Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 62581

Done in Washington, DC, on: October 12,

2000.

Thomas J. Billy,

Administrator.

[FR Doc. 00–26658 Filed 10–18–00; 8:45 am]

BILLING CODE 3410–DM–P

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.

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

spt@nrc.gov 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 NRCapproved 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 pdr@nrc.gov.

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

62582 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 sitespecific 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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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 sitespecific 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

§ 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

62584 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 sitespecific compliance with 10 CFR part 20

and §§ 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 B10 per

cm2 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.

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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 B10

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, offnormal 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 NS4-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

62586 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 11⁄2 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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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 sitespecific 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 NACUMS system are available for site

specific application without the need

for going through additional cask tipover 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 singlefailure-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

62588 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 longterm 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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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.

62590 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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.

62592 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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, midplane, 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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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, offnormal, 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

62594 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 sitespecific 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.

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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

Co60 skyshine experiment and a neutron

skyshine calculation, both reported by

Kansas State University (KSU). The Co60

skyshine experiment was performed for

a Co60 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 NRCapproved 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.

62596 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 multipurpose 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 §§ 72.210 and

72.212. This rulemaking addressed

question on the adequacy of the NAC–

UMS cask design and changes to

§ 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 (§ 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 °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 °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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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

62598 Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations

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 spt@nrc.gov.

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

Federal Register / Vol. 65, No. 203 / Thursday, October 19, 2000 / Rules and Regulations 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 § 72.214, CoC 1015 is added to

read as follows:

§ 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

DEPARTMENT OF COMMERCE

Bureau of Export Administration

15 CFR Part 705

[Docket No. 000601164–0164–01]

RIN 0694–AC07

Effect of Imported Articles on the

National Security

AGENCY: Bureau of Export

Administration, Commerce.

ACTION: Final rule.

SUMMARY: The Department of Commerce

is amending its regulation on the ‘‘Effect

of Imported Articles on the National

Security’’ to reduce the number of

copies of a request or application for an

investigation to be filed with the

Department from 12 copies to 1 copy,

plus the original, thereby reducing the

burden on the applicant.

EFFECTIVE DATE: This rule is effective

November 20, 2000.

FOR FURTHER INFORMATION CONTACT: Brad

Botwin, Director, Strategic Analysis

Division, Office of Strategic Industries

and Economic Security, Room 3876,