[Federal Register Volume 65, Number 203 (Thursday, October 19, 2000)]
[Rules and Regulations]
[Pages 62581-62599]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-26888]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG32


List of Approved Spent Fuel Storage Casks: NAC-UMS 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 Universal Storage System (NAC-UMS) 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 November 20, 2000.

FOR FURTHER INFORMATION CONTACT: Stan Turel, telephone (301) 415-6234, 
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 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-UMS cask system to the list of approved 
spent fuel storage casks 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 UMS 
Universal Storage System.'' The NRC evaluated the NAC submittal and 
issued a preliminary Safety Evaluation Report (SER) and a proposed 
Certificate of Compliance (CoC) for the NAC-UMS cask system. The NRC 
published a proposed rule in the Federal Register (64 FR 45918; August 
23, 1999) to add the NAC-UMS cask system to the listing in 10 CFR 
72.214. The comment period ended on April 5, 2000. Seven comment 
letters were received on the proposed rule.
    Based on NRC review and analysis of public comments, the NRC has 
modified, as appropriate, its proposed CoC and the Technical 
Specifications (TS) for the NAC-UMS cask system. The NRC has also 
modified its SER in response to some of the comments.
    The NRC finds that the NAC-UMS 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-UMS 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-UMS 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 
November 20, 2000. Single copies of the final CoC and SER will be 
available by November 2, 2000 for public inspection and/or copying for 
a fee at the NRC Public Document Room (PDR), 11555 Rockville Pike, 
Rockville, Maryland and electronically at http://ruleforum.llnl.gov.
    Documents created or received at the NRC after November 1, 1999, 
are also available electronically at the NRC's Public Electronic 
Reading Room on the Internet at http://www.nrc.gov/NRC/ADAMS/index.html. The public can gain entry from this site into the NRC's 
Agency wide Document Access and Management System (ADAMS), which 
provides text and image files of the NRC's public documents. An 
electronic copy of the final CoC, Technical Specifications, and SER for 
the NAC-UMS cask system can be found in ADAMS under Accession No. 
ML003737374. However, because the NRC must incorporate the date of 
publication of this Federal Register notice into the CoC, these 
documents are not yet publicly available. The NRC will make these 
documents publically available by November 2, 2000. Contact the NRC PDR 
reference staff for more information. PDR reference staff may be 
reached at 1-800-397-4209, 301-415-4737, or by e-mail at [email protected].

Summary of Public Comments on the Proposed Rule

    The NRC received seven comment letters on the proposed rule. The 
commenters included two utilities, an NAC-UMS cask users group, two 
States, and two members of the public. Copies of the public comments 
are available for review in the NRC Public Document Room, 11555 
Rockville Pike, Rockville, MD and electronically at http://ruleforum.llnl.gov.

Comments on the NAC-UMS Cask System

    The comments and responses have been grouped into nine subject 
areas: general, radiation protection, accident analysis, design, welds, 
structural, thermal, technical specifications (TS), and miscellaneous 
issues. Several of the commenters provided specific comments on the 
draft CoC, NRC's preliminary SER, and TSs. To the extent possible, all 
of the comments on a particular subject are grouped together. The NRC's 
decision to list the NAC-UMS 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's 
responses follow:

A. General

    Comment A-1: One commenter noted the regulatory analysis indicates 
that issuing a site-specific license would cost the NRC and the utility 
more time and money than the proposed action. The commenter asked for 
proof of this statement and suggested that a study or evaluation should 
be done. The commenter considers that in the long run it costs the NRC 
more time and

[[Page 62582]]

money to make all the site-specific changes needed later. Further, if 
each cask were site-specific, the vendor and utility would pay for a 
thorough analysis before presentation to the NRC, rather than the NRC 
``fixing up'' everything at taxpayer expense after certification for a 
general license.
    Response: The NRC disagrees with the comment. The scope of an NRC 
review of a cask design to be added under the listing of 10 CFR 72.214 
is enveloped by the NRC review efforts to license that same cask design 
for a site-specific license. The NRC's review of that same cask design 
for a site-specific license also includes, but is not limited to, 
evaluations of siting factors, licensee financial qualifications, 
physical protection provisions, emergency plan provisions, the quality 
assurance program and the decommissioning plan. Clearly, and as stated 
in the regulatory analysis, the NRC and licensee costs would increase 
to conduct multiple site-specific reviews associated with the use of 
the same cask design.
    Conducting site-specific reviews would ignore the alternative 
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 the NWPA direction to the NRC to approve technologies for 
the use of spent fuel storage at the sites of civilian nuclear power 
reactors without, to the maximum extent practicable, the need for 
additional site reviews. It 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. Also, 
because of the long experience with the CoC process and other similar 
processes the NRC has determined that site-specific licensing would be 
inefficient because of the significant number of amendments that would 
have to be processed and therefore would add to the costs of granting 
CoCs rather than being more efficient.
    Prior to storing spent fuel under the general license, each 
licensee must perform written evaluations to establish that: (1) The 
conditions set forth in the CoC have been met; (2) the reactor site 
parameters are encompassed by the cask design bases considered in the 
cask SAR and SER; and (3) other requirements detailed in 10 CFR 72.212 
have also been met. Each general licensee must retain a copy of these 
written evaluations until spent fuel is no longer stored under the 
general license. Furthermore, these written evaluations may be 
inspected at any time by NRC staff.
    The NRC's fee recovery structure in 10 CFR Parts 170 and 171 for 
the conduct of licensing and regulatory oversight activities under 10 
CFR Part 72 does not differentiate between the type of license used 
(i.e., general or specific).
    Comment A-2: One commenter commented that the proposed rules for 
casks and the environmental assessments have become almost a ``fill in 
the blank'' form, and said that this needs rethinking. The commenter 
also made several general statements about the overall waste program 
and that everything is going too fast, spent fuel pools are filling to 
capacity, more cask designs being built by more inexperienced workers 
with the cheapest materials. The commenter suggested that the NRC 
examine the program and carefully evaluate the end result.
    Response: This comment is beyond the scope of this rule that is 
focused solely on whether to add a particular cask design, the NAC-UMS 
cask system, to the list of approved casks. However, since the 
beginning of the CoC rulemaking process, the NRC and Congress have 
continuously evaluated the direction and progress of the program with 
the primary consideration continuing to be the health and safety of the 
public.
    Comment A-3: One commenter cited a news article stating that one 
utility is seeking an accelerated licensing review and approval 
schedule for storage of fuel in the NAC-UMS, and was concerned that 
there may be pressure because of the schedule. The commenter asked how 
much public comment is valued when the public knows the approval needs 
to be completed as fast as possible. The commenter stated that NRC's 
job is to ensure public and worker safety.
    Response: This comment is beyond the scope of this rule that is 
focused solely on whether to add a particular cask design, the NAC-UMS 
cask system, to the list of approved casks. However, since the 
beginning of the CoC rulemaking process, the NRC and Congress have 
continuously evaluated the direction, progress, and schedules of the 
program with the primary consideration continuing to be the health and 
safety of the public. The public comment and response procedure has 
always been and will continue to be an important part of the rulemaking 
process.
    Comment A-4: One commenter did not receive the reference section as 
listed in the Table of Contents for the SER and asked why. The 
commenter stated that the references and dates are important and that 
the public wants these references and dates. However, the references 
are often dated from the 1970's causing concern to the commenter. The 
commenter requested the missing pages from the SER.
    Response: The NRC separately provided the reference section of the 
SER issued with the preliminary SER to the commenter. The NRC had 
appropriately included the dates of references in the preliminary SER, 
and is uncertain why the commenter did not receive this section.
    Comment A-5: One commenter noted differences between NAC-MPC and 
NAC-UMS and stated that the terms ``multipurpose'' and ``universal'' 
are not explained. The commenter stated the casks are for storage only 
at this point and that is what they should be called in the documents.
    Response: Similarities or differences between the NAC-UMS cask 
design under consideration and any other cask design are beyond the 
scope of this rulemaking. The terms ``universal'' and ``multi-purpose'' 
have been selected by the applicant as descriptive of the system's 
design flexibility. The NRC agrees with the commenter that the NAC-UMS 
cask design evaluated in this rulemaking is limited to its 
acceptability for storage. However, the NRC does not consider 
descriptive nomenclature of the intended use beyond storage to be 
inappropriate.
    Comment A-6: One commenter asked what the ``M'' in UMS stands for 
and why is it not USS for Universal Storage System.
    Response: The NAC-UMS is the model name selected by the vendor. UMS 
stands for ``Universal MPC System,'' where MPC is intended to indicate 
``multi-purpose canister.''
    Comment A-7: One commenter agreed with one of the State's published 
comments. Several comments also were made on topics pertaining to the 
decommissioning of the Maine Yankee site.
    Response: The agreement with the State's published comments was 
noted. The State's comments in their entirety have been considered 
within this section. The comments pertaining to the decommissioning of 
the Maine Yankee site are outside the scope of this rule.

B. Radiation Protection

    Comment B-1: One commenter disagreed with the SER statement that it 
is unnecessary for the applicant to specify the source term for the 
confinement analyses and stated that the source term and corresponding 
dose consequence should be provided to the public in these documents. 
The

[[Page 62583]]

commenter stated there is no reason not to require this information 
that the NRC may need to know in the future.
    Response: The NRC disagrees with the comment. Revision 1 of Interim 
Staff Guidance (ISG) No. 5, ``Confinement Evaluation'' specifies that 
for storage casks having closure lids that are designed and tested to 
be leak tight as defined in ``American National Standard for Leakage 
Tests on Packages for Shipment of Radioactive Materials,'' American 
National Standards Institute (ANSI) N14.5-1997, detailed confinement 
analyses are not necessary. Therefore, the applicant is not required to 
provide a detailed analysis of the leakage of radioactive materials 
through the welded canister. As indicated in SAR Section 7.1, the 
confinement boundary is completely welded and inspected in accordance 
with both the American Society of Mechanical Engineers (ASME) Code and 
ISG No. 4, ``Cask Closure Weld Inspections,'' and is leak tested to 
ANSI leaktight standards. Further, the analyses presented in the SAR 
demonstrated that the stresses, temperatures, and pressures of the 
Transportable Storage Canister (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 for all credible 
accidents. The NRC concurs with the evaluation in the SAR and believes 
that the design of the confinement boundary, that includes the 
inspection of welds is adequately rigorous and meets the applicable 
regulations.
    Comment B-2: One commenter asked if there is an explanation in the 
SAR of detailed plans for how to dispose of the radioactive gases 
purged from the canister with nitrogen during unloading. The commenter 
asked if the disposal process has been clearly thought out so it could 
be performed the day after a cask is loaded, if necessary, and all 
personnel would know the process.
    Response: SAR Chapter 8 includes guidance for the development of 
site-specific operating procedures to be followed for unloading the TSC 
and includes consideration of the radioactive gases purged from the 
canister. 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 radioactive gaseous effluent released from the canister 
would be processed through High-Efficiency Particulate Air (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 radioactive effluents released to the environment must 
meet Federal and State regulations.
    Comment B-3: One commenter asked if the high peak dose rates could 
be reduced in some way for the transfer cask top during shield lid 
welding, the top of the transfer cask containing a sealed canister 
filled with Boiling Water Reactor (BWR) fuel, and the bottom of the 
transfer cask with a canister filled with Pressurized Water Reactor 
(PWR) fuel.
    Response: The high peak dose rates are based upon loading the 
design basis fuel and present the worst case scenario for estimating 
doses to workers. The actual doses received by workers should be less 
than the calculated doses because the actual fuel loaded may have a 
longer cooling time and a different, lower burnup. Under the facility's 
as low as reasonably achievable (ALARA) does exposure program, the 
licensee will have to evaluate ways to reduce the dose to those who 
will be working with the cask. For example, temporary shielding could 
be used to reduce dose to workers.
    Comment B-4: Three commenters noted that the Completion Time for 
Required TS Action A.1 of Limiting Condition for Operation (LCO) 3.2.1 
(Decontamination of Canister Surface Contamination) is unnecessarily 
restrictive. The commenters request that the Completion Time be revised 
to 25 days because this LCO is not time dependent.
    Response: The NRC disagrees with this comment. The applicant 
evaluated and proposed the 7-day time frame. During the review process, 
the staff evaluated and found acceptable the applicant's proposal. The 
NRC found the 7-day completion time reasonable to decontaminate the 
surface if contamination on the canister or transfer cask is 
identified. The commenters did not provide adequate justification for 
revising the LCO. If there is surface contamination on the canister or 
transfer cask, then it is good health physics practice to decontaminate 
the surface as soon as practicable but within the seven day completion 
time.
    Comment B-5: Three commenters stated that the Completion Time for 
Required Action A.2 of LCO 3.2.2 (Concrete Cask Average Surface Dose 
Rates) is unnecessarily restrictive, and request that the Completion 
Time be revised to 25 days.
    Response: The NRC disagrees with this comment. The applicant 
evaluated and proposed the 7-day time frame. During the review process, 
the NRC evaluated and found acceptable the applicant's proposal. The 
NRC found the 7-day completion time reasonable to verify compliance 
with the regulations. The comment did not provide adequate 
justification for revising the LCO.
    Comment B-6: Two commenters noted that the radiological dose to 
adjacent controlled or noncontrolled site areas is based on 20 loaded 
vertical storage modules (Preliminary Safety Evaluation Report [PSER] 
Sections 10.3 and 10.4), and that the prototypical modules are arranged 
in two rows with ten storage modules per row. The commenters stated 
this assumption is unrealistic in Independent Spent Fuel Storage 
Installations (ISFSIs) that support the complete decommissioning of an 
operating nuclear power plant where there may be 50 or more modules. 
The more storage modules, the greater the sky shine interaction that is 
available at the boundary of the site control area and the greater the 
onsite occupational dose. The commenters stated that the PSER does not 
analyze the more typical module configurations and, thus, does not meet 
the requirements of 10 CFR 72.236(d).
    Response: NRC disagrees with this comment. This application is for 
a general license and therefore a generic approach has been taken in 
evaluating the doses to site workers and the public. Prior to a general 
licensee using this cask, the licensee is required to meet the 
conditions stated in 10 CFR 72.212. Specifically, 10 CFR 
72.212(b)(2)(iii) states that the requirements in 10 CFR 72.104 (the 
criteria for radioactive materials in effluents and direct radiation 
from an ISFSI or Monitored Retrievable Storage Facility (MRS)) must be 
met. Therefore, to demonstrate compliance with 10 CFR 72.104, the 
Sec. 72.212 evaluation will have to contain a dose evaluation for the 
ISFSI site that includes the actual number and arrangement of storage 
canisters.
    Comment B-7: One commenter stated that compliance with required 
actions A.1 and A.2 for LCO 3.2.2 in the TS does not either restore 
compliance with the LCO or allow exiting the LCO. LCO 3.2.2 in the TS 
contains limits for the average surface dose rates of each concrete 
cask during loading operations. Surveillance requirement (SR) 3.2.2.1 
requires that the average surface dose rates be measured once after 
completion of transfer of a loaded canister into the concrete cask and 
before beginning storage operations. Condition A and required actions 
A.1 and A.2 for this LCO state that if the concrete average surface 
dose rate limits are not met, the

[[Page 62584]]

licensee must administratively verify correct fuel loading, and perform 
analysis to verify compliance with the ISFSI offsite radiation 
protection requirements of 10 CFR Parts 20 and 72. However, there is no 
provision in this LCO to allow the loaded concrete cask to be stored in 
the ISFSI after actions A.1 and A.2 are completed satisfactorily. The 
LCO does not provide for any course of action after actions A.1 and A.2 
are completed. SER Sections 5.4.3 and F5.3 state that the final 
determination of compliance with 10 CFR 72.104(a) is the responsibility 
of each applicant for a site license. Section 10.1.1 states that, as 
required by 10 CFR 72.212, a general licensee will be responsible for 
demonstrating site-specific compliance with 10 CFR part 20 and 
Secs. 72.104 and 72.106 requirements. The intent of LCO 3.2.2 is that a 
licensee may store a cask that does not meet the LCO average surface 
dose rate limits as long as the licensee completes an analysis showing 
compliance with 10 CFR parts 20 and 72 limits at the ISFSI. Therefore, 
in order for required actions A.1 and A.2 to restore compliance with 
the LCO, the LCO should state: ``The average surface dose rates of each 
Concrete Cask shall not exceed the following limits unless required 
actions A.1 and A.2 are met.''
    Response: NRC agrees with this comment. LCO 3.2.2 has been revised.
    Comment B-8: One commenter asked why there is axial reflection of 
neutrons from one tube to another bypassing the poison panels under 
full or partial flooding, and how this affects analysis. The commenter 
stated that if the NRC does not support NAC's claim that the infinite-
length approximation adds conservatism, it should be removed.
    Response: Although the NRC does not concur with NAC's statement 
that the infinite-length model adds conservatism, removal of the 
statement from the SAR is not necessary because the statement does not 
affect the overall conclusions of the safety analysis. The axial 
reflection of neutrons from one tube to another occurs when neutrons 
leaving the end of one fuel tube are scattered into another fuel tube 
by water, fuel hardware, or cask materials located beyond the ends of 
the poison panels. This phenomenon provides a neutron pathway between 
assemblies that is not considered in infinite-length models of the fuel 
and cask. The NRC's analysis shows that the resulting small increase in 
the computed reactivity roughly balances the small reactivity decrease 
arising from axial neutron leakage, which is likewise neglected in 
NAC's infinite-length model. The NRC therefore views the infinite-
length approximation as neutral; i.e., it neither adds nor subtracts 
conservatism.

C. Accident Analysis

    Comment C-1: One commenter noted that the thermal accident is 
postulated with 50 gallons of transporter fuel burning for 8 minutes 
and suggested that an evaluation for a possible jet crash and 
associated fire be performed.
    Response: The NRC staff's standard review plan for dry cask storage 
systems, Chapter 11 ``Accident Analysis,'' specifies that structures, 
systems, and components important to safety must be designed to 
withstand credible accidents and natural phenomena events. A cask 
transporter fire is considered credible for the NAC-UMS cask design, 
and is the basis for the 8-minute fire associated with the time it 
would take to burn 50 gallons of fuel. Other modes of transport causing 
the fire (such as airplanes, trains, and delivery trucks) are not 
considered plausible for this cask design and are beyond the scope of 
this rule. However, before using the NAC-UMS cask, the general licensee 
must evaluate the site to determine if the chosen site parameters are 
enveloped by the design bases of the approved cask as required by 10 
CFR 72.212(b)(3). The licensee's site evaluation should consider the 
effects of nearby transportation and military activities. Also included 
in this evaluation is the verification that the cask handling equipment 
used to move the Vertical Concrete Cask (VCC) to the pad is limited to 
50 gallons of fuel (as detailed in Technical Specification B 3.4.5-Site 
Specific Parameters and Analyses).
    Comment C-2: Three commenters requested that LCO 3.1.7 (Fuel 
Cooldown Requirement) be deleted from the TS because there are no 
design basis accidents that require fuel cooldown for removal from a 
sealed canister. The commenters believed that the applicant 
demonstrated that cooldown can be performed as shown by the ``Thermal 
Evaluation'' section of NUREG-1536, ``The Standard Review Plan for Dry 
Cask Storage Systems, January, 1997'' and that if the fuel cooldown 
requirements cannot be removed from the TS, the cooldown requirements 
should be moved to the ``Administrative Controls and Programs'' 
section.
    Response: The NRC agrees with the comment that the TS A 3.1.7, 
``Fuel Cooldown Requirements'' associated with canister unloading 
procedures can be deleted from the TS. The NRC agrees that this would 
be a highly unlikely scenario that could be adequately controlled by 
approved site-specific operating procedures developed based on the 
technical basis contained in SAR Chapter 8. Reuse of the canister after 
unloading would not be likely. The fuel would be returned to the spent 
fuel pool for subsequent dry cask storage in another canister and/or 
transport.
    Comment C-3: One commenter asked a number of questions related to 
the Boral panels regarding the continued efficiency over time, the 
number of casks that have utilized Boral, how the Boral is manufactured 
and tested, and whether the panels can structurally deform.
    Response: Boral has been used in the nuclear industry since the 
1950's and has been used in spent fuel storage and transportation cask 
baskets since the 1960's. Several utilities have also used Boral in 
spent fuel pool storage racks. Industry experience has revealed no 
credible mechanisms for a loss of Boral efficacy in the cask. 
Therefore, the NRC has reasonable assurance that the Boral panels in 
the PWR and BWR baskets of the TSC will perform their intended 
criticality function throughout the licensed storage period.
    Each Boral 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. 
As noted in SAR Section 6.1, criticality control in the PWR basket is 
achieved by surrounding the fuel assemblies with four panels of Boral 
for each fuel assembly. In the BWR basket, single panels of Boral 
placed between each fuel assembly are used for criticality control.
    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., 0.025 grams of B\10\ per cm\2\ for the PWR basket 
and 0.011 grams of B10 per cm2 for the BWR basket) is ensured by the 
acceptance testing procedures described in SAR Section 9.1.6.
    Comment C-4: One commenter noted that the NRC had reviewed the 
Boral vendor's product literature and believed this should be done for 
all materials because most cask vendors do not review this information. 
The commenter stated that nonstandard Boral sheets, are an area where 
mistakes may be made and verifications are not performed. The commenter 
asked why NAC was not ``up front'' with the issue of using nonstandard 
Boral sheets.

[[Page 62585]]

    Response: The NRC disagrees that most vendors do not review 
material specifications selected for use within cask designs. The 
vendor is responsible for implementing a quality assurance program. The 
NRC expects that the material used in the cask systems meets minimum 
design specifications. The NRC has no specific information that this or 
other vendors do not properly specify and confirm material properties. 
Furthermore, the NRC does specifically evaluate and consider the 
materials utilized in a proposed cask design. Regarding the use of 
``non-standard'' Boral sheets, the vendor had already committed to 
obtaining a specific B\10\ loading for the neutron absorbers, both in 
the SAR and as stipulated in the design features section of the TS. The 
NRC's safety evaluation fully describes the basis for the NRC's 
acceptance.
    Comment C-5: One commenter expressed a concern about the possible 
production of hydrogen from the aluminum heat transfer disks during 
loading and unloading operations.
    Response: The NRC has considered the possible production of 
hydrogen in its evaluation. As noted in SAR Section 3.4.1.2.2, the 
applicant anticipates that no hydrogen gas is expected to be detected 
prior to, or during, the loading or unloading operations. However, if a 
reaction between the aluminum heat transfer disks and the spent fuel 
pool water occurs, the loading and unloading procedures of SAR Chapter 
8 that 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.
    Comment C-6: Two commenters stated that the NAC-UMS system does not 
provide for a capability to verify periodically whether or not the 
storage conditions have changed, thus requiring canning or other 
remedial measures for fuel that has developed further damage during 
storage. The commenters stated that the fuel-containing canisters may 
need to be opened periodically in a hot cell and visually inspected, 
and that an ISFSI using the NAC-UMS system may require such a facility 
because the canisters may not be shipped under 10 CFR Part 71 without 
verification of fuel rod integrity. The commenters stated that the PSER 
should define verification requirements for the NAC-UMS system prior to 
shipment under Part 71 and evaluate the applicant's verification 
methods.
    Response: The NRC disagrees with the comment. The NRC, with the 
issuance of Interim Staff Guidance (ISG) No. 1, ``Damaged Fuel'' 
addressed the definition of damaged fuel and clarified the fuel 
conditions for which spent fuel should be placed in cans prior to 
storage for the purposes of retrievability. The NAC-UMS storage cask 
application, as considered in this rulemaking, did not seek approval 
for the storage of damaged fuel as defined in ISG-1. Additionally, both 
the design of the NAC-UMS system and the thermal, structural, and 
criticality analyses ensure that the fuel will not be disrupted under 
normal, off-normal and accident conditions once undamaged, or intact, 
fuel is placed into a storage canister. Further, the results of a cask 
demonstration program at Idaho National Engineering and Environmental 
Laboratory (INEEL) (where determinations were made of the effects of 
dry storage casks on spent fuel integrity) showed that there were no 
significant fuel failures that would require extraordinary handling of 
the fuel. Therefore, the NRC staff has reasonable assurance that the 
spent fuel is adequately protected against degradation that might 
otherwise lead to gross rupture during storage. As such, periodic 
verification of cladding conditions during the storage period or prior 
to transportation is not warranted.
    Regarding requirements associated with the safe transportation of 
spent fuel under 10 CFR Part 71, it is appropriate to establish the 
necessary conditions that ensure the health and safety of the public 
under the conditions of the 10 CFR Part 71 CoC. A 10 CFR Part 72 
storage cask design certification does not serve to authorize the 
shipment of the stored contents under 10 CFR Part 71. NRC does an 
independent evaluation of casks for shipping under 10 CFR Part 71. 
Similarly, conditions of any approval under 10 CFR Part 71 are 
independent of necessary conclusions pertaining to a cask design's 
capability to meet the requirements of 10 CFR Part 72 for storage.
    Comment C-7: Two commenters raised concerns about the radiation 
hardening of borated neutron absorber materials, including the NS-4-FR 
neutron shield employed in the NAC-UMS storage cask. The commenters 
stated there is no evidence and no analysis in the PSER to establish 
NS-4-FR's ability to maintain form over the expected lifetime 
integrated neutron flux.
    Response: The NRC has reasonable assurance that NS-4-FR will 
maintain its form over the expected lifetime integrated gamma and 
neutron doses. Independent laboratory tests of the NS-4-FR material 
have demonstrated that radiation exposures significantly higher than 
those of any neutron shield component of the NAC-UMS system have not 
resulted in any physical deterioration of the neutron shield material. 
Calculations have shown that over 500 continuous years of exposure to a 
design basis neutron source would have to occur before the transfer 
cask shield neutron exposure would reach the level of the laboratory 
tests. Similarly, over 50 years of continuous design basis gamma 
exposure would be required before the laboratory test exposure levels 
were reached. In actuality, the exposures would need to be considerably 
longer with spent fuel due to the continually declining source term.
    The NS-4-FR neutron shield material is used as a neutron shield in 
the transfer cask and the Vertical Concrete Cask (VCC) shield plug. It 
is not used in the storage cask. In the transfer cask, the amount of 
time this material will experience significant neutron fluxes is 
minuscule compared to the amount of time to cause radiation 
embrittlement of the material. In the VCC shield plug, the NS-4-FR 
material is placed above the canister lid and is exposed to 
significantly lower neutron fluxes than seen by the transfer cask.
    Further, for both the transfer cask and the VCC shield plug, the 
NS-4-FR neutron shield is completely enclosed within welded steel 
components. In the transfer cask, the top and bottom plates are seam 
welded to the shell with full penetration or fillet welds to enclose 
the NS-4-FR material. Similarly, the NS-4-FR in the VCC shield plug, is 
enclosed between the shield plug, a retaining ring and a cover plate 
using fillet welds. Since the NS-4-FR is sandwiched between these 
various steel shells for the transfer cask and VCC shield plug, the NRC 
has reasonable assurance that the NS-4-FR material will maintain its 
form over the expected lifetime of the transfer cask's or shield plugs 
radiation exposures. Even if the material were to become embrittled, 
its placement within the VCC shield plug and transfer cask components 
would not allow the material to redistribute.
    Comment C-8: One commenter stated that eight supply and two 
discharge lines in the transfer cask wall adds to confusion and 
mistakes, and that introducing forced air to cool the contents and 
allow the canister to remain longer in the transfer cask is asking for 
trouble because workers bank on the time being available.
    Response: The NRC disagrees with the comment. The number of supply 
and discharge lines is a specific design

[[Page 62586]]

objective to ensure uniform cooling so the spent fuel contents in the 
canister remain within the design envelope during loading and unloading 
operations. Activities associated with the safe and proper use of the 
transfer cask design are to be conducted in accordance with site-
specific operating procedures generated by the user. Appropriate 
identification and controls for the operation of the air supply and 
discharge lines, sufficient to minimize confusion and mistakes, are a 
responsibility of the general licensee. The objective of the option to 
provide forced air cooling to the transfer cask, although not intended 
to be routine, is to maintain the spent fuel contents within the design 
envelope at all times. If an operational situation results in the use 
of the forced air option, the spent fuel contents will remain under 
analyzed conditions, and thus the availability of this option is 
considered beneficial.
    Comment C-9: One commenter opposed the idea of using the transfer 
cask if a canister must be removed from a concrete cask. The commenter 
asked if the intent is to use the transfer cask for storage if there 
are problems and why.
    Response: The NRC evaluated and accepted the use of the transfer 
cask if a canister must be removed from a concrete cask, including 
unloading operations. The transfer cask is not an authorized 
configuration for long-term storage. The use of the transfer cask for 
loading and unloading operations is controlled by the TSs.
    Comment C-10: One commenter asked that preferential loading and 
administrative control of fuel assemblies not be allowed to leave a 
wide safety margin to protect the public.
    Response: The NRC disagrees with the comment. The NRC's safety 
evaluation determines with reasonable assurance that an adequate 
(rather than ``wide'') safety margin is ensured with respect to all 
cask activities. The proper selection and loading of candidate spent 
fuel assemblies necessarily relies on appropriate administrative 
controls. All 10 CFR Part 50 licensees that will use this cask design 
under the general license have extensive experience in selecting 
uniquely identified fuel assemblies for placement in uniquely 
identified locations, such as the reactor core or the spent fuel pool. 
Preferential loading specifications, in conjunction with the 
appropriate administrative loading controls, have been accepted by the 
NRC because they maintain an adequate safety margin and rely on similar 
existing administrative controls for safe fuel handling.
    Comment C-11: Three commenters requested the removal of the 
inference in Chapter 10 of the SAR that a daily inspection of the VCC 
vents is an expected or routine activity. The commenters stated that 
identification of blocked VCC vents is accomplished by use of the 
temperature monitoring systems, and that physical inspection of the VCC 
vents, especially daily, results in unnecessary exposure and is not in 
keeping with preferred As Low As Reasonably Achievable (ALARA) 
practices.
    Response: NRC disagrees with this comment. The cask user is 
required to verify the operability of the heat removal system by 
monitoring temperature instrumentation daily, as specified in TS 
A.3.1.6. As stated in SAR section 1.2.1.5.9, the temperature monitoring 
system can be read at a display device located on the outside surface 
of the cask or at a remote readout location. A daily inspection of the 
VCC vents is included in Chapter 10 of the SAR as an expected routine 
operation in determining a conservative, estimated annual dose due to 
routine operations as per ALARA practices. Whether to use a temperature 
monitoring system with a display on the outside of the casks or to use 
remote readout instrumentation is left to the cask user's discretion.
    Comment C-12: Two commenters stated that the operator testing and 
training exercises described in CoC Section A5.0 do not require 
training in the importance of sequence, and commented that the CoC 
implies that training will be conducted solely on the activity basis, 
and thus, the planned training loses the importance of the various 
interface requirements between activities that follow each other. This 
omission permits operator mistakes at activity intersections and may 
contribute to missing parameter values or conditions that must be met 
for safe loading and transfer of the assembly canister from the spent 
fuel pool to the storage cask. The commenters stated that individual 
procedures should include stated preconditions that must be satisfied 
by the previous sequential procedure and are necessary for safely 
performing the subsequent activity, and that without these procedures, 
the application does not satisfy the requirements of 10 CFR 72.236(l).
    Response: The NRC disagrees with the comment. The Administrative 
Controls and Programs section of the TS stipulates that the training 
program for the NAC-UMS system must be developed under the general 
licensee's systematic approach to training (SAT). The training modules 
must include comprehensive instructions for the operation and 
maintenance of the NAC-UMS System. The TS provides a detailed listing 
of the preoperational tests and training exercises that must be 
performed prior to the first use of the system to load spent fuel 
assemblies. Although the TS specifically recognizes that dry runs may 
be performed in an alternate step sequence from the actual procedures, 
it is the general licensee's responsibility under the SAT to establish 
and execute an effective preoperational testing and training program. 
With respect to the contents of individual procedures, Condition No. 2 
of the CoC specifies that the user's written site-specific operating 
procedures must be consistent with the technical basis described in 
Chapter 8 of the SAR. The preparation of written site-specific 
operating procedures that contain adequate and appropriate initial 
conditions, prerequisites, and verifications, is not necessary prior to 
this rulemaking to add the NAC-UMS cask design to the list of approved 
storage cask designs of 10 CFR 72.214.
    Comment C-13: One commenter asked why the speed of a vertical 
tornado-driven missile is assumed to be only 70 percent of the speed of 
a horizontal missile.
    Response: The primary wind velocities associated with tornadoes are 
in the horizontal direction, and thus wind velocities in the vertical 
direction are considered to be less as stated in NRC review guidance. 
Specifically, the NUREG-0800, Section 3.5.1.4, review guidance 
describes the basis for the assumption that the maximum speed of a 
vertical tornado-driven missile, at 88.2 mph, is specified as 70 
percent of a horizontal missile, at 126 mph. This vertical speed is 
enveloped by the horizontal missile speed of 126 mph considered 
conservatively in the SAR evaluation of the 1\1/2\ inch-thick VCC 
closure plate, that can only be hit by a vertical missile. The SAR has 
satisfactorily demonstrated that the VCC closure plate is adequate to 
withstand local impingement of a tornado missile traveling at the 
higher horizontal speed, e.g., 126 mph.
    Comment C-14: One commenter remarked that the transfer cask gets 
highly irradiated and exposed to high temperatures and contamination 
through repeated use and asked what happens to the transfer cask over 
time, especially the welds. The commenter stated that the trunnion area 
welds need inspection over time for possible leakage of pool water 
inside the transfer cask walls. The commenter stated that transfer 
casks for all cask designs need specific criteria for examination

[[Page 62587]]

periodically and that maybe the transfer casks are too neglected in NRC 
thinking. The commenter also asked what happens if water gets inside 
the walls starting chemical reactions and adding unaccounted for weight 
in lifts, and what are the requirements for transfer cask testing or 
checking over time.
    Response: The NRC agrees that the transfer cask will be subject to 
hostile environmental conditions such as high radiation, temperature, 
and contamination through repeated use. In SAR Section 9, NAC has 
committed to a transfer cask maintenance program to inspect the 
transfer cask trunnions and shield door assemblies for gross damage and 
proper function for each use. Annually, the lifting trunnions, shield 
doors, and shield door rails must be either dye penetrant or magnetic 
particle examined. The SAR states that the examination method must be 
in accordance with Section V of the ASME Code and the acceptance 
criteria Section III, Section NF, NF-5350, or NF5340, as required by 
ANSI N14.6. Therefore, the transfer cask, including trunnion welds, is 
examined periodically to ensure that it will function as designed over 
its entire service life. This provides reasonable assurance, 
supplemented by inspections prior to use that water will not get inside 
the wall to result in potential chemical reactions or unaccounted 
weight in lifts.
    Comment C-15: One commenter stated that if berms or shield walls 
are to be used for radiological protection, an evaluation of tornado 
missiles that could be generated as a result of their constituent 
materials should be performed.
    Response: The NRC agrees with the comment. Use of berms or shield 
walls for radiological protection is a site-specific consideration that 
is to be evaluated by the general licensee under 10 CFR 72.212 to 
ensure that the reactor sites parameters, including analyses of tornado 
missiles that could be generated due to the material constituency of 
any berms or shield walls, are enveloped by the cask design bases.
    Comment C-16: One commenter stated that explosion needs more 
evaluation, noting that where there is hydrogen, there can be an 
explosion.
    Response: The NRC disagrees with the comment. The NRC staff has 
found reasonable assurance that the possible generation of hydrogen due 
to cask loading and unloading operations has been evaluated, and that 
adequate controls are in place to detect and take corrective actions if 
significant quantities of combustible gases are generated. SAR 
Subsection 11.2.5 (explosion accident analysis under storage 
conditions) evaluates the NAC-UMS system subject to an external 
pressure up to 22 psig, has been accepted by the NRC staff, and 
provides part of the technical basis for site parameters evaluations 
performed in accordance with 10 CFR 72.212(b)(3). Further evaluation of 
the possible effects of an explosion involving hydrogen or other 
combustible materials under storage conditions is site-specific and 
beyond the scope of this rulemaking.
    Comment C-17: Three commenters stated that the parameters provided 
in B 3.4(6) of the Approved Contents and Design Features in Appendix B 
of CoC 1015 are not relevant to the drop accident condition and are not 
relevant to the tip-over provided that the allowable seismic 
accelerations are not exceeded (i.e., the cask does not tip over). As a 
result, the commenters request that Item 6 be revised to read: ``In 
addition to the requirements of 10 CFR 72.212(b)(2)(ii), the seismic 
acceleration at the top surface of the ISFSI pad cannot exceed the 
value provided in B 3.4 (3).''
    Response: The NRC agrees in part with the comment in that the 
parameters are not relevant to the SAR Subsection 11.2.4.3 VCC 24-inch 
vertical drop accident. These parameters have been removed from the TS 
as suggested. However, the same set of site concrete pad and soil 
parameters relevant to the tip-over analysis is being summarized in SAR 
Subsection 11.2.12 to ensure that the bounding side drop decelerations 
determined for the NAC-UMS system are available for site specific 
application without the need for going through additional cask tip-over 
analysis.
    Comment C-18: Two commenters stated the heavy load lifting ability 
of the transfer and storage systems (described in PSER Section 3.2.3) 
appears to be inadequately supported and that the systems are not 
redundant for either attachment or lift capability, and therefore, do 
not satisfy the requirements for single failure of the lifting 
equipment. The commenters also stated that the transfer cask trunnions 
and storage cask lifting lugs are not redundant and do not satisfy the 
requirements for single failure or the requirements of 10 CFR 
72.236(h).
    Response: The NRC disagrees with the comment on the adequacy of SAR 
evaluation for heavy load lifting abilities of the VCC lifting lugs and 
transfer cask trunnions.
    As noted in SER Subsection 3.2.3.4, the SAR demonstrates structural 
acceptance of the VCC components for the top lift operation in 
accordance with ANSI N14.6. The basic design stress factors of 3 and 5 
against materials yield (Sy) and ultimate (Su) 
strengths, respectively, are met with the allowable stress the lesser 
of Sy/3 or Su/5. The commenters were correct that 
the VCC lifting lugs do not meet the single-failure-proof lifting 
provision because the lifting lugs provide a single-load path. However, 
the SAR Subsection 11.2.4 VCC drop analysis is consistent with the 
assumption of non-single failure proof lifting lugs. Also, the VCC lift 
lugs do not need to be single failure proof because of accident 
analysis and administrative controls. The applicant's evaluation of a 
possible 24-inch vertical drop (limited by controls to a lift height of 
24 inches or less) of the VCC was shown to have no significant 
radiological consequences, and has been accepted by the NRC staff.
    On transfer cask trunnions, SER Subsection 3.2.3.1 recognizes that, 
for a two-trunnion lifting configuration, the maximum trunnion bending 
stress corresponds to the stress design factors of 9.4 and 20.7 that 
are larger than the required factors of 6 and 10 against the material 
yield and ultimate strengths, respectively. Therefore, the structural 
capability of the trunnions satisfies the ANSI N14.6, Section 7.1, 
requirements for lifting critical loads with either a dual-load path 
handling system (with the basic design stress factors of 3 and 5 
against materials yield and ultimate strengths, respectively), or a 
single-load path system with increased design stress factors that 
double the basic design stress factors.
    Comment C-19: Two commenters stated that the criticality analysis 
as discussed in the PSER Section 6.4 does not provide a listing of the 
fissile material in the spent fuel assemblies, without which the 
analysis is questionable and does not satisfy the requirements of 10 
CFR 72.236(c). Of particular concern is the concentration of Pu-239 
which continues to undergo spontaneous fission and therefore, increased 
neutron flux.
    Response: The NRC disagrees with the comment. The criticality 
analysis uses the conservative assumption of fresh fuel without 
burnable poisons. The analyzed fresh-fuel composition is always more 
reactive than the actual composition of irradiated fuel. Consistent 
with the fresh-fuel assumption, the criticality analysis lists only the 
fissile materials present in fresh fuel. Results of the analysis 
clearly demonstrate compliance with 10 CFR 72.236(c), the requirement 
that the spent fuel be maintained in a subcritical condition. The NRC 
notes that the neutron flux arising from spontaneous

[[Page 62588]]

fission or other fixed neutron sources in the cask has no bearing on 
the neutron multiplication factor, keff. Furthermore, as 
shown in the shielding analysis, the neutron flux in stored spent fuel 
arises mainly from the spontaneous fission of Cm-242 and Cm-244. 
Spontaneous fission of Pu-239 contributes very little to the neutron 
flux in spent fuel.

D. Design

    Comment D-1: One commenter expressed concern about icicles forming 
and covering the cask vent holes. The commenter stated that more study 
is needed for full cask array monitoring and cleaning in an ice storm, 
and that plans should be made for this situation.
    Response: TS A.3.1.6, ``Concrete Cask Heat Removal System'' 
requires that the cask user perform daily surveillance to verify the 
cask outlet temperature. The method of performing the daily check is a 
site-specific consideration of the cask user. If the daily temperature 
surveillance indicates a temperature outside of the acceptable range, 
then an inspection must be performed within 4 hours to verify that the 
inlets and outlets are not blocked or obstructed.
    Comment D-2: One commenter did not share the NRC's reasonable 
assurance that cladding will be protected in unloading because it has 
never really been tried and tested. The commenter stated that this 
testing needs to be performed on cladding material and that the 
commenter has been requesting the NRC to prove the cladding integrity 
for years.
    Response: The NRC disagrees with the comment. The NAC-UMS 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 conclusions that are based on sound 
engineering methods and practices. The NRC has reviewed the analyses 
performed by NAC and found them acceptable. However, as part of an 
ongoing cooperative research effort (NRC, DOE, and EPRI) regarding 
long-term performance of spent fuel storage, one spent fuel storage 
cask has been unloaded and inspected at INEEL in Idaho. Results to date 
are quite reassuring that the behavior of the casks and fuel assemblies 
is as expected.
    Comment D-3: One commenter asked what is the purpose of adding 
solar heat to the outer cask surface and averaging over a 12-hour 
period for the air flow and concrete cask model. The commenter also 
stated that reducing the view factor when analyzing thermal interaction 
among casks in an array, as was done for this design, should be done 
for all cask designs.
    Response: The purpose of adding insolation to the air flow and 
concrete cask model is to include the effect of solar heat on the cask 
that would heat the outer surface of the concrete cask and reduce heat 
removal from the canister through the concrete. The amount of solar 
heat is determined from 10 CFR Part 71 and may be averaged over a 24-
hour period per the guidance provided in NUREG-1536, the Standard 
Review Plan for Dry Cask Storage Systems. The comment that other cask 
designs should similarly reduce the view factor to compensate for an 
array arrangement is outside the scope of this NAC-UMS rule.
    Comment D-4: Three commenters requested that the language in B 
2.1.2 of the ``Approved Contents and Design Features'' addressing 
preferential loading and center position loading of shortest cooled 
fuel be revised as follows:
     The last two sentences of the first paragraph of this 
section should be deleted.
     The second paragraph should be revised to delete reference 
to the ``basket interior,'' which is described as the ``basket center 
positions'' in the previous paragraph.
     The third paragraph should be moved prior to the current 
first paragraph.
     The first sentence of the current second paragraph should 
be made a separate paragraph, as it is not related to the text that 
follows.
    Response: The NRC has no objection to editing Section B2.1.2 as 
suggested, because it does not change the loading configuration or the 
means of accomplishing preferential loading. The specification has been 
revised consistent with the comment.
    Comment D-5: One commenter noted that SER Section 1.1.1 does not 
specify the material of the tie rods of the BWR basket. The commenter 
asked why the change in materials to carbon steel for the BWR basket 
disks were made, necessitating the electroless nickel coating to 
protect from corrosion. The commenter also asked several other 
questions about the nickel coating including the criteria for applying 
the coating; how the coating is checked to ensure it is properly 
applied; how the coating is checked for long term storage and unloading 
pressures, stresses, and temperatures; if the NRC has checked the 
manufacturer's sheets for the coating; and if the BWR support disk 
coating has been evaluated for material reactions.
    Response: The tie rods of the PWR and BWR baskets are fabricated 
with ASME SA-479 Type 304 stainless steel. The applicant chose carbon 
steel as the BWR support disk material because it has higher allowable 
stresses and load carrying capability.
    The BWR support disks are coated with electroless nickel in 
accordance with American Society for Testing and Materials (ASTM) 
Specification B733-1997 (SC3, Type V, Class 1). The drawings specify 
the application in accordance with the ASTM specification, and the ASTM 
specification includes criteria to ensure proper application. All 
fabrication activities are to be carried out under a quality assurance 
program that meets the requirements of 10 CFR Part 72. As noted in SAR 
Section 3.4.1.2.4, the applicant demonstrated that the nickel coating 
is not expected to react with the spent fuel pool water during loading 
or unloading operations such that unsafe levels of flammable gas are 
produced. In the event flammable gases are produced from chemical or 
galvanic reactions, the procedures of SAR Sections 8.1 and 8.3, which 
specify that the cask user monitor the concentration of hydrogen gas 
during welding or cutting operations on the shield lid welds, ensure 
that accumulation of flammable gases is negligible and that workers are 
protected. Therefore, the NRC has reasonable assurance that the BWR 
support disk coating will not react with the spent fuel pool water 
during loading and unloading to produce unsafe levels of flammable 
gases.
    Comment D-6: Two commenters stated that neither the PSER nor the 
PSAR explain how consolidated fuel assemblies that have been canned 
will maintain confinement in the NAC-UMS system. They also note that 
the process of consolidation is expected to produce broken/damaged rods 
and that the screens will not confine the powder form 
(U3O8) of the fuel.
    Response: This comment is beyond the scope of this rule. For this 
rulemaking, the NAC-UMS storage system SAR only considers the storage 
of intact spent fuel that meets the limits as specified in the TS.
    Comment D-7: One commenter questioned the design and performance of 
the transfer cask extension and asked if it had been evaluated in 
relation to all evaluations for the TSC itself. The commenter asked if 
there is any possibility that the active fuel region could be pulled up 
into the extension area of the transfer cask and if all risks

[[Page 62589]]

associated with use of the extension have been evaluated.
    Response: The extension for the transfer cask is needed to provide 
gamma shielding to the workers while the transfer cask is being moved 
from the spent fuel pool to the VCC. The extension provides gamma 
shielding when the overall height of a standard fuel assembly has been 
increased due to the insertion of a control assembly. Because there is 
no neutron source associated with the control assembly, the NS-4-FR 
neutron shield is not needed. Because of the distribution of the active 
fuel region of a fuel assembly and the configuration of the transfer 
cask, the possibility of the active fuel region being pulled up into 
the extension is improbable.
    The structural performance of the bolts that attach the transfer 
cask extension to the PWR Class 2 transfer cask has been evaluated in 
SAR Subsection 3.4.3.3.4 for inadvertent TSC lifting against the 
retaining ring. Subsection 3.2.3.1 of the SER evaluates transfer cask 
load bearing components, including the transfer cask extension, and 
concludes that they are structurally acceptable.
    Comment D-8: Three commenters stated that a number of the NAC-UMS 
license drawings require some minor revisions, citing that the initial 
fabrication processes for the NAC-UMS have identified the need for 
additional clarifications and corrections to address editorial 
omissions for some of the current license drawings. The commenters 
noted that the requested revisions do not constitute design changes to 
the components or require revision of the existing SAR text or 
supporting evaluations. The commenters also stated that the 
incorporation of the requested revisions will significantly enhance the 
fabrication inspection process and allow authorized users of the NAC-
UMS System to fabricate the components without processing 10 CFR 72.48 
evaluations for minor variations with the current license drawings. The 
commenters' comments relate specifically to the following drawings: 
790-559, 790-560, 790-561, 790-562, 790-563, 790-564, 790-570, 790-575, 
790-581, 790-582, 790-583, 790-584, 790-585, 790-595, and 790-605.
    Response: The NRC agrees, with the exception of the addition of NS-
3 as a neutron shield material in the VCC shield plug, that the 
additional clarifications and corrections to address editorial 
omissions on the drawings do not constitute design changes to the 
components or require revisions to SAR text or the NRC's CoC, TS, or 
SER. The characteristics and evaluation of the use of NS-3 neutron 
shielding material have not been provided in the SAR; thus the NRC 
considers this aspect to be a design change. The NRC considers 
enhancements to the fabrication inspection process as a result of the 
drawing changes beneficial to all stakeholders.
    Comment D-9: Three commenters requested that B.2.2.3 of the 
Approved Contents and Design Features be revised to indicate the phrase 
``or demonstrate'' between the (existing) words ``restore'' and 
``compliance.''
    Response: The NRC agrees with the proposed clarification of the TS, 
and it has been revised accordingly.
    Comment D-10: Three commenters requested that the following 
additional note be added to both Tables B2-2 and B2-4 of the Approved 
Contents and Design Features: ``Parameters shown are nominal pre-
irradiation values.''
    Response: The NRC agrees with the proposed clarification of the TS, 
and it has been revised accordingly.
    Comment D-11: One commenter noted that a 24-inch drop would result 
in permanent deformation of the air inlets of the TSC pedestal and loss 
of part of the inlets. The commenter did not believe that the pedestal 
should be part of the inlets.
    Response: The NRC disagrees with the comment. The air inlets are an 
integral part of the pedestal or base weldment. The base weldment, that 
supports the TSC is expected to undergo yielding and partial collapse 
in a 24-inch drop of the VCC. SAR Subsection 11.2.4 presents the finite 
element analysis for calculating a bounding TSC deceleration and 
corresponding VCC base weldment deformation, that have been evaluated 
in SER Subsection 3.3.5.2. The NRC agrees with the SAR assessment that 
the 1-inch deformation of the air inlets is small compared to the 12-
inch height of the air inlet because the effect of this deformation is 
bounded by the blockage of half of the air inlets evaluated in SAR 
Subsection 11.1.2 for satisfying the radiological dose limits of 10 CFR 
72.102(a). It is important to note that although the accident 
evaluation for the concrete cask 24-inch drop has determined that the 
cask will remain functional and that there would be no radiological 
impact from the event, a full evaluation and corrective action of such 
an event's effects on cask performance, such as replacing the damaged 
VCC, would be performed according to the cask users corrective action 
and quality assurance processes.
    Comment D-12: Three commenters requested that B 3.5.2.1 (4) of the 
Approved Contents and Design Features be revised to read: ``The CHF 
design shall incorporate an impact limiter for CANISTER lifting and 
movement if a qualified single failure proof crane is not used.''
    Response: The NRC agrees with the comment. B 3.5.2.1 (4) has been 
revised as suggested.
    Comment D-13: Three commenters agreed that the following parameter 
definition clarifications are needed to Table B3-2 of the Approved 
Contents and Design Features: ``D'' should be revised to read ``Crane 
hook dead load'' and ``D*'' should be revised to read ``Apparent crane 
hook dead load''.
    Response: The NRC agrees with the comment. Table B3-2 has been 
revised as suggested.
    Comment D-14: Two commenters stated that the process of placing the 
spent fuel in the canister is not adequately justified as required by 
10 CFR 72.236(l). The industry consensus standard, ANSI/ANS-57.1, 
``Design Requirements for Light Water Reactor Fuel Handling Systems'' 
requires a translation inhibit for the spent fuel handling equipment. 
The commenters commented that although the standard permits an allowed 
bypass for this interlock, the bypass is limited to a jogging function. 
The NAC-UMS procedures do not make it clear that installed bypasses 
must be performed step-by-step as required by the standard, not in a 
continuous motion. The commenters stated that the handling equipment of 
a plant applying for approval to load dry storage canisters should be 
checked for continuous translation bypass in sensitive areas to 
eliminate the potential for a major radioactive dispersal accident.
    Response: This comment is beyond the scope of this rule. Safe fuel 
handling practices at reactor sites, including cask loading and 
unloading operations, are the responsibility of the 10 CFR Part 50 
licensee. Section 72.212 requires general licensees to determine if 
activities related to the storage of spent fuel involve any unreviewed 
safety question or change in the facility TS. The general licensee's 
evaluations and spent fuel handling practices are subject to regulatory 
oversight by the NRC's inspection process.
    Comment D-15: One commenter was concerned that a fuel assembly with 
too short bottom hardware can extend below the bottom of the poison 
panels, and asked if requiring a minimum length of bottom hardware will 
prevent this extension and if workers will measure it correctly. The 
commenter thought it would be safer to have longer poison panels and 
asked if cost-cutting is a factor.

[[Page 62590]]

    Response: Requiring a minimum length of bottom fuel hardware will 
indeed prevent the bottom of the active fuel from extending below the 
bottom of the poison panels under normal and accident conditions. The 
length of a fuel assembly's bottom hardware is usually known from the 
fuel design drawings or other fuel records. When this is not the case, 
the NRC sees no significant difficulties in the use of simple in-pool 
measurements (e.g., with a video camera and ruler) to adequately 
determine the bottom hardware dimensions. Because the required minimum 
length of fuel bottom hardware and spacer effectively precludes 
unanalyzed configurations of the fuel and poison, the NRC finds no 
basis for requiring NAC to use longer poison panels. The NRC has not 
considered cost factors in concluding that the cask design complies 
with the applicable safety regulations.

E. Welds

    Comment E-1. One commenter asked why partial penetration welds 
should be acceptable for the shield and structural lids. The commenter 
does not consider the closure redundant if the shield lid cannot be 
ultrasonically tested and stated that the structural lid needs a full 
penetration weld with ultrasonic testing because this area is crucial.
    Response: The NRC accepts the closure weld's configuration and 
examination in accordance with Interim Staff Guidance-4, Revision 1 
that allows the use of a partial penetration closure weld and a multi-
layer (i.e. progressive) liquid penetrant (PT) surface examination in 
lieu of a volumetric examination. Furthermore, ASME Code Case N-595-2, 
``Requirements for Spent Fuel Storage Canisters'' permits partial 
penetration welds for end closures using two cover plates and liquid 
penetrant examination of the weld.
    Comment E-2: One commenter was concerned about the pedestal 
weldment, stated that one inch may make a big difference in 
deformation, and asked if all possible problems have been examined.
    Response: The pedestal weldment that supports the TSC, is expected 
to undergo yielding and partial collapse in a 24-inch drop of the VCC. 
SAR Subsection 11.2.4 presents a finite element analysis for 
calculating a bounding TSC deceleration and corresponding pedestal air 
inlets deformation that has been evaluated in SER Subsection 3.3.5.2. 
The NRC agrees with the SAR assessment that the 1-inch deformation is 
small compared to the 12-inch height of the air inlet. Also, the effect 
of this deformation is bounded by that of the blockage of half of the 
air inlets that has been evaluated in SER Subsection 11.1.2 for 
satisfying the radiological dose limits of 10 CFR 72.102(a). See also 
related response D-11.

F. Structural Evaluation

    Comment F-1: One commenter asked why the pedestal plate and cask 
base plate are carbon steel and not stainless steel. The commenter 
asked for an explanation of the pedestal plate: how it is used, for 
what purpose, what shape it is, can it rust to the cask bottom plate 
and the canister bottom plate creating a problem in pulling out the 
canister, why is it not ceramic, why the VSC-24 necessitated ceramic 
tiles, and what it does long term in storage.
    Response: As depicted in SAR Figure 11.2.4-1 and Drawing 790-561, 
the pedestal or weldment plate is a 2-inch thick, 67.5-inch diameter, 
horizontal circular carbon steel plate. It provides a direct bearing 
surface to the TSC for transmitting gravity and impact vertical loads, 
through the vertical ring and inner cone baffle weldments, to the VCC 
support pad. Detail B-B of SAR Drawing 790-560 shows that a \1/4\-inch 
thick stainless steel plate is installed between the TSC bottom and the 
pedestal plate. The stainless steel plate isolates the TSC from the VCC 
carbon steel base plate. This configuration will prevent the carbon 
steel pedestal plate from rusting to the stainless steel TSC canister 
bottom. Therefore, no adherence force will develop to cause any 
shifting, deforming, or cracking of the pedestal plate in handling, as 
suggested.
    Analysis of the VSC-24 cask design is beyond the scope of this 
rule.
    Comment F-2: Two commenters noted that although the PSER structural 
analysis (Sections 3.1 and 3.4) discusses three types of tornado-
generated missiles, there is no analysis of a terrorist attack in the 
form of a fired missile. Foreign regulatory agencies are now requiring 
such an analysis. The commenters commented that the need for the 
analysis is driven further by a common location of the ISFSIs near 
international waters and that the recent introduction of high 
penetrating depleted uranium missile shells adds to the concern of a 
terrorist event. The commenters stated that an analysis of the 
vulnerability of an ISFSI to such an attack may identify the need for 
sturdier storage module surfaces, an expanded site security area, or a 
storage enclosure, and that without such an analysis, the application 
does not satisfy the requirements of 10 CFR 72.236(l).
    Response: The NRC disagrees with the comment. The NRC reviewed 
potential issues related to possible radiological sabotage of storage 
casks at reactor site ISFSIs in the 1990 rulemaking that added Subparts 
K and L to 10 CFR Part 72 (55 FR 29181; July 18,1990). The 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.
    The physical protection systems at an ISFSI and its associated 
reactor are similar in design features to ensure the detection and 
assessment of unauthorized activities. Alarm annunciations at the 
general license ISFSI are monitored by the alarm stations at the 
reactor site. Response to intrusion alarms is required. Each ISFSI is 
periodically inspected by the NRC. The licensee conducts periodic 
patrols and surveillances to ensure that the physical protection 
systems are operating within their design limits. It is the ISFSI 
licensee who is responsible for protecting spent fuel in the casks from 
sabotage rather than the certificate holder. Therefore, the commenter's 
interpretation of 10 CFR 72.236(l) as requiring the cask design to be 
analyzed for specific forms of terrorist attacks is beyond the scope of 
this rule.
    Comment F-3: One commenter noted that the NAC-MPC VCC weighs 
155,000 pounds and that the NAC-UMS VCC weighs between 221,000 and 
238,000 pounds empty, and asked if this weight has been evaluated for 
all systems. The commenter also asked why the UMS wall is 7 inches 
thicker than the MPC and the carbon steel liner thickness is 1 inch 
less in the UMS than in the MPC, suggesting that more concrete and less 
steel was used to cut costs.
    Response: The weights for five classes of VCC listed in SAR Table 
1.2-5 have been considered to establish bounding values for evaluating 
structural performance of the NAC-UMS system. The design for the 
thickness of the concrete wall and its liner plate for different 
storage cask systems is NAC's choice to meet various cask performance 
objectives such as protection from tornado missiles and radiation 
shielding and heat rejection. The design has been

[[Page 62591]]

evaluated in the SAR and found acceptable by the NRC.
    Comment F-4: One commenter asked why in Section 3.1.1.3 of the SER 
the transfer cask extension is identified as ``low alloy steel'' 
instead of ``carbon steel.''
    Response: The NRC recognizes that the transfer cask extension is 
fabricated with the ASTM A516, Grade 70, carbon steel, per SAR Drawing 
790-560. Accordingly, SER Subsection 3.1.1.3 is revised to read: ``The 
transfer cask extension is a carbon steel ring designed to be bolted to 
the transfer cask.''
    Comment F-5: Three commenters noted that either plate or forging 
material specified in ASME SA240 or ASME SA 182 should be permitted for 
both the shield lid and structural lid of the TSC. The commenters 
stated that only minor differences exist between the properties of each 
material and that these differences do not affect the performance of 
the components in the NAC-UMS System.
    Response: The NRC agrees with the comment. NAC has noted in SAR 
Section 3.4.4.1.11 that the forged material is required to have 
ultimate and yield strengths that are equal to or greater than the 
plate material. This ensures that the critical flaw size determination 
is applicable to both the SA-240 and SA-182 materials. SAR Drawing 790-
584 has been revised to permit the use of ASME SA182 as an alternate to 
SA240 for both the shield and structural lids of the TSC.

G. Thermal Evaluation

    Comment G-1: One commenter asked how the NRC can assure the public 
that determination of the design basis decay heat load was done 
properly and who checks this determination.
    Response: The design basis heat load is determined by the 
applicant, supported by their calculations, loaded in accordance with 
their procedures, and demonstrated to be in compliance with the design 
by TS surveillance measurements of the cask air inlet and air outlet 
temperatures. The NRC reviewed the SAR to provide assurance that the 
thermal design meets the regulations and performs as intended. The NRC, 
as stated in Section 4.3 of the SER, confirmed through analysis a 
sample of the decay heat loads identified in the SAR and verified 
through independent analysis that the design bases heat load is 
bounding. The NRC has concluded that the design bases heat load was 
determined properly. The user has the responsibility to load the 
canister in accordance with site-specific operating procedures that 
reflect the TS limits, including those limits imposed on heat load.
    Comment G-2: One commenter considered the fuel cladding temperature 
increase and reduction in normal temperature margin to be quite large 
when a sensitivity analysis was performed on fabrication tolerances on 
gap size between the support/heat transfer disks and the canister 
shell. The commenter asked if the fabrication tolerances can be 
tightened.
    Response: The NRC evaluated the effect of fabrication tolerances 
and has determined that the consequences are acceptable. Further 
``tightening'' of tolerances may hinder fabrication of the canister/
basket assembly and possibly adversely effect spent fuel loading and 
unloading operations.
    Comment G-3: Three commenters requested that the language of LCO 
3.1.1 (Canister Maximum Time in Vacuum Drying) with respect to ``in-
pool cooling'' be clarified to not restrict this cooling to only the 
spent fuel pool. The commenters noted that in some plant 
configurations, the use of the cask loading area or area other than the 
fuel pool may be desirable for providing cooling. The commenters also 
request that the second frequency for both surveillance requirement 
3.1.1.1 and surveillance requirement 3.1.1.2 be revised to read: ``as 
required to meet the Limiting Condition for Operation (LCO) time 
limits.''
    Response: The NRC disagrees with the comment. Insufficient 
information has been provided to describe the alternative to in-pool 
cooling. Spent fuel pools are maintained in a specific temperature 
range whereas the proposed alternative appears not to be limited in 
either temperature or configuration. Currently, more than one cooling 
method is provided because the referenced LCO 3.1.1 does allow forced 
air cooling as an alternative to in-pool cooling. Adding ``as required 
to meet Limiting Condition for Operation (LCO) time limits'' to the 
second frequency of surveillance requirement (SR) 3.1.1.1 and SR 
3.1.1.2 more clearly identifies the required time intervals, is 
acceptable to the NRC staff, and has been revised accordingly.
    Comment G-4: Three commenters stated that under LCO 3.1.6 (Concrete 
Cask Heat Removal System), SR 3.1.6.2 should be deleted. The commenters 
noted that this surveillance is already required under A 5.4, 
``Administrative Controls and Programs'' and that A 5.4 should be 
revised to clearly state for which off-normal, accident, or natural 
phenomena events the surveillance should be performed. The commenters 
stated that reference to Chapter 11 of the SAR, NUREG-1536, or 10 CFR 
72.24 and 72.122 would identify events that would require surveillance.
    Response: The NRC agrees with the comment to delete SR 3.1.6.2 
because Administrative Control A 5.4 ensures that the ISFSI will be 
inspected within 4 hours of an off-normal, accident, or natural 
phenomena event to ensure that at least half of the air inlets and 
outlets on each concrete cask are free of blockage within 24 hours. 
Also, SR 3.1.6.1 requires a comparison of the cask outlet temperature 
to the ambient temperature every 24 hours. However, the NRC does not 
agree to list the specific events in A 5.4 that could cause blockage 
because SAR Chapter 11 does not provide a comprehensive listing, but 
instead gives examples of possible events.
    Comment G-5: Two commenters noted the NAC-UMS system dissipates 
heat through conduction from the center of the fuel assembly-filled 
canister to the canister walls and away from the canister through 
natural convection by air circulation over the canister's outer 
surface. The commenters stated that the analysis of the expected 
configuration described in the PSER Section 4.4.1.2 is based on an 
unrealistic physical model that assumes concentrically centered fuel 
assemblies. In fact, conduction is radial (not axial) and is based 
solely on the physical contact of the fuel assembly with the basket 
holding the assemblies. The commenters stated that because the NAC-UMS 
system is a vertical storage system, there is a potential for 
nonuniform physical contact between the basket and the fuel assembly 
and that for this reason, hot spots may develop along the axial 
direction of the fuel rod. The commenters stated that the PSER does not 
analyze the degradation effects of these hot spots to assure cladding 
integrity throughout the license storage period and thus, the 
application does not satisfy the requirements of 10 CFR 72.236(b), (e), 
(f), and (l).
    Response: The NRC disagrees with the comment. The SAR clearly 
states that conduction and radiation are modeled in the axial and 
radial directions. Certain aspects of heat transfer are conservatively 
ignored (e.g. radiation heat transfer from the fuel tubes, and contact 
between fuel assemblies and fuel tubes, fuel tubes and support/heat 
transfer disks, and support/heat transfer disks and the canister wall). 
Consideration of these omissions would only increase the heat transfer 
from the basket assembly and result in a lowering of the calculated 
fuel cladding temperature.

[[Page 62592]]

    Comment G-6: Three commenters stated that to provide for a safer 
approach and greater flexibility in the loading and use of the NAC-UMS 
System, the TS should be revised to extend the LCO completion time 
frames based on a variable heat loading, as appropriate. The commenters 
noted that the design basis heat load time frames do not provide for an 
optimal approach to the loading and use of the first canister or those 
canisters that contain fuel with significantly lower heat loads. The 
commenters indicated that lower thermal loading will provide for 
extended time frames for many of the current LCO's and enhance 
operational safety when loading a canister with lower heat loads. The 
commenters propose that time frames for 20kW, 17kW, 14kW, 11kW, and 8kW 
be added to the current 23kW design maximum heat load used in 
developing the current LCO time frame.
    Response: The NRC agrees with the comment in principle; however, 
the NRC considers the certificate amendment process the most 
appropriate vehicle for implementing such a change at this time. The 
NRC has already completed its evaluation and solicited public comments 
by the rulemaking process, based on the request contained in the 
application. Extensive changes to the TS to include 5 levels of lower 
cask heat loads, with corresponding changes to the LCO completion time 
frames, would necessitate additional NRC review and changes to the CoC 
and SER to an extent that would warrant soliciting additional public 
comments on the proposed changes. The NRC notes that similar 
modifications have already been submitted for NRC review in connection 
with a certificate amendment request to accommodate the contents of the 
Maine Yankee spent fuel pool.

H. Technical Specifications

    Comment H-1: One commenter stated that the evacuated envelope 
helium leak test sounds inadequate and that the sniffer probe is not 
the greatest test either. The commenter said that if the shield lid 
weld cannot be ultrasonically tested, the weld cannot be called a 
redundant seal. The commenter has concerns for future leakage, 
especially in shield lid welds, because of the perceived flaws possible 
in these lid welds.
    Response: The NRC disagrees with the comment. For the types of 
helium leak tests proposed, the NRC found that these tests are capable 
of detecting leaks to the required sensitivity provided they are 
performed properly. Furthermore, liquid penetrant examinations are 
performed on all field welds' root and final surfaces, or progressive 
liquid penetrant examinations (i.e. root, mid-plane, and final surface 
of the structural closure weld) in accordance with Interim Staff 
Guidance ISG-4. For the type of welding process, the environmental 
conditions near the weld, and the austenitic stainless steel weld base 
material, there are no known delayed cracking mechanisms that could 
cause the weld to crack after it has been examined. Subsequent to 
completing the shield lid field weld, a pneumatic pressure test is 
performed and then a helium leak test is conducted in accordance with 
the leak-tight criteria of ANSI N14.5. These tests and examinations 
have been accepted by the NRC as assurance that the requirements of 10 
CFR 72.236(e) for redundant sealing of the confinement boundary have 
been met.
    Comment H-2: One commenter objected to the use of progressive 
liquid penetrant examination (PT) instead of ultrasonic examination 
(UT) for the structural lid-to-shell weld. The commenter stated the 
NRC's justification of allowable flaw size is inadequate and needs 
reevaluation. The commenter commented that the NRC admits progressive 
PT is not in agreement with ASME code and that making it easier to test 
welds and accept flaws is in the favor of the utility and vendor, not 
the safety of the public and workers. The commenter also stated that 
``sufficient intermediate layers'' is an inadequate requirement that 
should be more specific.
    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) liquid penetrant (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 failure to detect a 
flaw of this size because it did not break the surface is low because 
the liquid penetrant test is undertaken at intermediate weld pass 
levels (i.e. at \3/8\ inch for the \7/8\-inch thick structural lid 
closure weld) as well as at the root and final weld passes.
    Comment H-3: Three commenters stated that LCO 3.1.6 (Concrete Cask 
Heat Removal System) should be revised to modify Required Action B.2.2 
to allow for the use of supplemental cooling to the concrete cask with 
a completion time of 12 hours. The commenters also requested a deletion 
of the reference to transferring the canister to the transfer cask, as 
use of the transfer cask only is overly restrictive and may not be 
feasible in some conditions.
    Response: The NRC disagrees with the request to change LCO 3.1.6 to 
provide an alternative to cooling the canister (by presumably providing 
some form of forced convection) prior to being required to remove it 
from the concrete cask. No details have been provided that describe how 
this would be accomplished. Therefore, this request is not acceptable 
to the NRC. Additionally, in the NRC's judgment, the use of the 
transfer cask to provide a means of cooling should remain as an option.
    Comment H-4: Three commenters stated that the language of LCO 3.1.5 
(Canister Helium Leak Rate) should be revised to read ``demonstrate a 
helium leak rate of less than or equal to'' rather than ``demonstrate a 
helium leak rate of less than.''
    Response: The NRC agrees with the comment. The TS has been changed 
to incorporate the change in wording.
    Comment H-5: One commenter noted that ISG No. 3 lets the vendor and 
utility ``off the hook'' as to letting the public know an analysis of 
the dose consequence from a ground level canister breach with 100% fuel 
rod failure because it is not credible and the analysis is unnecessary. 
The commenter's view was that vendors and utilities do not want this 
analysis out to the public to reduce fear of such a failure. The 
commenter stated that dry cask storage is in its infancy and that such 
a failure is possible. The commenter said that the public deserves to 
know dose consequences of all related events, the NRC should be for 
public and worker safety, and the more information and education the 
public can get on dry cask storage, the more the public can help solve 
the problems and ask the right questions.
    Response: The NRC disagrees with the implication that ISG-3 was 
developed to reduce the fear of the public to nonmechanistic accidents 
such as noncredible failures of the confinement boundary. ISG-3 
clarifies the distinction between retrievability and postaccident 
recovery, and focuses on the identification and evaluation of all 
credible accident scenarios affecting public health and safety. ISG-3 
specifically places emphasis on identifying accidents with potential 
consequences resulting in the failure of

[[Page 62593]]

the confinement boundary and also recommends the modification of 
emergency plans and event detection capabilities to ensure that 
licensees have the ability to identify an accident or non-compliance 
situation. The NRC agrees with the remainder of the comment regarding 
the rights of the public pertaining to the dose consequences of 
credible events, concerns regarding public and worker safety, and 
providing information that enhances the overall understanding of dry 
cask storage.
    Comment H-6: Three commenters requested that Section A5.2 [after 
A5.2 (n)]of the TS be revised to add the following sentence: 
``Appropriate mockup fixtures may be used to demonstrate and/or to 
qualify procedures, processes, or personnel in welding, weld 
inspection, vacuum drying, helium backfilling, leak testing, and weld 
removal or cutting.''
    Response: The NRC agrees with the proposed clarification of the TS 
and it has been revised accordingly.
    Comment H-7: Three commenters requested that Table A5-1 of the TS 
be revised to indicate a Lifting Height Limit of ``24 inches.'' The 
commenters noted that this requested change is consistent with Section 
11.2.4.2 of the Preliminary Safety Evaluation Report.
    Response: The NRC agrees with the comment. Table A5-1 of the TS has 
been revised as suggested.

I. Miscellaneous

    Comment I-1. One commenter recommended that the SAR title shown in 
the proposed cask CoC state ``as amended'' instead of ``Revision 2.'' 
The commenter commented that identifying a specific SAR revision in the 
CoC may imply that a CoC amendment requiring prior NRC approval would 
be required to amend or revise the FSAR. However, the approved changes 
to 10 CFR 72.48 will allow the cask certificate holder to make changes 
to the FSAR without prior NRC approval. Also, 10 CFR 72.248 requires 
the cask certificate holder to periodically update the cask FSAR. 
Therefore, it would be more accurate and reflect the 10 CFR 72.48 
change process and the 10 CFR 72.248 FSAR update requirement if the SAR 
title shown in the CoC were to state ``as amended.'' This is typically 
how Part 50 reactor operating licenses refer to the reactor FSAR.
    Response: The NRC agrees with the comment. The SAR Title shown on 
the CoC has been revised to delete a reference to a particular SAR 
revision number.
    Comment I-2: Two commenters stated that neither the applicant nor 
the NRC has analyzed the impact of pinhole and hairline crack cladding 
defects over the 20-year license period, much less over the likely 
storage duration. The commenters stated that extraordinary attention 
must be given to the removal of water from the loaded canister and that 
the proposed vacuum drying process will not remove the water 
completely. They also asserted that available water will react with 
UO2 based fuel to form a U3O8 phase 
that could lead to unzipping of the cladding with hairline cracks or 
pinhole leaks. Therefore, they believe emerging research shows that 
incomplete drying of the spent fuel before storage combined with 
demonstrated physical processes can enlarge those defects and ``unzip'' 
the cladding, thus breaching a primary containment barrier for the 
fuel.
    Response: The NRC agrees that vacuum drying is an important 
procedure to prevent the degradation of the spent fuel cladding during 
storage. However, the NRC disagrees that the impacts of pinhole and 
hairline crack cladding defects on long term storage have not been 
evaluated.
    All spent fuel storage cask licensees are required to conduct 
vacuum drying and inert gas backfilling operations to remove oxidizing 
species from the cask and prevent cladding degradation. As discussed in 
the report, ``Evaluation of Cover Gas Impurities and Their Effects on 
the Dry Storage of LWR Spent Fuel'' (Report Number PNL-6365), and as 
described in the Standard Review Plan for Dry Cask Storage Systems 
(NUREG-1536), the combination of the low pressure and elevated 
temperature of the spent fuel during vacuum drying should remove all of 
the water from the cask and oxidizing species to an amount less than 
1.0 gram-mole. More specifically, after the liquid water has been 
removed from the storage cask, the air and water vapor are evacuated 
from the cask until a steady pressure of less than or equal to 3 
millimeters of mercury (mm Hg) is achieved and maintained for 30 
minutes. Then, the cask is backfilled with helium gas before a second 
cycle of vacuum drying (i.e., 3 mm Hg for another 30 minutes) is 
performed. The cask user is required, by the operating procedures in 
the SAR and in the TS to perform the vacuum drying procedure to ensure 
there is less than 1.0 gram-mole of oxidizing gases in the cask. These 
procedures reduce the levels of oxidizing gases to concentrations below 
those that could cause the fuel to oxidize to the 
U3O8 phase and produce larger gaps in cladding 
with existing pinhole or hairline crack defects. Therefore, the NRC has 
reasonable assurance that, if cask licensees conduct the vacuum drying 
and inert gas backfilling procedures in accordance with the TS of the 
SAR, the cladding will be protected from gross ruptures (or 
``unzipping'') during storage.
    Comment I-3: Two commenters stated that the applicant has not 
provided reasonable assurance that the NAC-UMS storage system will 
maintain the required level of confinement integrity in the proposed 
dry storage installation under the known, normal conditions; has not 
provided the required assurance that the single failure-proof 
confinement requirements for cladding and cask integrity will be 
unimpaired during the expected storage interval; and in particular, has 
not provided assurance that the integrity of the primary confinement 
barrier (cladding) will be maintained during the licensed period from 
cask closure until relicensing or shipment. The commenters also stated 
that the absence of a primary barrier violates the single failure 
requirement in 10 CFR 72.236(e) for confinement of the radioactive 
material.
    Response: The NRC disagrees with the comment. In general, the spent 
fuel cladding is not considered to be the primary confinement boundary 
of a dry storage cask. Cladding integrity is very important to prevent 
the fuel from redistributing in the storage cask and to ensure that any 
release of radioactive material from the cladding has been analyzed in 
the SAR. For example, one assumption of the confinement analysis is 
that 1%, 10%, and 100% of the fuel source term are available for 
release from the cladding under normal, off-normal, and hypothetical 
accident conditions, respectively. As a conservative approach, the 
analyses are conducted with those source term release fractions even 
though there may be no pinholes or hairline cracks in the cladding 
under normal, off-normal, and hypothetical accident conditions. 
Further, the NRC has reasonable assurance that existing cladding 
integrity will be preserved by both maintaining cladding temperatures 
below the calculated temperature limits and conducting vacuum drying 
operations in accordance with the TS. (Also, refer to the responses to 
comments C-6 and I-2.)
    As noted in SER Section 7.1, the primary confinement boundary of 
the NAC-UMS storage system includes the TSC shell, bottom baseplate, 
shield lid (including the vent and drain port cover plates), and the 
associated welds. The shield lid (with the vent and drain port cover 
plates welded to the lid) and the structural lid are independently 
welded to the upper part of the TSC shell. This

[[Page 62594]]

design provides redundant sealing of the confinement boundary and 
satisfies the requirements of 10 CFR 72.236(e). Therefore, through the 
analyses presented in SAR Chapter 7, the applicant has demonstrated 
that the NAC-UMS storage system will maintain the required level of 
confinement integrity under all conditions of storage. As documented in 
SER Chapter 7, the NRC concludes that the design of the confinement 
system of the NAC-UMS storage system is in compliance with 10 CFR Part 
72.
    Comment I-4: One commenter stated that control components should be 
low level waste and that only high level waste should be allowed in 
high level waste containers being sent to a repository. The commenter 
thinks that failure to separate high and low level waste will result in 
more handling and confusion in the long run.
    Response: The NRC has issued Interim Staff Guidance No. 9, 
entitled, ``Storage of Pressurized Water Reactor (PWR) Fuel Assembly 
Integral Hardware'' to address the authorized storage of control 
components in spent fuel storage casks. Although control rods are 
specifically excluded from the NAC-UMS authorized contents, other 
integral components (e.g., burnable poison inserts and thimble plugs) 
associated with fuel assemblies have been requested as authorized 
contents. The NRC's evaluation considered the guidance of ISG-9 in the 
preliminary SER as it relates to storage under 10 CFR Part 72. The 
aspects of the comment pertinent to the separation of high and low-
level wastes and the future acceptance criteria at a repository are 
beyond the scope of this rule.
    Comment I-5: One commenter noted that two cycles of vacuum drying 
and helium backfilling are specified for this cask design, and asked if 
the VSC-24 casks at Palisades and Pt. Beach did not have this done, how 
safe are those casks and is there any water vapor in the casks.
    Response: This rule pertains solely to the evaluation and safe 
operation of the NAC-UMS storage cask design. Comments pertaining to 
the VSC-24 or any other cask design were not a subject of the NRC's 
evaluation of the NAC-UMS design, and are thus beyond the scope of this 
rule.
    Comment I-6: One commenter stated that during cooldown for 
reflooding, very detailed definite criteria are needed for the steam 
and water being discharged. The commenter also stated that each cask 
user should have site-specific procedures in place to add to generic 
procedures so that all is ready before any cask is loaded, and that the 
NRC needs to check this activity.
    Response: The NRC has reviewed and accepted the generic unloading 
procedure guidance contained in SAR Chapter 8 that includes detailed 
criteria to control the evolution. Detailed loading and unloading 
procedures prepared using the technical basis established in the SAR 
are a site-specific aspect that is beyond the scope of this rule.
    Comment I-7: One commenter stated there should be definite criteria 
regarding records as to what are permanent and not left up to the 
licensees to decide, resulting in faded photographs and videos that 
have disappeared. The commenter suggested checking with experts on 
permanent recordkeeping.
    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 be 
maintained until decommissioning of the cask is complete. Criteria for 
records are specified in Subpart G.
    Comment I-8: One commenter remarked that ``mobile lifting frame'' 
sounds very vague. The commenter asked if the mobile lifting frame is a 
transporter, how it works, and if it has been developed.
    Response: The TS in the ``Design Features'' section establishes 
requirements for the design and operation of a canister handling 
facility, including any mobile lifting devices. The specific design for 
a mobile lifting frame was not, and is not required to be, submitted as 
part of the approval for the NAC-UMS storage cask design. Such a 
design, if implemented in the future, must be consistent with the cask 
design basis described in the SAR, the TS, and implemented on a site-
specific basis in accordance with existing heavy-loads provisions at a 
facility licensed under 10 CFR Part 50.
    Comment I-9: One commenter stated that the off-normal and accident 
conditions always assume a cask is fabricated correctly, and asked what 
problems could occur if there were fabrication problems. The commenter 
thought fabrication problems and worker mistakes are the leading 
concerns with dry casks, stated that is why the design has to have the 
best review possible, and that instructions and criteria have to be 
simple and clear. The commenter said that the casks will be on the pads 
forever and the issuance of a CoC should not be rushed.
    Response: The NRC agrees that instructions and criteria should be 
clear and that issuance of a CoC should not be rushed. Part 72 CoCs are 
issued for 20 years and are then subject to review for renewal, if 
applicable. The NAC-UMS design has been under NRC review since 1997.
    The NRC's approval of cask designs does rely, in part, on the 
design, fabrication and operation being conducted under an approved 
quality assurance (QA) program. An approved QA program includes 
programmatic controls of non conformances, corrective actions, and 
audits. The NRC has found reasonable assurance that the approved 
design, manufactured under an approved QA program, will ensure public 
health and safety under all normal, off-normal, and accident 
conditions.
    Comment I-10: One commenter stated that a quality assurance program 
is only as good as it is put to use, and that NRC's unannounced visits 
to contractors and subcontractors are very important. The commenter 
also stated that licensees need to give full documentation to changes 
in the design and keep the SAR current.
    Response: The NRC agrees with the comments.
    Comment I-11: One commenter stated the ``main problem'' is that 
nothing in the review considers or involves the review of ultimate 
disposal of spent nuclear fuel and speculated that Yucca Mountain will 
never open. The commenter made several general comments about storage 
and disposal of nuclear waste and alternative forms of energy, and 
suggested that as more spent nuclear fuel is handled and transported, 
the probability of more problems will arise.
    Response: Comments regarding the future use of a repository, 
transport and disposal of nuclear waste, and alternative energy forms 
are beyond the scope of this rule. The NRC recognizes its 
responsibility to ensure the public's health and safety, independent of 
the amount of spent fuel handling and transport that occurs under its 
regulatory oversight, now and in the future.
    Comment I-12: One commenter asked how the 5-inch carbon steel 
temporary shield is used during welding, draining, drying, and helium 
backfill operations.
    Response: A carbon steel temporary shield is placed over the 
transport cask top to shield workers from the loaded canister. Because 
gamma radiation is the predominant radiation emitted from the top of 
the canister, the 5-inch thick carbon steel temporary shield will 
reduce the gamma radiation dose to the workers.

[[Page 62595]]

    Comment I-13: One commenter asked for an explanation and dates of 
the skyshine experiments performed at Kansas State University.
    Response: The Skyshine-III, version 4.0.0 code was benchmarked with 
a Co\60\ skyshine experiment and a neutron skyshine calculation, both 
reported by Kansas State University (KSU). The Co\60\ skyshine 
experiment was performed for a Co\60\ source in a concrete silo with 
two different thickness roofs and no roof. The KSU neutron benchmark 
computations were performed for upward directed conical neutron point 
sources. Skyshine experiments are performed at KSU on an on-going 
basis. Discussions of skyshine experiments can be found in the book, 
``Radiation Shielding'' by J. Kenneth Shultis and Richard E. Faw, 
published by Prentice Hall PTE, 1996 and also in the SKYSHINE-III PC 
and SKYSHINE-KSU computer code manuals. The codes and manuals are 
available from the Oak Ridge National Laboratory's Radiation Safety 
Information Computational Center.
    Comment I-14: One commenter was concerned with computer models and 
the wording in Section 5.3 of the SER that states ``input for these 
codes * * * appears to be appropriate.'' The commenter asked if the 
input is correct.
    Response: The input data used by NAC for determining the source 
term of the design basis PWR and BWR nuclear fuel is acceptable. The 
NRC staff performed independent calculations to confirm NAC's 
evaluation of the source terms. The SAS2H module of the SCALE computer 
code uses a free form style for inputting the data that must be 
carefully reviewed to determine which keywords and variables have been 
used in the input. Also, the various fuel parameters can have a range 
of acceptable values that may be used in the input.
    Comment I-15: Three commenters requested that Section 1.b (page 2 
of 4, last paragraph) of the CoC be revised to read: ``To minimize 
contamination of the Transportable Storage Canister (TSC) exterior and 
interior of the transfer cask, clean water is circulated in the gap 
between the transfer cask and the Transportable Storage Canister (TSC) 
during loading.''
    Response: NRC agrees with this comment. The CoC has been revised 
accordingly.
    Comment I-16: Two commenters stated the PSER does not address the 
impact of the NAC-UMS cask storage system on stormwater quality.
    Response: Stormwater quality is beyond the scope of this rule. Any 
applicable stormwater quality issues will be addressed in the 10 CFR 
Part 72.212 site-specific evaluations performed prior to using the 
cask.
    Comment I-17: One commenter recommended that the wording in SER 
5.4.3 be: ``Consequently, final determination of compliance with 
72.104(a) is the responsibility of each Independent Spent Fuel Storage 
Installation (ISFSI) licensee'' instead of ``responsibility of each 
applicant for a site license.'' The commenter commented that the 
reference to an ``applicant for a site license'' is contrary to the SER 
introduction which states that the cask may be used by an ISFSI general 
licensee under 10 CFR Part 72. An ISFSI general licensee would be 
required to have site-specific evaluations in accordance with 10 CFR 
72.212 but would not be required to apply for a site license. Further, 
an ISFSI licensee would be responsible for compliance with 10 CFR 
72.104(a) at all times, not just during an application for a license.
    Response: The NRC agrees. The SER has been revised accordingly.
    Comment I-18: Three commenters requested that the first paragraph 
of Section 8.2 of the SER be revised to refer to CoC Appendix A, 
Section A 5.6 for the transport evaluation program, not Section A 5.5.
    Response: The NRC agrees with the proposed clarification of the 
SER, that has been revised accordingly.
    Comment I-19: Two commenters expressed concerns about the 
implications of long-term storage of spent nuclear fuel. One of the 
commenters had an acute interest in NRC's evaluation of this 
application because of Maine Yankee's intended use of this system for 
long-term storage following decommissioning. The commenters expected 
that the DOE will not remove all the spent nuclear fuel for 20 years or 
longer after plants cease operations and stated that whatever storage 
system is chosen must ensure the public's health and safety for an 
extended period and must ensure that the fuel will be acceptable for 
removal when the DOE is prepared to take it years in the future. One 
commenter commented that because spent fuel with pinholes or hairline 
cracks may deteriorate during storage, the NRC's evaluation of the NAC-
UMS system does not provide the necessary assurance that the spent fuel 
will be acceptable to the DOE for permanent disposal.
    Response: The NRC agrees with and shares the commenters' concerns 
regarding the safe storage of spent nuclear fuel for any and all 
lengths of time.
    The NRC's cask certification regulations stipulate that the user's 
general license to store spent fuel in a particular cask design 
terminates 20 years after the cask design's first use by that licensee. 
If the CoC has been renewed, the general license expires 20 years after 
the CoC's renewal date. The NRC will review spent fuel storage cask 
designs periodically to consider any new information, either generic to 
spent fuel storage or specific to cask designs, that may have arisen 
since issuance of the cask's CoC. The 20-year time limitation expressly 
provides an opportunity for the NRC to address any and all safe storage 
implications associated with storing spent fuel, including spent fuel 
whose cladding has pinhole leaks or hairline cracks, in particular 
casks for longer than 20 years. The NRC's initial and recertification 
reviews of cask designs are independent of the DOE's capabilities to 
accept spent fuel for permanent disposal at any point in time. However, 
the NRC's initial and renewal evaluations of a cask design have and 
will consider both the public health and safety and the retrievability 
of the spent fuel contents.
    Regarding the DOE's acceptance of spent fuel for permanent disposal 
in the future and the impact of storing spent fuel cladding with 
pinholes or hairline cracks, Dr. Ivan Itkin, Director of the Office of 
Civilian Radioactive Waste Management, addressed that issue for the 
Maine Yankee reactor. Dr. Itkin confirmed in a letter to Maine's 
Governor Angus S. King dated May 3, 2000 that DOE's contract for 
disposal with Maine Yankee covers the acceptance, transport, and 
disposal of all spent nuclear fuel from the Maine Yankee reactor, 
regardless of the condition of the spent fuel. Dr. Itkin further noted 
that, although the DOE may be currently delayed in its ability to begin 
the disposal of the Nation's commercial spent nuclear fuel, DOE has 
every intention of fulfilling its contractual obligations to all of its 
utility customers.
    Comment I-20: Two commenters requested that as a prerequisite to 
approving the proposed rule, the NRC acquire binding assurances from 
the DOE that the DOE will accept spent fuel for transport and disposal 
that has been stored in accordance with NRC-approved procedures. Those 
procedures must ensure that stored spent fuel will remain in a 
condition the DOE can accept. The commenters stated that these 
considerations and 10 CFR 72.236 preclude approval of the proposed 
certification until the NRC and the applicant have thoroughly analyzed 
and resolved critical outstanding issues.

[[Page 62596]]

    Response: The NRC disagrees that 10 CFR 72.236 requires the NRC to 
obtain binding assurances from the DOE regarding the acceptance of 
spent fuel for disposal prior to approving a storage cask design.
    DOE's efforts to develop a multi-purpose canister (MPC) program 
gave rise to several recent dual purpose (storage and transportation) 
cask design applications, including the NAC-UMS. With dual purpose 
designs, fuel no longer must be returned to the reactor spent fuel pool 
for repackaging. Dual purpose cask designs have the capability of being 
prepared for offsite transportation without having to handle individual 
fuel assemblies or return to a spent fuel pool. DOE is continuing to 
develop the cask design characteristics and parameters for disposal.
    Regarding the DOE's acceptance of spent fuel for permanent disposal 
in the future, Dr. Ivan Itkin, Director of the Office of Civilian 
Radioactive Waste Management, recently addressed that issue for the 
case of Maine Yankee reactor. Dr. Itkin confirmed in a letter to 
Maine's Governor Angus S. King dated May 3, 2000, that DOE's contract 
for disposal with Maine Yankee covers the acceptance, transport, and 
disposal of all spent nuclear fuel from the Maine Yankee reactor, 
regardless of the condition of the spent fuel. Dr. Itkin further noted 
that, although the DOE may be currently delayed in its ability to begin 
the disposal of the Nation's commercial spent nuclear fuel, DOE has 
every intention of fulfilling its contractual obligations to all of its 
utility customers. Because the DOE's spent fuel acceptance criteria for 
ultimate disposal has not yet been formalized, it would be not be 
practical to preclude a storage approval on this basis at this time.
    Comment I-21: Two commenters stated that the PSER does not address 
the necessary financial capability of a license holder to operate and 
maintain the NAC-UMS cask storage system over the 20-year license 
period.
    Response: This comment is beyond the scope of this rule. The 
financial capabilities of a certified cask design's user, a general 
licensee, are not required to be addressed in an application under 10 
CFR Part 72, Subpart L. The NRC published a proposed rule in the 
Federal Register on November 3, 1999 (64 FRN 59677) that would clarify 
the portions of 10 CFR Part 72 that apply to activities associated with 
the general license, a specific license, and a CoC. Requirements 
regarding the financial capabilities of a cask user are not identified 
as being applicable to activities associated with obtaining a CoC in 
the proposed rule.
    Comment I-22: Two commenters stated that the PSER does not address 
the necessary technical capability of the license holder to operate and 
maintain the NAC-UMS cask storage system.
    Response: This comment is beyond the scope of this rulemaking. 
Requirements on the technical capabilities of a general licensee are 
principally contained in Secs. 72.210 and 72.212. This rulemaking 
addressed question on the adequacy of the NAC-UMS cask design and 
changes to Sec. 72.214. Therefore, the preliminary SER was not required 
to address questions on the adequacy of a general licensee who may wish 
to use the NAC-UMS cask design. The NRC's requirements on the adequacy 
of a cask design are contained in Subpart L of Part 72. These 
requirements apply to an applicant for a CoC and a certificate holder, 
not a general licensee. The NRC recently added a new section 
(Sec. 72.13) to Part 72 in a final rule to clarify which requirements 
apply to a specific licensee, a general licensee, or a certificate 
holder (see 65 FR 50606; August 21, 2000). Section 72.13 specifies that 
requirements for the qualification of a spent fuel storage cask design 
do not apply to a general licensee. Rather, they apply to the 
certificate holder (and applicant for a CoC).
    Comment I-23: One commenter preferred that sensitivity studies for 
the canister deceleration g-loads and the tipover analysis be done by 
an independent party, not by NAC, and that sensitivity checks should be 
done by independent evaluation.
    Response: The SAR sensitivity analyses examine how the structural 
performance, including impact decelerations of the NAC-UMS system, 
varies with changes of modeling parameter values for the 24-inch 
vertical drop and tip-over accidents. These analyses follow standard 
engineering practice for evaluating applicability of analytical 
modeling and results. In evaluating the SAR analyses, the NRC 
determined that the analyses were adequate. Therefore, additional 
independent evaluation is not warranted.
    Comment I-24: One commenter expressed concern about long-term cask 
materials performance issues such as lead slumping and thermal aging, 
specifically as reactions that could cause creation of new materials 
and new interactions between the newly formed materials.
    Response: As part of any storage cask application review, the NRC 
evaluates the long term materials issues, such as thermal aging and 
lead slumping. The maximum calculated temperatures of the various cask 
materials do not exceed the temperature limits for any conditions of 
storage. Therefore, the NRC is assured from the analyses provided in 
SAR Chapter 4 that the thermal load from the spent fuel will not 
adversely impact the ability of those materials to perform their 
intended functions during storage. Further, lead slumping would only be 
a concern for the lead in the annulus of the transfer cask while the 
TSC is contained inside (i.e., during transfer of the fuel from the 
spent fuel pool to the VCC). When the transfer cask is not being used, 
the lead is assumed to be at ambient temperatures. As noted in SER 
Section 3.1.4.2, no softening or flow of lead is expected in the 
annulus due to lead slumping.
    Comment I-25: One commenter stated that Charpy testing of materials 
needs to be verified before any casks are loaded. The commenter asked 
who verifies the Charpy test of materials, where is the verification in 
the documents, and is the information clear.
    Response: In general, some steel materials require minimum Charpy 
impact properties for structural applications as required by the 
governing consensus standard or codes (e.g., ASTM, ASME Boiler and 
Pressure Vessel Code, etc.). The NAC-UMS storage cask utilizes several 
types of steel including stainless and carbon steel. The PWR support 
disks are fabricated with ASME SA-693, Type 630 (H1150) precipitation-
hardened steel. A typical minimum impact absorption energy requirement 
for Type 630 stainless steel is 48 foot-pounds at -110  deg.F. 
Therefore, for the NAC-UMS storage cask, there is enough ductility in 
the material so that fracture of the material is not expected at the 
minimum specified service temperature of -40  deg.F. The BWR support 
disks are fabricated from ASME SA-533, Type B, carbon steel. As noted 
in SER Section 3.1.4.1, the applicant has committed to specifying 
Charpy impact testing for each plate of material in accordance with 
ASME Code Section III, Subsection NG-2320. With regard to testing the 
Charpy impact energy, it is the responsibility of the supplier of the 
material to perform the necessary tests in accordance with the purchase 
order and to document the results of those tests on the Certified 
Materials Test Record that accompanies each lot of material shipped to 
a customer. For the NAC-UMS cask, documentation for the materials used 
to fabricate a cask will be controlled in accordance with a quality

[[Page 62597]]

control program that conforms to the requirements of 10 CFR part 72, 
Subpart G.
    Comment I-26: One commenter asked if ferritic steel is different 
than carbon steel. The commenter asked if the ferritic steel anchor 
base plate and optional lifting anchors should be stainless steel.
    Response: Ferritic steel is one of several classifications of 
stainless steel. In general, stainless steels are more resistant to 
rusting than plain-carbon and low-alloy steels. Stainless steels also 
have superior corrosion resistance because they contain relatively 
large amounts of alloying elements (e.g., chromium). Carbon steels, 
also known as plain carbon steels, have no minimum quantity for any 
alloying elements and contain only a small amount of elements other 
than the commonly accepted carbon, silicon, manganese, copper, sulfur, 
and phosphorus. Carbon steels are generally much less corrosion 
resistant than stainless steels.
    The use of the ASTM A537, Class 2, carbon steel for the VCC lifting 
lug and its anchor plate is NAC's choice for meeting its design 
objectives. SAR Subsection 3.4.3.1.3 evaluates the lifting lug and its 
anchor plate, that has been reviewed and determined structurally 
adequate in SER Subsection 3.2.3.4.
    Comment I-27: One commenter asked what the word ``chemical'' means 
in the term ``interlocking chemical lead bricks'' in Section 3.1.4.2 of 
the SER and what are the chemicals. The commenter also asked what could 
the chemicals create if water leaked into the lead chamber.
    Response: Interlocking chemical lead bricks are used in the 
transfer cask for gamma shielding. There are no chemicals added to the 
lead. The term ``chemical'' refers to a grade of lead that is specified 
in the ASTM Standard B29 for lead materials. The grade specified as 
``Chemical-Copper Lead'' is almost identical to the ``Pure Lead'' 
grade. Chemical-copper lead has 99.90% elemental lead (versus 99.94% 
elemental lead for the Pure Lead grade) and has 0.04% more alloying 
elements (e.g., copper) than Pure Lead. Because the lead is encased 
between the inner and outer shells and the top and bottom end plates of 
the transfer cask, the lead is not expected to come in contact or react 
with the spent fuel pool water.
    Comment I-28: One commenter asked several questions about the NS-4-
FR shielding material including: what other cask systems use NS-4-FR; 
how long has NS-4-FR been in use; what does the word ``reliably'' mean 
as used in SER Section 3.1.4.2; how has the NS-4-FR been tested for 
fire resistance; what can happen if the NS-4-FR gets wet because of a 
transfer cask leak; where NS-4-FR has been tested to prove it will work 
well in long term dry cask storage; and if the NRC has checked the 
materials sheets from the manufacturer of NS-4-FR for the 
specifications.
    Response: NS-4-FR has been used as a neutron shield in two licensed 
storage casks in the United States for up to 10 years and in more than 
50 licensed casks in Japan, Spain and the United Kingdom. Various 
research groups have performed both radiation and thermal stability 
testing over the last 15 years. Data from these tests adequately 
demonstrate long-term thermal and radiation stability. Further, the NRC 
has not received any reports that the shielding effectiveness of the 
NS-4-FR material has become degraded. Therefore, the NRC staff believes 
that this material is reliable for the purpose of shielding neutrons 
from personnel and the environment.
    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 the NS-4-FR resists fire and the 
generation of hydrogen gas that could cause the material to combust. 
Data supplied by the applicant show that approximately 90% of the gases 
that evolve from the NS-4-FR material when it is exposed to relatively 
high temperatures consists of water.
    The neutron shields in the transfer cask and the VCC shield plug 
are enclosed in welded steel shells so water and direct flames from a 
fire cannot get in contact with the NS-4-FR. If water were to contact 
NS-4-FR, the material is inert. Therefore, gases will not form due to 
contact between the NS-4-FR and water. Further, if fire were to contact 
the shield material, data show that the material only becomes charred 
on the surface and rapid extinguishing of the flame after the source of 
the flame is removed.
    Thermal and radiation testing of the NS-4-FR material was conducted 
in the United States by Bisco Products, Inc. and by several Japanese 
organizations to assess the material's long term performance under dry 
cask storage conditions. As part of the SAR review, the NRC staff 
routinely checks any manufacturer specification sheets to ensure that 
the material is being used in accordance with manufacturer's 
recommendations.
    Comment I-29: One commenter asked if Keeler & Long and Carboline 
epoxy enamel paint has been checked for use on casks in actual 
situations. The commenter also asked whether paint patch-up jobs 
exacerbate corrosion.
    Response: The Keeler & Long E-Series Epoxy and Carboline 890 paint 
coatings that are used to coat the exposed surfaces of the transfer 
cask are routinely recommended by the paint manufacturers for use in 
nuclear power plant applications. Further, these particular paint 
coatings have been used extensively under radiation and spent fuel pool 
water immersion conditions. Therefore, the NRC staff agrees with the 
applicant's statements in SAR Section 3.4.1.2.4 that there will be no 
adverse effects from contact between either of the paint coatings and 
spent fuel pool water because the paint will be applied in accordance 
with the paint manufacturer's recommendations. With regard to 
repainting areas where the coating has been removed (e.g., by 
scratching), paint patching will be done in accordance with the paint 
manufacturer's recommendations and the transfer cask maintenance 
program described in SAR Chapter 9, and is specifically performed to 
not exacerbate corrosion.
    Comment I-30: One commenter asked what is the date of ASME Code 
Section III, Part D, referenced in Section 3.1.4.6 of the SER. The 
commenter also asked what are the other acceptable references and their 
dates, and that the references be included in the SER.
    Response: The 1995 Edition of the ASME Boiler and Pressure Vessel 
Code (B&PVC), Section II, Part D, is referenced in Section 3.1.4.6 of 
the SER. Other acceptable sources of information are referenced in SAR 
Section 3.2 and include: the 4th Editions of the Metallic Materials 
Specification Handbook, 1992; Military handbook MIL-HDBK-5G, U.S. 
Department of Defense, 1994; ASME B&PVC Code Cases--Nuclear Components, 
1995 Edition, Code Case NC-71-17; and the Genden Engineering Services & 
Construction NS-4-FR Product Data Sheet.
    Comment I-31: One commenter stated that the dry spent fuel loading 
and unloading referenced in Evaluation Finding F3.9 should not be in 
the SER unless it has been evaluated. The commenter asked what dry 
loading procedures are being referenced.
    Response: The SAR procedures only address wet loading and unloading 
fuel from the NAC-UMS storage cask. Dry loading or unloading procedures 
are not included with this application and were not a part of the NRC's 
review. The SER finding was modified to indicate that

[[Page 62598]]

the materials are compatible with wet loading and unloading operations 
and facilities.

Summary of Final Revisions

    Based on the responses above, the NRC has modified the CoC, the TSs 
and the SER as follows:
     LCO 3.2.2 has been revised (Comment B-7).
     TS A 3.1.7, ``Fuel Cooldown Requirements'' associated with 
canister unloading procedures has been deleted from the TS (Comment C-
2).
     Parameters provided in B 3.4(6) of the ``Approved Contents 
and Design Features'' in Appendix B of CoC 1015 have been removed from 
the TS. This same set of site concrete pad and soil parameters is 
relevant to the tip-over analysis are being summarized in SAR 
Subsection 11.2.12 (Comment C-17).
     Section B 2.1.2 of the ``Approved Contents and Design 
Features'' has been edited (Comment D-4).
     B 2.2.3 of the ``Approved Contents and Design Features'' 
has been revised (Comment D-9).
     Tables B2-2 and B2-3 of the ``Approved Contents and Design 
Features'' have been revised. (Comment D-10).
     B 3.5.2.1 (4) of the Approved Contents and Design Features 
has been revised (Comment D-12).
     Table B3-2 of the Approved Contents and Design Features 
has been revised (Comment D-13).
     SER Subsection 3.1.1.3 has been revised (Comment F-4).
     SAR Drawing 790-584 has been revised to permit the use of 
ASME SA182 as an alternate to SA240 for both the shield and structural 
lids of the TSC (Comment F-5).
     The second frequency of surveillance requirement (SR) 
3.1.1.1 and SR 3.1.1.2 within LCO 3.1.1 has been revised (Comment G-3).
     SR 3.1.6.2 has been deleted (Comment G-4).
     LCO 3.1.5 (Canister Helium Leak Rate) has been revised 
(Comment H-4).
     Section A 5.2 [after A 5.2 (n)]of the TS has been revised 
(Comment H-6).
     Table A5-1 of the TS has been revised (Comment H-7).
     The SAR title on the CoC has been revised (Comment I-1).
     Section 1.b (page 2 of 4, last paragraph) of the CoC has 
been revised (Comment I-15).
     SER 5.4.3 has been revised (Comment I-17).
     Section 8.2 of the SER been revised to refer to CoC 
Appendix A, Section A 5.6 for the transport evaluation program, while 
the Section A 5.5 reference to the transport evaluation program has 
been deleted (Comment I-18).
     SER Evaluation Finding F3.9 has been revised (Comment I-
31).

Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the NRC 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.

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-UMS 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.

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 
NRC. 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, 11555 Rockville Pike, Rockville, MD and 
electronically at http://ruleforum.llnl.gov. Single copies of the 
environmental assessment and finding of no significant impact are 
available from Stan Turel, Office of Nuclear Material Safety and 
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 
telephone (301) 415-6234, 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.

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 
Nuclear Waste Policy Act (NWPA) direction to the Commission to approve 
technologies for the use of spent fuel storage at the sites of civilian 
nuclear power reactors without, to the

[[Page 62599]]

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 NRC 
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.

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.

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.

List of Subjects in 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

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

    Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 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).


    2. In Sec. 72.214, CoC 1015 is added to read as follows:


Sec. 72.214  List of approved spent fuel storage casks.

* * * * *
    Certificate Number: 1015.
    SAR Submitted by: NAC International, Inc.
    SAR Title: Final Safety Analysis Report for the NAC-UMS Universal 
Storage System.
    Docket Number: 72-1015.
    Certificate Expiration Date: November 20, 2020.
    Model Number: NAC-UMS.
* * * * *

    Dated at Rockville, Maryland, this 2nd day of October ,2000.
    For the Nuclear Regulatory Commission.
William D. Travers,
Executive Director for Operations.
[FR Doc. 00-26888 Filed 10-18-00; 8:45 am]
BILLING CODE 7590-01-P