel.'' As stated in this report, if the amount of oxidizing species is 
less than the 1 gram-mole limit, damage to the fuel cladding as a 
result of fuel oxidation will be precluded.
    Comment I.7: One commenter asked the difference between a suction 
pump and a vacuum pump, and why a suction pump is used. The commenter 
further questioned the amount of water removed, the basis for the 
specific amount, and why the quantity is not the same for each plant.
    Response: A suction pump is used to remove water from the canister 
cavity. Approximately 50 gallons of water corresponding to an air space 
of about 3 inches by 70 inches in diameter are removed from every cask 
(independent of which plant is using the cask) to keep moisture away 
from the regions that need to be welded (e.g., shield lid-to-shell 
weld, etc.). Removal of this amount of water is adequate to perform the 
welding operations and still provide enough shielding to the workers 
performing the welding and inspection operations. On the other hand, a 
vacuum pump is used to remove residual moisture, air, and other gases 
during vacuum drying after all of the water has been removed from the 
TSC. Removal of the water and vacuum drying reduce the quantity of 
oxidizing species in the cask to below 1 gram-mole recommended in PNL-
6365, ``Evaluation of Cover Gas Impurities and Their Effects on the Dry 
Storage of LWR Spent Fuel.'' As stated in this report, if the amount of 
oxidizing species is less than the 1 gram-mole limit, damage to the 
fuel cladding as a result of fuel oxidation will be precluded. The 
amount of water removed is specific to this cask-design to facilitate 
welding operations and for ALARA considerations, and is not appropriate 
as a specific criterion for other cask designs.
    Comment I.8: One commenter asked if all water evaporates due to the 
vacuum, even the water in gas or in solids, and fuel debris inside the 
tubes.
    Response: After most of the water has been removed from the cask, 
there may be a small amount at the bottom of the cask trapped in 
crevices or other small confined spaces that the suction pump cannot 
remove. The combination of the heat from the spent fuel and the low 
pressure (i.e., 3 mm mercury pressure) during vacuum drying will aid in 
the removal of residual water and moisture from the cask. As noted in 
the previous response (response to comment I7), the vacuum drying 
procedures described in SAR Section 8.1 will ensure there is less than 
1 gram-mole of oxidizing species in the TSC.
    Comment I.9: One commenter questioned the makeup of the pool water, 
whether the canister changed the composition, what kind of chemical 
reactions can take place, whether they have been evaluated, and who 
checks the water.
    Response: The maintenance of the spent fuel pool water chemistry is 
beyond the scope of a 10 CFR Part 72 cask review. However, a Part 72 
cask

[[Page 12458]]

review does include consideration of chemical and galvanic reactions 
that may take place while a storage canister and associated hardware 
are in the spent fuel pool. The materials employed for the transfer 
cask and the TSC are compatible with wet loading and unloading 
operations and facilities, and no reactions that affect the spent fuel 
pool chemistry or water quality are expected.
    Comment I.10: One commenter asked who the experienced fabricators 
are who will ensure the process chosen for a durable cask.
    Response: In general, NRC reviews and approves the applicant's 
quality assurance (QA) program as described in SAR Chapter 13. However, 
the cask user is ultimately responsible for ensuring the fabricator's 
QA programs comply with 10 CFR Part 72, Subpart G. Additionally, most 
storage cask fabricators are certified by the American Society of 
Mechanical Engineers and are N-Stamp Certificate holders. The N-Stamp 
Certificate is a certificate that enables a vendor to fabricate 
certified components for nuclear applications.
    Comment I.11: One commenter asked if the characteristics of the 
epoxy enamel have been checked and considered, and referred to a 
problem at Trojan concerning curing time.
    Response: For the NAC-MPC cask, the applicant demonstrated in SAR 
Section 3.4.1 that there will be no adverse reactions caused by the 
epoxy enamel coating. The NRC concurs with the SAR evaluation and 
concludes the designs of the TSC and transfer cask meet the regulatory 
requirements. The NRC staff has reviewed the problems at Trojan with 
basket coatings and has concluded that the Trojan issues do not affect 
our acceptance of the NAC-MPC coating.
    Comment I.12: One commenter questioned whether cobalt impurity and 
other contaminants had been fully evaluated for interaction concerns in 
storage and unloading.
    Response: The level of cobalt impurity and other contaminants have 
been evaluated in determining the source term and dose rates that are 
applicable to loading, storage, and unloading operations. The cobalt 
and other contaminants are mainly gamma emitters that will increase the 
dose rate on the surface of the concrete cask. The source term and dose 
rate evaluations have been reviewed and have been found to be 
acceptable.
    Cobalt is an unintended impurity element that is incorporated in 
fuel component materials during fabrication. Accordingly, there is such 
a small amount of cobalt (i.e., parts per million concentration) and 
other impurities in fuel component hardware that no reactions with 
other cask components during loading, storage, or unloading is 
expected.
    Comment I.13: One commenter asked what transfer operations ocur in 
loading and unloading in relation to the use of lead bricks in the 
transfer cask.
    Response: Section 3.1.4.2 of the SER indicates that the temperature 
of the lead bricks during transfer operations are well below the 
melting point of this material. The use of the words ``transfer 
operations'' in this sentence refers to the time that the TSC is loaded 
inside the transfer cask. Thus, the highest temperature that the lead 
bricks will experience (i.e., 191 deg.F, as noted in SAR Section 4 and 
Table 4.1-4) is expected to occur only when the TSC contains design 
basis fuel and is loaded inside the transfer cask.
    Comment I.14: One commenter asked what temperatures would be 
expected if vacuum drying or helium refill took longer than expected.
    Response: In general, the longer vacuum drying or helium transfer 
takes, the higher the temperatures will be. The rate at which these 
temperatures would increase is shown graphically in SAR Figure 4.4.3-5, 
``History of Maximum Component Temperatures for the Nominal Transfer 
Conditions''. However, the temperatures of components like the fuel 
cladding are prevented from exceeding their respective temperature 
limit of 806 deg.F by imposing time limitations during vacuum drying 
and helium transfer operations. If these time limits are exceeded, 
required actions are imposed that would prevent the temperature limits 
from being exceeded.
    Comment I.15: One commenter asked what happens if the fuel reaches 
the temperature limit when conducting a ultrasonic test and if the test 
is done when the TSC is in the transfer cask.
    Response: The welding during loading operations and their 
associated examinations are performed while the canister is in the 
transfer cask. The NAC-MPC is designed and operated to preclude the 
spent fuel from reaching its cladding temperature limits. Therefore, 
the possibility of performing a UT examination (which is optional to 
the licensee in lieu of a progressive PT) while the fuel cladding is at 
its maximum temperature limit is very remote, if not non-existent. 
However, if the licensee was concerned that the cladding temperature 
limit was being approached, the licensee would follow the technical 
specifications and initiate forced air cooling or in pool cooling, and 
there would be no adverse consequences.
    Comment I.16: One commenter asked how can a cask user be certain of 
the temperature of the lead in the transfer cask. The commenter further 
questions whether the cask user would know if the lead slumps and hot 
spots form on the outside of the transfer cask.
    Response: The temperature of the lead being below its melting point 
is assured by design analysis, thermal testing of the first loaded 
canister above a threshold heat load, and by operating procedures. 
During unloading, if the canister was placed in the transfer cask for a 
relatively long period of time (approximately 48 hours for maximum 
decay heat load) without commencing the cool down in accordance with 
LCO 3.1.7, some material temperature limits could be exceeded. 
Therefore, a new LCO 3.1.10 has been added to provide restrictions on 
the time a canister can be in the transfer cask during unloading 
operations.
    Comment I.17: One commenter questioned how the NS-4-FR neutron 
shielding could have a high hydrogen content and be fire resistant. The 
commenter further questioned if hydrogen gas could be created from the 
neutron shielding.
    Response: The NS-4-FR material consists of many elements including 
hydrogen. The chemistry of the material (e.g., the way the elements are 
bonded to one another) contribute significantly to the fire retardant 
capability of the NS-4-FR. Even though the material contains hydrogen, 
the ingredients were selected so that the NS-4-FR resists fire and 
hydrogen gas generation that could cause the material to combust.
    Comment I.18: One commenter asked if all the chemical analysis for 
a cask drop or tipover in the transfer cask had been evaluated for 
possible interaction due to water leaks or gas generation.
    Response: Cask drops and tipover analyses of the transfer cask are 
beyond the scope of the review.
    Comment I.19: One commenter questioned why the word ``if'' was used 
in describing the need for girth welds. The commenter stated that they 
should know if it is needed.
    Response: The NRC agrees with the comment. The SAR drawings 
indicate that both seam and girth welds will be used to fabricate the 
TSC. The SER has been modified accordingly.
    Comment I.20: One commenter asked about lead slumping.
    Response: Lead slumping is a term that describes the metal flow 
processes that can occur due to impact, stress, or softening at 
temperatures close to the melting point of lead (e.g., around

[[Page 12459]]

600 deg.F). This phenomenon would only be a concern for the lead that 
is in the annulus of the transfer cask while the TSC is contained 
inside. When the transfer cask is not being used, the lead is assumed 
to be at ambient temperatures. Further, the calculated maximum 
temperature of the lead during transfer of the TSC from the spent fuel 
pool facilities to the VCC is 191 deg.F under the conditions the 
applicant has analyzed in SAR Section 4.4.3. Because this temperature 
is significantly lower than the melting temperature, no softening or 
flow of lead in the annulus due to lead slumping is expected.
    Comment I.21: One commenter asked how the fuel debris could affect 
unloading if it clogs the drain tubes during reflooding and stated that 
this issue should be addressed along with the operating procedures to 
transfer a loaded cask.
    Response: Fuel debris is defined in the TS and is handled within 
individual fuel tubes in an 8 x 8 array within an RFA. The fuel tubes 
and RFA are designed to preclude the release of gross particles to the 
canister. Similar radiological precautions would need to be taken by 
the cask user for both the loading and the unloading evolution when 
handling fuel debris. The technical basis for the development of site-
specific operating procedures for transferring a loaded canister to the 
NAC-STC for transport have been approved for Certificate of Compliance 
No. 71-9235.

Summary of Final Revisions

    As a result of the staff's response to public comments, or to 
rectify issues identified during the comment period, the following 
items in the TSs have been modified: (1) TS Design Feature Section 
4.4.6 (see comments H.2, H.3, H.4 and H.5); (2) TS LCO 3.1.5 (see 
comments H.7 and H.10); (3) TS LCO 3.1.6 (see comments H.9 and H.11); 
and (4) TS Table 3-1, Canister Limits (see comment H.13). In addition 
TS LCO 3.1.1 was deleted (see comments G.3, G.7 and H.6) and TS 3.1.10 
was added (see comment G.5). The staff has also updated the CoC, 
including the addition of explicit conditions governing acceptance 
tests and maintenance program, approved contents, and design features, 
and has removed the bases section from the TSs attached to the CoC to 
ensure consistency with NRC's format and content.
    The title of the SAR has been revised to delete the revision number 
so that in the final rule the title of the SAR is ``Final Safety 
Analysis Report for the NAC Multi-Purpose Canister (NAC-MPC) System.'' 
The staff has also modified the rule language by changing the word 
``certification'' to ``certificate'.

Agreement State Compatibility

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

Finding of No Significant Environmental Impact: Availability

    Under the National Environmental Policy Act of 1969, as amended, 
and the Commission's regulations in Subpart A of 10 CFR Part 51, the 
NRC has determined that this rule is not a major Federal action 
significantly affecting the quality of the human environment and 
therefore, an environmental impact statement is not required. This 
final rule 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 Merri Horn, 
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone (301) 415-8126, 
e-mail [email protected].

Paperwork Reduction Act Statement

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

Public Protection Notification

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

Voluntary Consensus Standards

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

Regulatory Analysis

    On July 18, 1990 (55 FR 29181), the Commission issued an amendment 
to 10 CFR Part 72. The amendment provided for the storage of spent 
nuclear fuel in cask systems with designs approved by the NRC under a 
general license. Any nuclear power reactor licensee can use cask 
systems with designs approved by the NRC to store spent nuclear fuel if 
it notifies the NRC in advance, the spent fuel is stored under the 
conditions specified in the cask's CoC, and the conditions of the 
general license are met. In that rule, four spent fuel storage casks 
were approved for use at reactor sites and were listed in 10 CFR 
72.214. That rule envisioned that storage casks certified in the future 
could be routinely added to the listing in 10 CFR 72.214 through the 
rulemaking process. Procedures and criteria for obtaining NRC approval 
of new spent fuel storage cask designs were provided in 10 CFR Part 72, 
Subpart L.
    The alternative to this action is to withhold approval of this new 
design and issue a site-specific license to each utility that proposes 
to use the casks. This alternative would cost both the NRC and 
utilities more time and money for each site-specific license. 
Conducting site-specific reviews would ignore the procedures and 
criteria currently in place for the addition of new cask designs that 
can be used under a general license, and would be in conflict with NWPA 
direction to the Commission to approve technologies for the use of 
spent fuel storage at the sites of civilian nuclear power reactors

[[Page 12460]]

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

Small Business Regulatory Enforcement Fairness Act

    In accordance with the Small Business Regulatory Enforcement 
Fairness Act of 1996, the NRC has determined that this action is not a 
major rule and has verified this determination with the Office of 
Information and Regulatory Affairs, Office of Management and Budget.

Regulatory Flexibility Certification

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

Backfit Analysis

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

List of Subjects 10 CFR Part 72

    Criminal penalties, 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. 553; the NRC is proposing to 
adopt 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

    a. 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. 10d-
48b, sec. 7902, 10b Stat. 31b3 (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).


    b. In Sec. 72.214, Certificate of Compliance 1025 is added to read 
as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1025.
    SAR Submitted by: NAC International.
    SAR Title: Final Safety Analysis Report for the NAC Multi-Purpose 
Canister System (NAC-MPC System).
    Docket Number: 72-1025
    Certificate Expiration Date: April 10, 2020.
    Model Number: NAC-MPC.

    Dated at Rockville, Maryland, this 24th day of February, 2000.

    For the Nuclear Regulatory Commission.
Carl J. Paperiello,
Acting Executive Director for Operations.
[FR Doc. 00-5588 Filed 3-8-00; 8:45 am]
BILLING CODE 7590-01-P