Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25241

IMPORT ASSESSMENT TABLE—

Continued

[Raw Cotton Fiber]

HTS No. Conv. fact. Cents/kg.

6302600020 .......... 1.052 1.0344

6302600030 .......... 1.052 1.0344

6302910005 .......... 1.052 1.0344

6302910015 .......... 1.1689 1.1494

6302910025 .......... 1.052 1.0344

6302910035 .......... 1.052 1.0344

6302910045 .......... 1.052 1.0344

6302910050 .......... 1.052 1.0344

6302910060 .......... 1.052 1.0344

6303110000 .......... 0.9448 0.929

6303910000 .......... 0.6429 0.6322

6304111000 .......... 1.0629 1.0451

6304190500 .......... 1.052 1.0344

6304191000 .......... 1.1689 1.1494

6304191500 .......... 0.4091 0.4023

6304192000 .......... 0.4091 0.4023

6304910020 .......... 0.9351 0.9195

6304920000 .......... 0.9351 0.9195

6505901540 .......... 0.181 0.178

6505902060 .......... 0.9935 0.9769

6505902545 .......... 0.5844 0.5746

* * * * *

Dated: April 24, 2000.

Kathleen A. Merrigan,

Administrator, Agricultural Marketing

Service.

[FR Doc. 00–10709 Filed 4–28–00; 8:45 am]

BILLING CODE 3410–02–P

NUCLEAR REGULATORY

COMMISSION

10 CFR Part 72

RIN 3150–AG 31

List of Approved Spent Fuel Storage

Casks: Holtec HI–STORM 100 Addition

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is amending its

regulations to add the Holtec HI–

STORM 100 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 May 31, 2000.

FOR FURTHER INFORMATION CONTACT:

Merri Horn, telephone (301) 415–8126,

e-mail mlh1@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 NRC

approved dry storage of spent nuclear

fuel in NRC-approved casks under a

general license, publishing a final rule

in 10 CFR Part 72 entitled, ‘‘General

License for Storage of Spent Fuel at

Power Reactor Sites’’ (55 FR 29181; July

18, 1990). This rule also established a

new Subpart L within 10 CFR Part 72

entitled, ‘‘Approval of Spent Fuel

Storage Casks,’’ containing procedures

and criteria for obtaining NRC approval

of dry storage cask designs.

Discussion

This rule will add the Holtec HI–

STORM 100 cask system to the list of

NRC approved casks for spent fuel

storage in 10 CFR 72.214. Following the

procedures specified in 10 CFR 72.230

of Subpart L, Holtec International

submitted an application for NRC

approval with the Safety Analysis

Report (SAR) entitled ‘‘Topical Safety

Analysis Report for the HI–STORM 100

Cask System.’’ The NRC evaluated the

Holtec International submittal and

issued a preliminary Safety Evaluation

Report (SER) and a proposed Certificate

of Compliance (CoC) for the Holtec

HISTORM 100 cask system. The NRC

published a proposed rule in the

Federal Register (64 FR 51271;

September 22, 1999) to add the Holtec

HI–STORM 100 cask system to the

listing in 10 CFR 72.214. The comment

period ended on December 6, 1999.

Four comment letters were received on

the proposed rule.

Based on NRC review and analysis of

public comments, the NRC staff has

modified, as appropriate, its proposed

CoC, including its appendices, the

Technical Specifications (TSs), and the

Approved Contents and Design

Features, for the Holtec HI–STORM 100

cask system. The NRC staff has also

modified its preliminary SER. Finally,

comments were received from other

industry organizations suggesting

changes to the TSs and the Approved

Contents and Design Features. Some of

these were editorial in nature, others

provided clarification and consistency,

and some reflected final refinements in

the cask design. The NRC staff agrees

with many of these suggested changes

and has incorporated them into the final

documents, as appropriate. The NRC

staff has also modified the rule language

by changing the word ‘‘Certification’’ to

‘‘Certificate’’ to clarify that it is actually

the Certificate that expires.

The NRC finds that the Holtec

International HI–STORM 100 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. Thus, use of the Holtec HI–STORM

100 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 Holtec

HI–STORM 100 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 May 31, 2000. Single copies of the

CoC and SER are available for public

inspection and/or copying for a fee at

the NRC Public Document Room, 2120

L Street, NW (Lower Level),

Washington, DC.

Summary of Public Comments on the

Proposed Rule

The NRC received four comment

letters on the proposed rule. The

commenters included a industry users

group, two members of the public, and

a State. Copies of the public comments

are available for review in the NRC

Public Document Room, 2120 L Street,

NW (Lower Level), Washington, DC

20003–1527.

Comments on the Holtec HI–STORM

100 Cask System

The comments and responses have

been grouped into eleven areas: General,

radiation protection, accident analysis,

criticality, design, welds, structural,

materials, thermal, technical

specifications, and miscellaneous.

Several of the commenters provided

specific comments on the draft CoC, the

NRC staff’s preliminary SER, the TSs,

and the applicant’s SAR. Some of the

editorial comments have been grouped.

25242 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

To the extent possible, all of the

comments on a particular subject are

grouped together. The listing of the

Holtec HI–STORM 100 cask system

within 10 CFR 72.214, ‘‘List of approved

spent fuel storage casks’’ has not been

changed as a result of the public

comments. A review of the comments

and the NRC staff’s responses follow.

A. General

Comment A.1: One commenter

expressed concern over the number of

cask designs being certified because

there would be more problems and a

lack of standardization and integration

in the country’s total waste system. The

commenter stated that this amendment

would change existing environmental

concerns as it would add one more

design, complicating the waste system

for workers at a plant. The commenter

asked how many designs would be

certified by the NRC and how many

designs could be used at one plant.

Additional designs add to more

mistakes and human error because each

design has different fabrication criteria

and handling procedures.

Response: These comments are

beyond the scope of this rule that is

focused solely on whether to add a

particular cask design, the Holtec HI–

STORM 100 cask system, to the list of

approved casks. Pursuant to the general

license, each licensee must determine

whether or not the reactor site

parameters are encompassed by the cask

design bases considered in the cask SAR

and SER. Further, each general licensee

must document this determination in

accordance with 10 CFR 72.212.

Comment A.2: One commenter stated

that the tiered environmental impact

statement (EIS) is outdated for current

dry cask design and should be redone,

particularly looking at terrorism and

sabotage at an independent spent fuel

storage installation (ISFSI).

Response: The NRC disagrees with the

comment. The environmental

assessment (EA) and finding of no

significant impact (FONSI) prepared as

required by 10 CFR Part 51 conform to

National Environmental Policy Act

(NEPA) procedural requirements.

Tiering on past EISs and EAs is a

standard process under NEPA. As stated

in the Council on Environmental

Quality’s 40 Frequently Asked

Questions, the tiering process makes

each EIS/EA of greater use and meaning

to the public as the plan or program

develops without duplication of the

analysis prepared for the previous

impact statement.

The NRC reviewed potential issues

related to possible radiological sabotage

of storage casks at reactor site ISFSIs in

the 1990 rule that added Subparts K and

L to 10 CFR Part 72 (55 FR 29181; July

18, 1990). The NRC still finds the results

of the 1990 rule current and acceptable.

In addition, each Part 72 licensee is

required by 10 CFR 73.51 or 73.55 to

develop a physical protection plan for

the ISFSI. The licensee is also required

to install systems that provide high

assurance against unauthorized

activities that could constitute an

unreasonable risk to the public health

and safety.

Comment A.3: One commenter

questioned whether the NRC was

including interim storage away from

reactors in the EA, such as at a Federal

or private storage site in Nevada or

Utah. The commenter further

questioned whether it was the NRC’s

intent to include transfer and storage at

a second site in the EA. The commenter

asked if the certification covered use at

an interim site in Nevada or Utah.

Response: The EA supports the

generic use of the Holtec HI–STORM

100 cask system under a general license.

The storage could occur at any site that

meets the definition of a general

licensee under 10 CFR Part 72. The

general licensee must evaluate the site

to determine whether or not the chosen

site parameters are enveloped by the

design bases of the approved cask as

required by 10 CFR 72.212(b)(3). The EA

does not cover transportation from one

site to another.

Comment A.4: One commenter

questioned whether the NRC claims to

have done research on the condition of

spent fuel after 20 to 50 years of storage

at a reactor in pools and dry casks, after

being unloaded twice and being

transported across the country. The

commenter stated that a detailed

analysis of what can happen to spent

fuel before it gets to Nevada or Utah

should be conducted by the NRC. The

commenter asked what the spent fuel

will be like and what the potential

environmental impacts will be after the

fuel is unloaded and transported.

Response: The NRC staff has reviewed

numerous research reports regarding the

long term condition of spent fuel in wet

and dry storage. Additionally, the NRC

has ongoing confirmatory research with

spent fuel removed from dry storage

after 10 to 20 years. Analysis of spent

fuel has included the loads from routine

shipping; and the effects, primarily due

to vibration, were found to be negligible.

The HI–STORM 100 MPC is a dualpurpose canister. Once loaded in the

MPC, the fuel is not intended to be

unloaded and reloaded as the questioner

suggests. The lid welding and testing

requirements and the structural and

thermal analyses in the SAR give the

NRC staff reasonable assurance that cask

confinement and fuel integrity will be

maintained under design basis normal,

off-normal, or accident events.

Therefore, fuel unloading should not be

necessary. Regardless of whether

unloading may be necessary, each cask

user is required to develop detailed sitespecific unloading procedures. Proper

unloading does not cause any particular

degradation to occur to the fuel.

Comment A.5: One commenter stated

that the no action alternative was

acceptable because the NRC should not

be certifying numerous designs. The

commenter stated that other agencies

such as NWTRB, EPA, OCRWM, and

DOE should be contacted for their views

on what happens to the whole waste

system as more designs are certified.

Response: The NRC disagrees with the

comment. The NRC found no inherent

design features that would result in

significant environmental impacts and

that the HI–STORM 100 design meets

regulatory requirements. Therefore,

there is no basis for denial of the

application. The NRC does not limit the

number or types of casks that may be

certified. The NRC is not required to

contact the agencies mentioned by the

commenter and we have not specifically

solicited their input. The commenter

may contact these other agencies if

interested in their views.

Comment A.6: One commenter

recommended finding a reference

(reference 1 on page 3–16 of the SER)

that is more recent than 1962.

Response: The NRC disagrees with

this comment. This reference refers to

the change of the coefficient of friction

from static to dynamic condition. The

rational behind this engineering

principle has not changed with time.

Comment A.7: One commenter stated

that the NRC should request simpler

designs because of material interactions

instead of approving designs with new

materials that have never received long

term testing for material interactions.

Response: The NRC staff disagrees

with this comment. The materials used

in casks are selected upon the basis of

the needed properties. Casks are

constructed from a limited number of

materials. The materials used in the

Holtec HI–STORM design have a long

history of use in the nuclear industry

and the performance of those materials

is well known.

Comment A.8: One commenter objects

to site specific changes that are made to

generic designs.

Response: This comment is beyond

the scope of this rule that is focused

solely on whether to place the HI–

STORM 100 cask system on the list of

approved casks. Section 72.48 permits

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25243

changes to the spent fuel storage cask as

described in the FSAR and defines the

conditions under which these changes

may be made without prior NRC

approval.

Comment A.9: One commenter stated

that it appeared that Holtec split what

appears to be one generic system into

two separate rules and asked why the

system was not certified together.

Systems should be complete when they

are proposed for rulemaking. The

commenter further stated that vendors

should apply for storage and transport at

the same time and that NRC should not

allow loading until the transportation

portion is certified.

Response: The NRC disagrees with the

comment. The HI–STAR 100 Cask

System and HI–STORM 100 Cask

System are two separate spent fuel

storage cask systems. Each is a complete

spent fuel storage cask system that

satisfies the requirements of 10 CFR Part

72. Regarding the dual-purpose (storage

and transportation) use of a cask system

or its components, separate

certifications are required for approval

of a cask design (or individual

components such as a canister) under

the provisions of use for 10 CFR Parts

71 and 72. There is no regulatory

requirement that the certification be

simultaneous.

Comment A.10: One commenter asked

a number of site-specific questions

related to Private Fuel Storage’s plans to

use the Holtec HI–STAR and HI–

STORM cask systems at the Utah site.

These issues related to cask handling,

dry transfer, sabotage scenarios,

infrastructure for unloading, etc. One

commenter stated that they understood

that Private Fuel Storage plans to use

the HI–STAR system for storage and

transport with the HI–STORM as a

companion concrete overpack, that the

metal HI–STAR overpack would be used

as a backup, and that the commenter

objected to these plans.

Response: The comment is beyond the

scope of this rule that is focused solely

on whether to add a particular design,

the Holtec HI–STORM 100 cask system,

to the list of approved casks. The rule

will enable licensees to use this cask

system under the general license

provisions of 10 CFR Part 72. The rule

does not address site-specific issues

related to potential users.

Comment A.11: One commenter

objected to calling the cask a multipurpose cask (MPC) because that stands

for storage, transport, and disposal, and

stated that the cask is not approved for

these functions which can cause

confusion when real MPCs are certified.

Response: The NRC disagrees with the

comment. The name or model number

given to the cask design is developed by

the applicant. The CoC for the Holtec

HI–STORM 100 is intended for the

interim storage of spent fuel. The use of

MPC in a dry storage cask application or

an NRC SER/CoC is not a certification

under 10 CFR Part 71 for the transport

of radioactive materials or an approval

for disposal at a high-level waste

repository.

Comment A.12: One commenter

stated that Holtec should not be allowed

to approve its own suppliers and that

the suppliers should be ASMEapproved.

Response: The NRC disagrees with the

comment. NRC regulations do not

require an ASME stamp for a cask or the

use of ASME-approved suppliers. The

design and fabrication requirements for

a certified dry cask storage system are

described in 10 CFR Part 72 and the

NRC staff’s Standard Review Plan,

NUREG–1536, ‘‘Standard Review Plan

for Dry Cask Storage Systems’’ (SRP).

Applicant submittals are reviewed to

the criteria in the SRP. Cask fabrication

activities are audited by the licensees

and inspected by the NRC staff to ensure

that components are fabricated as

designed. The CoC holder and licensee

are responsible for verifying that

fabricators are qualified. The CoC holder

and licensee must have a Quality

Assurance (QA) Program that has been

approved by the NRC as part of the

licensing or CoC issue process. This QA

program must meet the requirements of

10 CFR 72.148 and 10 CFR 72.154 for

the selection of fabricators. Also, the

procurement documents issued to the

fabricator must comply with 10 CFR

21.31. The licensee/CoC holder is

required to verify that all regulations

and CoC conditions applicable to the

container are met. The NRC inspects the

licensee/CoC holders and fabricators to

verify compliance. Additionally, many

storage cask fabricators are certified by

the American Society of Mechanical

Engineers and are N-Stamp Certificate

holders.

Comment A.13: One commenter

stated that issues should not be resolved

in telephone conferences but in public

meetings with a record in the public

document room.

Response: The NRC disagrees with the

comment. Telephone conferences are an

important mode of communication with

applicants and licensees and enable the

NRC staff to conduct its official business

efficiently. If, in these telephone

conferences, the NRC staff receives

information that would form the basis

for its regulatory decision, that

information is documented and made

available for public inspection under 10

CFR Parts 2 and 9.

Comment A.14: One commenter

stated that all details of the design

should be finalized and open for public

comment.

Response: The NRC disagrees that all

design details need to be finalized and

open for public comment before a

design is approved. The NRC staff

focuses its review on those design

details that are significant with respect

to the health and safety of the public

and/or are required to make a regulatory

finding. Design details that are pertinent

to the NRC staff’s findings are finalized

and made available for public

inspection and comment under 10 CFR

Parts 2 and 9.

B. Radiation Protection

Comment B.1: One commenter

objected to the use of less shielding for

the 100-ton transfer cask and allowing

the utilities to perform a cost-benefit

analysis to justify the use of the 100-ton

transfer cask at the expense of the

worker. The workers should receive the

minimum achievable dose and not the

maximum allowable dose. The NRC

should not allow the use of the 100-ton

transfer cask because the dose is 3 times

higher and workers should not be

treated as guinea pigs. The commenter

stated that the utilities should be

required to use the 125-ton transfer cask

which is safer and modify their facilities

to accommodate the transfer cask or

choose a cask that works for their

specific site limitations because the

utilities shouldn’t limit the shielding for

workers.

Response: NRC disagrees with this

comment. Each cask user will operate

the HI–STORM 100 under a 10 CFR Part

20 radiological protection program.

ALARA means making every reasonable

effort to maintain exposures to radiation

as far below the dose limits while taking

in account the state of technology, the

economics of improvements in relation

to the state of technology, and the

economics of improvements in relation

to benefits to the public health and

safety. As stated in Section 2.0.3 of the

SAR, the general licensee should utilize

the 125-ton transfer cask provided it is

capable of using it. However, licensees

not capable of using the more shielded

design may employ ALARA

considerations when evaluating whether

to modify its plant or use the 100-ton

transfer cask. The NRC found this

acceptable as discussed in Section 10.2

of the SER.

Comment B.2: One commenter asked

why the specific dose rate criteria for

the HI–TRAC was not given and

indicated that the criteria should be

included.

25244 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

Response: The applicant did not

provide explicit dose rate values as

design criteria for the transfer cask

designs, but stated that the radiological

requirements of 10 CFR Parts 72 and 20

as the overall shielding design

objectives for the cask system. The NRC

found this acceptable. The TSs in

Appendix A of the CoC specify dose rate

limits for the transfer casks that are

based on the applicant’s shielding

calculations.

Comment B.3: One commenter

questioned the bounding analysis for

cobalt impurities, asked how much

cobalt is really in the fuel, and if the

quantity had been tested and verified for

the real thing.

Response: The applicant’s analysis of

cobalt impurities is discussed in Section

5.2.1 of the SER. The applicant showed

that the cobalt impurity values that are

assumed in its shielding analyses were

appropriate based on industry data and

analysis of post-irradiation cooling of

older fuel. The NRC found this

acceptable. The cask user is not required

to measure the actual quantity of cobalt

in its spent fuel. The cask user will

operate the cask under a 10 CFR Part 20

radiological protection program and

verify that the cask system meets the

dose rate limits specified in the TSs.

Comment B.4: One commenter asked

why backscattering was not considered

for all cask designs.

Response: This comment is beyond

the scope of this rule that is focused

solely on whether to add a particular

cask design, the Holtec HI–STORM 100

cask system, to the list of approved

casks. Note that backscatter was

considered for the Holtec HI–STORM

100 cask system.

Comment B.5: One commenter asked

what are the various array

configurations allowed and what are the

differences between them. The

commenter asked if the cask array is

limited to two rows and for the

applicable NRC criteria.

Response: The use of the HI–STORM

design is not limited to two rows. The

NRC requires the applicant to perform

off-site dose calculations from a typical

ISFSI array to demonstrate that

radiation shielding features are

sufficient to meet the radiological

requirements of 10 CFR Parts 72.104

and 72.106. As discussed in Section

5.3.1 of the SER, the applicant used a

two-row cask array model as part of its

methodology to estimate off-site dose

rates. The values obtained by this

method can be applied to dose rate

calculations for typical cask arrays that

may consist of multiple rows. NRC

found the dose estimates to be

acceptable. Each general licensee will

identify an ISFSI configuration and

perform a site-specific dose evaluation

to demonstrate compliance with Part 72

radiological requirements.

Comment B.6: One commenter asked

why the dose rate for the bottom of the

MPC–68 was higher than for the MPC–

24 when the dose rates at the side and

top were higher for the MPC–24. The

commenter stated that the trunnion

doses showed that extreme care needs to

be taken in those areas and that the

bottom doses are really high and don’t

get enough attention.

Response: The applicant

appropriately assumed design basis fuel

loadings for each canister and estimated

dose rates at various locations. The NRC

notes that dose rates at the bottom of the

canister depend on several factors such

as the fuel hardware characteristics,

irradiation and cooling history, and the

relative position of each fuel type

within the cask system. The NRC found

that the applicant appropriately

addressed these and other factors, and

that the calculated dose rates at the

bottom and at the trunnions of the

transfer cask were acceptable. In

addition, each cask user will operate the

HI–STORM 100 under a 10 CFR Part 20

radiological protection program and

monitor dose rates during loading and

unloading.

Comment B.7: One commenter asked

what the dose for the 2x5 cask array was

at 100 meters.

Response: Figure 5.1.3 of the SAR

indicates that the dose rate for a 2x5

array at 100 meters is approximately 600

to 700 mrem/yr assuming a design basis

fuel loading and 100 percent occupancy.

Each general licensee will identify an

ISFSI configuration and perform a sitespecific dose evaluation, based partly on

site-specific characteristics, to

demonstrate compliance with Part 72

radiological requirements.

Comment B.8: One commenter asked

why other cask designs do not account

for approximate atmospheric

conditions. The commenter also asked

the conditions of weather or location for

which the air density decreases.

Response: Atmospheric density

changes daily. The measure of the

density is provided by local weather

forecasters through the barometric

pressure. When a high pressure front

passes an area, the air density is greater

than when a low pressure weather front

passes the same location.

The comment concerning other cask

designs is beyond the scope of this rule

that is focused solely on whether to

place the Holtec HI–STORM 100 cask

system on the list of approved casks. For

the HI–STORM 100, each general

licensee should consider atmospheric

conditions relevant to its ISFSI as

indicated in Section 5.4.2 of the SER.

Comment B.9: One commenter asked

how much the releases from dry storage

add to the effluent from a reactor site

and the duration of a release, and what

happens to the cask and fuel during the

release.

Response: Specific effluent releases

from reactors operated by general

licensees are beyond the scope of this

rule. However, NRC does not expect any

effluent release from the HI–STORM

100 under credible conditions. Design

basis public exposures from direct

radiation and hypothetical releases are

discussed in SER Sections 10.4 and

10.5.

Comment B.10: One commenter

approved of the condition in Appendix

B of the CoC regarding the evaluation of

engineering features (e.g. berm) that are

used for radiological protection by the

user.

Response: No response is necessary.

Comment B.11: One commenter stated

that average surface dose rates in TS

3.2.1 for transfer cask dose rates should

not be used, that the highest value

should be used, and the limit should not

be exceeded. The commenter also asked

why the side dose rates are measured

along the middle of the flat surface

section of the neutron shield rather than

on the radial steel fins where dose rates

are assumed by the commenter to be

higher.

Response: The NRC disagrees with the

comment. The specification of surface

average dose rates and the measuring

locations on the side of the neutron

shield are consistent with health

physics methods that are used to

characterize radiation fields around a

cask. The measuring locations are also

consistent with the dose rate

calculations presented in the applicant’s

shielding analysis. The cask user will

operate the HI–STORM 100 under a 10

CFR Part 20 radiological protection

program. NRC has reasonable assurance

that the general licensee’s radiological

protection and ALARA program will

detect and mitigate exposures from the

radiation fields that are expected during

operation of the HI–STORM 100 system.

Comment B.12: One commenter asked

why the dose rate for the bottom of the

transfer cask is not provided in TS 3.2.1

and what is that dose rate.

Response: Dose rate limits for the

bottom of the transfer casks are not

needed because they would not provide

a significant benefit in ensuring

compliance with regulatory limits on

occupational dose and dose to the

public. The dose limits at the top and

side of the transfer casks are adequate to

help ensure that the cask system is

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25245

safely operated in compliance with 10

CFR Part 20 and Part 72. Calculated

dose rates at the bottom of the transfer

casks are reported in Sections 5.1 and

5.4 of the SAR.

Comment B.13: One commenter

recommended that Section 5.1.2 of the

SER be revised to clarify that overpack

surface dose rates are design objectives

and are shown to be met by analysis,

and that the TSs are equal to or more

conservative than the design objectives.

Response: The NRC disagrees with

this comment. The NRC staff does agree

that the vent dose rates calculated by

the applicant are significantly less than

the applicant’s proposed design criteria.

However, the differences between the

calculated vent dose rates and the

proposed design criteria are not relevant

to the bases and findings in the SER.

The TSs in Appendix A of the CoC

specify vent dose rate limits for the

overpack that are based on the

applicant’s shielding calculations.

Therefore, a revision to the SER to

reflect the dose rate difference is not

necessary.

Comment B.14: One commenter

recommended that Section 5.4.11 and

Table 5.4–1 of the SER be clarified to

indicate that the dose rates are not peak

or maximum values.

Response: The NRC agrees with the

comment. The SER has been clarified to

state the vent dose rates are average over

the area of the vent opening. A footnote

has been added to Table 5.4.1 to clarify

values are average over surface detector

areas.

Comment B.15: One commenter

recommended that Section 10.5.1 of the

SER be revised to indicate that the

maximum MPC leak rate is utilized in

the calculations.

Response: The NRC agrees with the

comment. The SER text has been revised

accordingly.

Comment B.16: One commenter

indicated there was an inconsistency

between the accident condition whole

body and thyroid dose values referenced

in Chapter 11 of the draft SER and the

dose values calculated in Section 7 of

the applicant’s SAR.

Response: The NRC agrees with the

comment. The SER has been revised to

indicate the correct whole body and

thyroid dose values calculated by the

applicant. The accident condition whole

body total effective dose equivalent

(TEDE) is 44.1 mrem and the thyroid

dose is 4.1 mrem.

Comment B.17: One commenter

objected to the use of a 30-day duration

of a radiological release during an

accident. The commenter noted that this

assumption is stated in Interim Staff

Guidance 5 but that it is not justified in

the guidance or any accompanying

report. The commenter pointed out that

NRC regulations for ISFSIs do not

require offsite emergency planning, or

planning for the ingestion pathway

zone, and therefore, there is no basis for

assuming that something happens

within 30 days to stop the release.

Response: The NRC disagrees with the

comment. As indicated in ISG–5, Rev.1,

the 30-day assumption is consistent

with the time period that is used to

demonstrate compliance with

radiological dose requirements

associated with reactor facilities that

operate under 10 CFR Part 50. The

applicant specified corrective actions

for each accident in Chapter 11 of the

SAR. NRC believes that these corrective

actions can be reasonably achieved

within 30 days. Although NRC does not

expect effluent release from the HI–

STORM 100 under credible accident

scenarios, the 30-day assumption in the

analysis is acceptable because the NRC

staff has reasonable assurance that in

the 30-day timeframe adequate

protective measures can and will be

taken for the public in the event of a

radiological emergency. These

protective measures include

implementation of the general licensee’s

Part 50 emergency plan, evacuation of

the surrounding public, and mitigation

of radiological ingestion pathways.

Comment B.18: One commenter

objected to the assumption that a person

at the fence post (500 meters) would be

exposed for only 2000 hours/year which

is the number of working hours in a

year. The commenter stated that 8,760

hours/year should be used because a

licensee can not control who would be

in the area outside the fence or how

long they would be there. For

conservatism, the applicant should have

assumed that people, such as mothers

with pre-school aged children, the

elderly, ranchers, and farmers are

present at the fence post day-long and

year-round.

Response: The NRC agrees that 8,760

hours/year should be used and notes

that Section 7.2.9 of the HI–STORM

SAR explicitly states that: ‘‘The

individual at the site boundary is

exposed for 8,760 hours [7.0.2].’’ The

NRC staff’s independent calculations

confirmed Holtec’s calculated results, as

stated in the NRC staff’s SER. In

addition, Section 7.2.9 also assumed in

its calculations that: ‘‘The distance from

the cask to the site boundary is 100

meters.’’ With respect to hypothetical

individual exposed at the site boundary,

the methods used in the dosage

calculations cover children, the elderly,

ranchers, farmers, etc. The overall

public dose limit is protective of all

individuals because the variation of

sensitivity with age and gender was

accounted for in the selection of the

lifetime risk limit, from which the

annual public dose limit was derived.

The NRC continues to believe that the

existing regulations and approved

methodologies adequately address

public health and safety. The issue of

dose rates to children was addressed in

the Federal Register on May 21, 1991

(56 FR 23387).

Comment B.19: One commenter stated

that the dose due to direct gamma and

ingestion of radionuclides should be

considered in the dose calculation

because to ignore these pathways

underestimates the dose. The

commenter further objected to the NRC

staff stating (in the Holtec HI–STAR 100

final rule) that these pathways would be

addressed in the general licensee’s sitespecific review. The commenter stated

that there is no regulatory requirement

for these actions to be taken by the

general licensee. The commenter stated

that it is misleading for the applicant to

do a calculation that provides a

reassuring result, based on assumptions

that have nothing to do with the real

requirements of the regulations because

licensees tend to rely heavily on the

generic analyses that have been

performed by cask manufacturers.

Response: The NRC disagrees with the

comment. Although the NRC does not

expect effluent release from the HI–

STORM 100 under credible conditions,

the applicant’s method used to

determine design basis dose rates from

a hypothetical release are adequate to

demonstrate that the confinement

features are sufficient to meet the

radiological requirements of 10 CFR

72.106. The NRC staff believes the

methods applied by the applicant

conservatively bound hypothetical dose

rates to the general public. Further, 10

CFR 72.212(b)(6) requires the general

licensee to review its reactor emergency

plan and radiation protection program

to determine its effectiveness and make

changes if necessary when using a cask

listed in 10 CFR Part 72, Subpart L.

Comment B.20: One commenter stated

that the thyroid and whole body doses

should consider chlorine-36 (Cl–36)

because it will be present in the

irradiated fuel and will significantly

contribute to the dose. The commenter

points out that the Department of

Energy acknowledges that Cl–36 is one

of the significant radionuclides in

Appendix A, of the Yucca Mountain

Draft EIS.

Response: The NRC disagrees with the

comment. The NRC staff’s independent

analysis of the thyroid and whole body

dose was based on independent

25246 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

calculations using the ORIGEN

computer code, as referenced by the

commenter. The calculated contribution

of the chlorine gas was below the

truncation limit used in the calculation.

Cl–36 has an inconsequential

contribution on the total dose to an

individual.

C. Accident Analysis

Comment C.1: One commenter asked

if lead could be a missile strike barrier

from a tornado or from current weapons.

The commenter asked if missiles could

penetrate the transfer cask and canister

inside, and when the missile strike is

assumed to occur (i.e. when a loaded

transfer cask is on top of the overpack.)

The commenter stated that this needs to

be updated and evaluated.

Response: The lead backed outer shell

of HI–TRAC has been evaluated for the

required tornado missile strike. The

analysis shows that there is no

penetration consequence that would

lead to a radiological release. The threat

of missiles from weapons is beyond the

scope of this rule.

Comment C.2: One commenter

expressed concern that the transfer cask

is a real target on top of the storage cask

and asked if it had been fully evaluated

for terrorism and sabotage, particularly

when it was on top of the storage cask.

The commenter asked if the overpack

was put in place while on the pad; the

commenter felt that this would be a

target for terrorists. The commenter

asked if the transfer cask, with inner

cannister inside, could be knocked off

by a terrorist blast and fall, crash, or roll

into other casks or be upended so that

the fuel is upside down.

Response: The performance of the

transfer cask in a sabotage or terrorist

event was not evaluated. The threat of

terrorism or sabotage is beyond the

scope of this rule. See also the response

to C.8.

Comment C.3: One commenter asked

if the seismic event was based on the

actual pad analysis and not the reactor

building seismic analysis because the

conditions between the reactor building

and pad location could significantly

differ.

Response: The storage pad is a sitespecific issue and is beyond the scope

of this cask design rule. Under 10 CFR

72.212, the cask operators are required

to perform written evaluations to ensure

that storage pads have been designed to

adequately support the stored casks.

The licensee using a particular cask

design has the responsibility under the

general license to evaluate the match

between reactor site parameters and the

range of site conditions (i.e. the

envelope) reviewed by the NRC for an

approved cask.

Comment C.4: One commenter asked

how a full cask array would behave in

a seismic event. The commenter asked

what buildings or equipment are

allowed on the pad that could crash into

the casks during a seismic event, such

as the transfer equipment. The

commenter asked if a crack or ‘‘push

up’’ of the pad could cause the cask to

roll (down an incline or into water).

Response: The SAR indicates that the

HI–STORM 100 overpack will neither

slide nor tip over due to a seismic event

with the design-basis earthquake input

listed in Section 3.4.2 of the SER. The

use of a general licensed cask by a

utility requires that the user ensure that

the site is not subject to any potential

accident that has not been analyzed for

the general license. This would include

any potential design basis earthquakes

that were not enveloped by the NRC

SER for the cask or any site conditions

associated with the actual pad and cask

locations that could affect the cask

design.

Comment C.5: One commenter asked

what the design-basis earthquake on top

of the surface pad was and where it

occurred. The commenter questioned

why the bottom surface was not

evaluated because the ground can push

up and crack or cause heaving in the

concrete and how the condition of the

bottom surface is known.

Response: The design basis

earthquake is the most severe

earthquake that has been historically

reported for a particular site and

surrounding area, with sufficient

margins for the limited accuracy,

quantity, and period of time in which

historical data have been accumulated.

Structure, systems, and components

important to safety are designed to

withstand the effects of this earthquake

without loss of capability to perform

their safety functions. The design basis

earthquake is described by an

appropriate response spectrum

anchored at the peak ground

acceleration. The response is then

amplified through the pad to obtain the

input response spectrum at the top of

the pad (or at the bottom of the cask) for

cask seismic evaluation. Soil and

storage pad interaction is a site-specific

issue that will be addresses in the cask

user’s 10 CFR 72.212 evaluation and is

beyond the scope of this rule.

Comment C.6: One commenter asked

what happens if the pad is cracked and

heaving up as the cask is tipping over

because a tornado or seismic event will

likely affect both the pad and the casks.

Response: The NRC does not consider

the scenario described by the

commenter to be credible. The

evaluation in Section 3 of the SAR

shows that tipover will not occur.

However, as a defense-in-depth

measure, cask tipover is also evaluated

in Section 3 of the SAR and discussed

in Section 3.4.2 of the SER.

Comment C.7: One commenter asked

if the cask could become upside down

in a tornado or seismic event and if it

happened would the top of the fuel hit

the underside of the MPC lid with the

weight on the overpack lid studs.

Response: The HI–STORM 100

overpack is evaluated for tornado,

tornado missiles, and seismic events in

Section 3 of the SAR. The results

indicate that the cask will not tip over.

Therefore, the cask will not become

upside down.

Comment C.8: One commenter stated

that an airplane crash with its fuel fire

should be evaluated, including crash

into a full cask array, damage to the pad,

and a fuel and airplane explosion after

the crash. The commenter stated that an

anti-missile device with an incendiary

device and a truck bomb should be

analyzed for the cask transfer facility

(CTF).

Response: The NRC disagrees with the

comment. Before using the HI–STORM

100 casks, the general licensee must

evaluate the site to determine whether

or not the chosen site parameters are

enveloped by the design bases of the

approved casks as required by 10 CFR

72.212(b)(3). The licensee’s site

evaluation should consider the effects of

nearby transportation and military

activities.

The NRC reviewed potential issues

related to possible radiological sabotage

of storage casks at reactor site ISFSIs in

the 1990 rule that added Subparts K and

L to 10 CFR Part 72 (55 FR 29181; July

18, 1990). NRC regulations in 10 CFR

Part 72 establish physical protection

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

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

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25247

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 NRC. Also,

the licensee conducts periodic patrols

and surveillances to ensure that the

physical protection systems are

operating within their design limits. The

ISFSI licensee is responsible for

protecting spent fuel in the casks from

sabotage not the certificate holder.

Comments on the existing regulations

specifying what type of sabotage events

must be considered are beyond the

scope of this rule.

Comment C.9: One commenter

questioned why the tornado missile test

simulated a pulse impact of a vehicle

and stated that a sharp object such as a

metal pole or other items that might be

in the vicinity of a real pad would do

more penetration damage.

Response: In addition to the 4,000-

pound automobile impacting at a 126

mph velocity, the SAR also provided

analyses for two more missiles

impacting at 126 mph velocity: a 1-in

diameter steel sphere and an 8-in

diameter rigid cylinder. Results of the

analyses show that the 4,000 pound

automobile produces the highest impact

force on the cask because it has the

largest mass. Based on these results, the

NRC staff has reasonable assurance that

the 4,000 pound vehicle bounds the

effect of other credible types of tornado

missiles.

Comment C.10: One commenter stated

that the 15-minute transporter fire is not

valid. A big plane crash with its fuel

should be evaluated as well as a

sabotage missile penetration with an

incendiary device.

Response: The NRC disagrees with

this comment. The basis for the 4.8-

minute fire (not a 15 minute fire, see

response to comment C.18) is associated

with the time it would take to burn 50

gallons of fuel, presumably carried by

the transporter. The CoC, Appendix B,

Section 3.4, states that ‘‘the on-site

transporter fuel tank will contain no

more than 50 gallons of diesel fuel

while handling a loaded OVERPACK or

TRANSFER CASK.’’ Other modes of

transport causing the fire (e.g.,

airplanes, trains, delivery trucks or

missiles) are not considered as plausible

and are beyond the scope of this rule.

Before using the HI–STORM casks, the

general licensee must evaluate the site

to determine whether or not the chosen

site parameters are enveloped by the

design bases of the approved cask as

required by 10 CFR 72.212(b)(3).

Included in this evaluation is the

verification that no credible source of an

external explosion that would produce

an external pressure above that analyzed

in the SAR and that any cask handling

equipment used to move the HI–STORM

cask to the pad is limited to 50 gallons

of fuel (refer to CoC, Appendix B,

Section 3.4—Site Specific Parameters

and Analyses).

Comment C.11: One commenter asked

why there were no calculations for the

bottom plate, overpack lid, etc. in a fire

because the temperatures of these plates

were important to know and could

affect the pad or fire fighting equipment.

Response: The applicant did calculate

the temperatures for the bottom plate

and the overpack lid. However, these

temperatures were not reported in the

SAR. Not all calculated temperatures are

reported in the SAR. With respect to the

impact of fire on the pad or fire fighting

equipment, a postulated 50 gallon fuel

source would have minimal impact on

those components. The heat generated

by the pool of fuel is directed upward

where the fuel is in a gaseous state. The

limiting temperatures will occur above

the surface of the concrete pad. Because

the fuel has to vaporize in order to burn,

the liquid fuel on the concrete will have

minimal impact on the bottom plate of

the overpack lid (in a liquid state, the

fuel is cool). The duration of the fire is

less than 4 minutes. The impact on the

fire fighting equipment would be

minimal, if any. Table 4–3 of the SER

was modified to indicate that the

temperatures were not reported.

Comment C.12: One commenter asked

how the 45,000 MWD/MTU for 5 years

related to the sabotage and terrorist

evaluation for radiation disposal and

stated that the evaluation is outdated.

Response: The comment on the

sabotage report is beyond the scope of

this rule. See the discussion in the

response to C.8.

Comment C.13: One commenter asked

if the water jacket could be pierced with

an anti-missile gun or if a terrorist could

shoot the jacket full of holes, and what

are the consequences if these events did

occur.

Response: The specific threat of an

anti-missile gun or other small arms

against the HI–STORM 100 is beyond

the scope of this rule. However, the

resultant dose rate for an assumed

complete loss of the water jacket is

addressed in Section 5.1.2 of the SAR.

The analysis indicates that the off-site

dose at 100 meters will be below the 5

rem accident limit in 10 CFR 72.106.

Comment C.14: One commenter asked

why a burial under a landslide during

a seismic event is not considered.

Response: Burying a cask due to

seismic event, landside, or tornado is

considered a very unlikely event.

Considering the unlikeliness of the

event coupled with the casks being able

to withstand these events make burying

and any adverse consequence in the

opinion of the NRC not credible.

Comment C.15: One commenter asked

why a vertical drop of a loaded transfer

cask is not considered a credible

accident, particularly as it is perched on

top of the concrete overpack to load.

Response: A vertical drop of a transfer

cask is not considered credible because

vertical lifting of a loaded transfer cask

must be performed with structures and

components designed to prevent a cask

drop. The criteria for those structures

and components are specified in Section

3.5 of Appendix B to CoC No. 1014. The

restrictions on vertical lifting are

specified in Section 5.5 of the TSs

(Appendix A to the CoC).

Comment C.16: One commenter stated

that defense-in-depth is needed for

sabotage events which could cause a

tipover.

Response: The NRC disagrees with

this comment. Sabotage events are

beyond the scope of this rule. They are

considered in Part 73. Furthermore, the

SAR demonstrates that the HI–STORM

100 overpack will not tipover due to a

design basis accident. However, as an

added defense-in-depth measure, cask

tipover is evaluated in Section 3 of the

SAR and discussed in Section 3.4.2 of

the SER.

Comment C.17: One commenter asked

why a postulated explosion from a truck

bomb at the pad fence was not

evaluated.

Response: The specific threat of a

truck bomb is beyond the scope of this

rule. The response to C.8 addresses

radiological sabotage of storage casks at

generally licensed ISFSIs.

Comment C.18: One commenter asked

the basis for the 217-second fire for the

overpack and the 4.8-minute fire for the

transfer cask. The commenter also asked

if the NRC assumed that nothing on the

vehicle or in the vicinity (such as grass

or trees or other structures) will burn

and cause the fire to burn longer.

Response: The duration of a fire burn

is based on several factors. One factor is

the rate at which the fire burns,

normally categorized as inches of fuel

burned per minute. The burn rate

(inches per minute) is the same for both

the overpack and the transfer cask

because the source of fuel is the same

(e.g., diesel fuel). The duration of the

burn comes from the postulated depth

of the pool of fuel. A conservative

estimate of the time of burn is to assume

that the spilled fuel does not extend

beyond 1 meter of the surface of the

overpack or the surface of the transfer

cask. (In reality, the fuel will spill

significantly farther than one meter on

25248 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

a flat surface, just as spilling a bucket of

water on the ground, and will not

accumulate to any significant depth

which creates a shorter fire burn time.)

Because the outer diameter of the

overpack and the outer diameter of the

transfer cask are different, the

postulated depth or height that 50

gallons of fuel is postulated to reach

will differ for the two cases. The case

with the higher column of fuel will burn

longer. Because the surface area of the

pool of fuel for the overpack is 1.3 times

larger than for the transfer cask, the pool

of fuel for the overpack will be lower

(given the same volume of available

fuel, e.g., 50 gallons). A lower pool of

fuel will burn quicker. (Note that the

burn rate is in inches of fuel per minute,

and a smaller column of fuel will burn

quicker than a higher column of fuel).

Therefore, the burn time for the

overpack is shorter than the burn time

for the transfer cask.

With respect to other flammable

sources that could catch fire, before

using the HI–STORM cask, the general

licensee must evaluate the site to

determine whether or not the chosen

site parameters are enveloped by the

design bases of the approved cask as

required by 10 CFR 72.212(b)(3).

Included in this evaluation is the

verification that the cask handling

equipment used to move the concrete

cask to the pad is limited to 50 gallons

of fuel (refer to CoC, Appendix B,

Section 3.4.5) and that the assumptions

used in the SAR bound the

consequences for the proposed site.

Additional assessments would have to

be performed if other sources are

identified that could result in a more

limiting fire.

Comment C.19: One commenter

objected to the use of the leakage rate

used by Holtec because it is based on an

analysis of a transportation cask rather

than a storage cask, for which the NRC

and industry have different design and

testing requirements. The commenter

noted that the small assumed leakage

rate and calculation methodology in

NUREG/CR–6487 are based on ANSI

standard N14.5 for transportation casks.

ANSI N14.5 assumes that casks will be

leak-tested periodically before shipment

and after maintenance and repair. The

commenter pointed out that some

ISFSIs have no design provisions for

testing helium leakage during storage

and no provisions for repairing and

maintaining casks and testing for

leakage after repair and maintenance.

Therefore, it is inappropriate to assume

that these storage casks will have the

same small leakage rate as

transportation casks for which leakage

potential is designed and planned to be

monitored. The commenter stated that

neither Holtec nor the NRC has any

basis for relying on NUREG–1617 to

assume a small leakage rate in a storage

cask breach.

Response: The NRC disagrees with the

comment. The ANSI N14.5 standard

was developed to determine allowable

leak rates for shipping packages that

employ mechanical seals, which

typically undergo repetitive use.

Periodic testing is prescribed for the

mechanical seal to ensure it has not

degraded from repetitive use and/or seal

maintenance. The analytic technique in

ANSI N14.5 that is used to determine a

leak rate across an assumed leak path is

valid for determining an assumed leak

rate across the confinement boundary of

a welded canister. An off-site dose can

be subsequently calculated using

standard atmospheric dispersion

principles and assuming a partial

release of the cask constituents at the

calculated leak rate. The welded closure

is leak-tested to a sensitivity equal to the

calculated leak rate to ensure integrity

of the confinement system before

storage operations. Periodic testing of

the confinement boundary is not

applicable because the welded

confinement boundary is designed to

remain intact during normal, off-normal,

and accident conditions for the lifetime

of the canister.

Comment C.20: One commenter stated

that the methodology used in NUREG/

CR–6487 may not apply for accidents

that exceed the design basis accident.

The allowed leak hole size can easily be

exceeded in accidents involving

sabotage such as an impact with a

MILAN or TOW–2 hand held anti-tank

device, a jet engine, or military

ordnance.

Response: The NRC disagrees with the

comment. Consideration of accidents

that exceed design basis is not required

by 10 CFR Part 72 and is beyond the

scope of the NRC staff’s review. The

threat of accidents involving sabotage is

beyond the scope of this rule. Sabotage

issues are covered by 10 CFR Part 73.

Comment C.21: One commenter stated

that Holtec should consider a 300 gallon

fire or a 6,000 gallon fire. The

commenter stated that these are credible

accidents at an ISFSI and should be

considered. A heavy haul tractor carries

300 gallons of fuel and is likely to be

used at ISFSIs. Locomotives that carry

casks to ISFSIs may carry 6,000 gallons

of diesel fuel.

Response: The NRC disagrees with the

comment. The analysis need only

address the maximum permissible

source of fuel at the storage site near the

HI–STORM 100 system (10 CFR Part

72). Section 3.4 of Appendix B to the

CoC limits the source of fuel near the

HI–STORM 100 system to 50 gallons.

Licensees are required to verify that all

conditions of the CoC are met.

Comment C.22: One commenter stated

that Holtec’s fire analysis is deficient

because the fire calculations assume

that the fire takes place outside the

concrete storage cask and does not

consider the possibility of a fuel fire

being drawn into the intake vent of the

HI–STORM 100 cask.

Response: The NRC staff disagrees

with the comment. The purpose of the

fire analysis is to assess the

consequences of a postulated fire on the

HI–STORM system. The elements of

interest are the impact of the fire on the

peak clad temperature and the impact of

the fire on the system materials. A 50-

gallon fuel supply will have a very short

burn duration. Applying the

conservative assumptions of 10 CFR

Part 71, a 50-gallon fuel supply would

theoretically result in a pool size of 0.54

inches if limited to a one-meter spread

around the overpack. The burn duration

of the fuel in this configuration is 3.6

minutes. This burn duration will have

insignificant impact on the peak clad

temperature. The heat capacity of the

system is too great to have an

appreciable feedback on the peak clad

temperature for a short duration

transient. The greatest impact of a fire

will be felt on the overpack. A bounding

analysis was performed on the overpack

by imposing a maximum burn

temperature (specified in 10 CFR Part

71) on the entire outer surface of the

overpack. This maximizes the impact on

the steel liner and the concrete. In a less

conservative calculation, the maximum

temperature will be limited to the lower

portion of the overpack. For additional

conservatism, the applicant increased

the inside temperature of the overpack

to 300°F to account for heating of the air

as it passes through the vents. As

illustrated in SAR Figures 11.2.1—

11.2.5, this bounding calculation

illustrates that only the outer boundary

of the concrete exceeds the temperature

limit for concrete. At a depth of one

inch into the concrete the temperature

limit is not challenged. If a conservative

assumption postulates the fire to occur

inside the vent, similar results would

occur because there is only a limited

amount of energy (BTU) that can be

deposited into the massive overpack

structure. Exceeding the concrete

temperature limit only at the concrete

surface does not lead to a safety

concern, and therefore, the SAR analysis

is acceptable.

Comment C.23: One commenter stated

that the consequences of a hit by an

anti-tank missile, such as the MILAN or

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25249

TOW–2 missile should be considered.

The commenter noted that the

regulations only require a licensee to

install systems that protect against

unauthorized entry; however, entry to a

site is not necessary to successfully

carry out sabotage using an anti-tank

missile. The commenter stated that the

NRC should place additional conditions

in the CoC to lower the probability of a

sabotage event. The commenter further

pointed out that the NRC has been

inconsistent and arbitrary in

determining whether to treat sabotage

issues as site-specific or generic.

Response: The NRC disagrees with the

comment. The threat of an anti-tank

missile and other sabotage events is

beyond the scope of this rule.

Requirements on radiological sabotage

are covered in 10 CFR Part 73 and apply

to both ISFSIs and spent fuel storage

cask designs. Therefore, comments on a

specific threat or mode of attack are

beyond the scope of this Part 72 rule.

See also the response to C.8 addressing

radiological sabotage of storage casks at

reactor site ISFSIs.

D. Criticality

Comment D.1: One commenter

objected to the assumption on the

continued efficacy of the boral over a

20-year storage period because it has

never been tested or proven.

Response: The NRC disagrees with the

comment. The NRC staff does not

consider the loss of fixed neutron

poisons credible after installation into

the cask because the poisons are fixed

in place and contained. The neutron

absorber is designed to remain effective

in the HI–STORM system for a storage

period greater than 20 years and there

are no credible means to lose the

neutron absorber. Section 6.3.2 of the

HI–STORM SAR describes the neutron

absorber and its environment, and

evaluated boron depletion due to

neutron absorption. Section 9.1.5.3 of

the SAR describes the testing

procedures for the neutron absorber

material. The neutron absorber material

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. The

compositions and densities for the

materials in the computer models were

reviewed by the NRC staff and

determined to be acceptable. This

material is not unique and is commonly

used in other spent fuel storage and

transportation applications.

Comment D.2: One commenter asked

if Boral had ever been used in any dry

storage casks before and if it had, how

long and had it been tested. The

commenter asked if this was an

experiment with a new application. The

commenter further asked what proof

was available to show the continued

efficacy of a boral panel. The

commenter asked what other fuel

storage and transport applications

utilized Boral and stated that it should

be documented in the SER.

Response: As described in SAR

section 1.2.1.3.1, Boral has been used in

environments comparable to those in

spent fuel storage casks since the 1950s,

and in spent fuel shipping casks for

Canadian spent fuel in the 1960s. In the

United States, Boral has been used in

numerous other spent fuel

transportation casks since the early

1980’s and in storage casks since the

early 1990’s. Some of the casks that use

Boral are the NAC–I28 S/T, NAC–S/T,

the NUHOMS–24P, NUHOMS MP–187,

and BMI–1. The NRC disagrees that the

HI–STORM SER should include a list of

other casks that use Boral. Information

on other spent fuel casks and Boral is

publicly available. The response to

comment D.1 discusses the efficacy of

Boral and why testing other than initial

fabrication testing is not necessary.

Comment D.3: One commenter stated

that a test should be conducted to verify

the presence and uniformity of the

neutron absorber in fabrication.

Response: The presence and

uniformity of the neutron absorber is

verified as described in Section 9.1.5.3

of the SAR. The neutron absorber

material 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.

Comment D.4: One commenter asked

if water injection in unloading reflood

could result in large amounts of steam

generation and two-phase flow

conditions inside the MPC cavity

causing over pressurization of the

confinement boundary and a potential

criticality.

Response: As stated in SAR section

6.4.2.1, the HI–STORM system was

evaluated with various water densities

inside the cask. The cask met the design

criterion of keff less than or equal to 0.95

for all credible flooding conditions. The

cask is most reactive when filled with

full density water. As can be seen in

SAR Table 6.4.1, the cask reactivity

decreases when filled with low density

water (i.e., steam).

In addition, Section 4.5.1.1.6

describes the cask cooldown and reflood

analysis during fuel unloading

operation. This section of the SAR states

that before reflooding the cask with

water, the helium inside the MPC is

cooled to below 200°F which is below

the boiling point of water. The

procedures are outlined in Section 8.3.1

of the SAR with reference to TS 3.1.3.

These procedures limit steam generation

and two-phase-flow interactions with

the fuel to acceptable levels, thereby

preventing over pressurization of the

MPC.

E. Design

Comment E.1: One commenter asked

if there are three MPCs that are NRCcertified for storage and transfer because

the SER states that they are evaluated

and approved.

Response: As stated in Condition 1.b.

of the CoC and in Section 1.1 of the

SER, there are three types of MPCs that

can be used in the HI–STORM 100 Cask

System: The MPC–24, the MPC–68, and

the MPC–68F. The MPC–24 holds up to

24 PWR fuel assemblies that must be

intact. The MPC–68 holds up to 68 BWR

fuel assemblies that may be intact or

damaged (i.e., with known or suspected

cladding defects greater than hairline

cracks or pinholes). The MPC–68F holds

up to 68 BWR fuel assemblies that may

be intact, damaged, or in the form of

fuel debris (i.e., with known or

suspected defects such as ruptured fuel

rods, severed fuel rods, and loose fuel

pellets). All three MPCs have the same

external dimensions. Section 1.2.1.1 and

Table 1.2.1 of the SAR has been revised

to clarify that there are three types of

MPCs.

Comment E.2: One commenter asked

how and to what the trunnion is

attached, and what it is made of.

Response: The trunnions are attached

by welds to the inner and outer shell

and to the HI–TRAC top flange. The

trunnions are fabricated of SB–637–

N07718 steel and SA–350–LF3 steel.

Comment E.3: One commenter stated

that the concrete for the overpack

should be reinforced and asked why it

is not reinforced.

Response: The NRC disagrees with

this comment. The main function of the

concrete encased between the steel

shells in the HI–STORM 100 overpack

is shielding. The structural strength of

the HI–STORM 100 overpack is

provided by the inner and outer carbon

steel shells. The concrete, on the other

hand, will provide an added benefit to

the HI–STORM 100 overpack because it

will increase the stiffness and weight of

the overpack to resist external forces

due to seismic, tornado, and tornado

missiles.

Comment E.4: One commenter asked

if the pedestal could shift in movement

and touch the liner or if it could corrode

to the carbon steel liner or baseplate.

The commenter also asked what the

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baseplate was made of and if a ceramic

baseplate should be used.

Response: The pedestal consists of

concrete, 17 inches thick, encased in a

steel shell. This shell is welded to the

steel overpack baseplate, and the weld

is examined according to the ASME

Code Section V. Stresses in the pedestal

have safety factors exceeding 16. The

pedestal will not shift. The exterior and

interior surfaces of the overpack are

coated with an epoxy paint to prevent

corrosion. The overpack baseplate is

made of carbon steel that meets the

design criteria.

Comment E.5: One commenter stated

that jamming of parts could be a

problem because unjamming the parts

could cause damage (during both

loading and unloading). The commenter

further asked if the cask had been tested

for jamming and what the situation

would be after 20 to 40 years in storage.

Response: Stainless steel shims,

depicted in Detail T of drawing 1495,

sheet 5, prevent the MPC from

contacting the overpack interior and

preclude the paint from being scraped

during the operational steps. The drop

accident analyses cause stresses which

significantly bound the stresses that

could occur during normal handling

operations. Therefore, damage to the

MPC during loading and unloading into

the overpack is not credible.

The calculation in the SAR

demonstrated that there will be no

jamming of the MPC in the overpack

under the most severe stack-up of

tolerances. The cask has not been tested

for jamming; however, a dry run of all

operational steps is required before use

of the system.

The license life of all overpack and

MPC components is 20 years. The

applicant engineered the overpack, HI–

TRAC, and MPC for 40 years of design

life. More detailed information

regarding the service life of the

overpack, HI–TRAC, and MPC can be

found in Sections 3.4.11 and 3.4.12 of

the SAR.

Comment E.6: One commenter stated

that the clearances were not adequate.

The commenter asked if the wet fuel

would be inserted into the overpack in

the same way as the dry run and what

happens if the crane does not have the

MPC completely vertical when inserting

it in the overpack or if the HI–TRAC or

pad is not level.

Response: There is no adverse

tolerance stack-up that would prevent

the insertion of the MPC into the

overpack. Additionally, the dry run will

verify that the MPC can be inserted into

the HI–TRAC and overpack. All cell

plates of the MPC are constructed of

stainless steel that is not effected by

immersion in water; therefore, the

tolerances for the dry run would not

change and the wet fuel will go in the

same way as in the dry run.

Comment E.7: One commenter is

concerned that the manufacturer’s

tolerances are not clear if fabrication is

the minimum margin of safety or

minimum clearance allowed.

Response: The most severe ‘‘stack-up’’

of manufacturer’s tolerances provides

sufficient clearance for insertion of the

MPC into the HI–TRAC. The minimum

clearance allowed is thus met. The cask

could be manufactured to the minimum

allowed clearances, but this would not

reduce the minimum margin of safety.

Comment E.8: One commenter asked

if there would be a problem if the radial

clearance of the HI–TRAC MPC is at a

maximum and the radial clearance of

the MPC overpack is at the minimum

allowed. The commenter asked how

much leeway is allowed in fabrication

in both of these radial clearance

measurements.

Response: No operational problem

exists if the radial clearance of HI–

TRAC/MPC is at a maximum tolerance

‘‘stack-up’’ and the radial clearance of

the MPC overpack is at the minimum

tolerance ‘‘stack-up.’’ These tolerances

have been evaluated for all

manufacturer’s design criteria

requirements and for all design

temperatures. The largest allowable

radial dimension of the HI–TRAC is

greater than the smallest allowable

radial dimension of the overpack.

Fabrication requirements, including

tolerances, are stated on the drawings.

These tolerances provide sufficient

clearance for operations.

Comment E.9: One commenter

expressed concern over the 13⁄16-inch

difference in the maximum MPC

diameter and minimum overpack

internal diameter because it was a

minuscule amount for fabrication. The

commenter also asked what was meant

by average radial clearance of about 0.4

inches and stated that it was not a lot

of clearance.

Response: The 13⁄16-inches is the

minimum clearance accounting for

tolerances between the MPC diameter

and the channels/shims that are

attached to inner shell of overpack. The

channels/shims provide guidance for

MPC insertion, position MPC within the

overpack, and allow the cooling air flow

to circulate through the overpack. The

minimum clearance between the MPC

and overpack inner shell is

approximately 5 inches without the

channels. Both the clearance between

the MPC and channels/shims and

between the MPC and overpack inner

shell are considered to be acceptable.

The SER has been changed to clarify

that 13⁄16-inches is the clearance

between the maximum MPC diameter

and the channels/shims that are

attached to inner shell of overpack

rather than between the MPC and

overpack inner shell. The average radial

clearance is diametral clearance divided

by two.

Comment E.10: One commenter asked

what the computed decrease (page 3–9

of the SER) was related to. The

commenter expressed concern that these

were very small calculation amounts

(0.11 inches) to depend on computer

accuracy.

Response: The computed decrease of

0.11 inches is the calculated maximum

decrease in the inner diameter of the

overpack shell due to a tipover accident.

The 0.11 inches decrease in the inner

diameter of the overpack shell is not

computed by computer simulation.

Rather, it is computed by using a

standard text book equation for

deformation calculation. The

deformation due to tipover is expected

to be small. This calculation has been

evaluated by the NRC staff and found to

be acceptable.

Comment E.11: One commenter asked

if the base under the pads would be the

same at all sites and asked what is under

the pad. The commenter is concerned

that the pad evaluation has not received

adequate attention because it is a crucial

part of the tipover and drop evaluation.

Response: Each user is required to

meet the site parameters in CoC,

Appendix B, Section 3.4 that include

specific requirements for the pad. Site

characteristics will be investigated by

each cask user and addressed in the

cask user’s 10 CFR 72.212 evaluation.

The pad is a site-specific issue. Sitespecific issues are beyond the scope of

this rule that is focused solely on

whether to add the HI–STORM 100 cask

system to the list of approved casks.

Comment E.12: One commenter asked

why there were two different weights

for the transfer cask.

Response: As discussed in Section

1.2.1.2.2 of the SAR, the 100-ton transfer

cask weighs less than the 125-ton

transfer cask because it has a reduced

thickness in lead and water. The 100-

ton transfer cask is designed for

facilities not capable of handling the

heavier 125-ton transfer cask.

Comment E.13: One commenter asked

why the bottom pool lid supported the

weight of a loaded MPC plus water.

Response: During lifting of the

transfer cask from the fuel pool there is

water in the MPC and the annulus.

Therefore, the structural evaluation of

the bottom pool lid of the transfer cask

must consider all the applicable weights

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25251

supported by the pool lid, including the

water.

Comment E.14: One commenter stated

that the cask should be up on something

to air out the area under the cask to

prevent rusting. The commenter

questioned if the baseplate rusted if that

could cause the cask to tipover or lean.

The commenter is concerned that if the

canister ended up leaning against the

inner liner of the concrete shell, it

would cause blockage of the venting

annulus and create a hotspot in the

concrete.

Response: The NRC disagrees with

this comment. The baseplate is coated

with an epoxy type coating to prevent

corrosion. Some rusting may occur at

scratches in the coating. However, even

a postulated extreme case, assuming no

coating present, would not result in

sufficient corrosion to cause an amount

of leaning that would be significant.

Comment E.15: One commenter asked

if there is any leeway for the pressure

in the concrete encasement between the

two carbon steel outer and inner liners,

if the concrete had room to move, and

if the concrete could split the outer

carbon steel encasing it, particularly at

the welds.

Response: The coefficient of thermal

expansion of steel is only slightly

greater than that of concrete, and the

thermal gradient through the overpack

wall, experienced during the extreme

temperature criteria, was calculated to

be approximately 40°F. This

temperature difference and thermal

coefficient of expansions do not cause

the inner steel to apply significant force

to the concrete in the overpack. The

outer steel shell expands somewhat

more than the concrete; therefore, the

concrete has room for expansion and

exerts no force on the outer steel plates.

Comment E.16: One commenter asked

what a bottom pool lid is and how it is

replaced by the heavier shielded

transfer lid and if it has been tested.

Response: The bottom pool lid is

described in Section 1.2.1.2.2 of the

SAR. The lids are interchanged with a

transfer slide device as described in

Section 8.1.1 of the SAR. The NRC did

not require test results for lid changing

operations. The NRC found the pool and

transfer lid design to be acceptable for

the HI–STORM 100 system.

Comment E.17: One commenter asked

if the 17.000 inches of concrete for the

overpack baseplate was a typo and if the

number of significant figures was

correct.

Response: The value in the SER has

been revised to state 17.0 inches that

reflects the thickness assumed in the

shielding analysis.

Comment E.18: One commenter stated

that the configuration discussion in

Section 6.4.1 of the SER is not clear

because the HI–STAR doesn’t have a

transfer cask.

Response: As stated in SER section

6.4.1, the HI–STORM system has a

transfer cask, not the HI–STAR system.

The transfer cask, the HI–STORM

overpack, and the HI–STAR overpack

are constructed of different materials.

The effectiveness of these materials to

reflect neutrons affects the criticality

safety of the system; therefore, each was

explicitly evaluated. The other

parameters affecting the criticality safety

of the HI–STORM system, including the

transfer cask, are identical to the HI–

STAR system.

Comment E.19: One commenter asked

if the closure ring was a ring or a lid.

Response: The closure ring is a ring.

In the MPC, the lid and the closure ring

are two different components.

Comment E.20: One commenter asked

how many rings are included in the

design.

Response: There is one closure ring

included in the design.

Comment E.21: One commenter asked

why voids in the installation of the lead

shield are only minimized instead of

being disallowed completely, if the

shield was composed of lead bricks or

poured, and which was more prone to

voids. The commenter asked if lead

bricks could be used and then have lead

poured into the cracks between the

bricks, and how the lead shield is

installed.

Response: The HI–STORM 100 must

be fabricated and tested in accordance

with the drawings specified in the SAR

and under a quality assurance program

that meets the requirements of 10 CFR

Part 72, Subpart G. The proper

fabrication of the lead shield, including

potential voids, will be evaluated under

this quality assurance program. As

discussed in Section 9.1.5.2 of the SAR,

effectiveness of the lead pours are

verified during fabrication by

performing gamma scanning on all

accessible surfaces of the transfer cask

in the lead-pour regions. Installation of

the lead shields is discussed in Section

9.1.5.1 of the SAR. The SAR specifies

the use of poured lead and does not

allow the installation of lead bricks.

Comment E.22: One commenter asked

what the relief valve was and what type

of maintenance it received.

Response: A relief valve is a

mechanical device that opens when

pressure inside a system exceeds the

actuation pressure of the valve (pressure

that will open the valve). Relief valves

are common pressure limiting devices.

Relief valves are placed on water heaters

in homes to ensure that the water pipes

in a house will not fail due to excessive

pressure. Similarly, relief valves are

attached to the radiator in a car to

ensure that the coolant hoses do not

burst from excessive pressurization of

the engine coolant system. Maintenance

of the relief valves are discussed in SAR

Section 9.2.4. The relief valves are

calibrated annually to ensure that their

pressure relief setting is correct or they

are replaced with factory-set relief

valves.

Comment E.23: One commenter asked

if there were holes in the shield jacket

to add and drain things and indicated

that holes would be a potential sabotage

threat for someone to drain the jacket or

add something dangerous to the water.

Response: There are drain holes in the

water jacket end plate. The 125-ton HI–

TRAC has two 11⁄2-inch drain holes and

the 100-ton HI–TRAC has four 3⁄4-inch

drain holes. The resultant dose rate for

an assumed loss of the water jacket is

addressed in Section 5.1.2 of the SAR.

The analysis indicates that the off-site

dose at 100 meters will be below the 5

rem accident limit in 10 CFR 72.106. In

addition, NRC regulations in 10 CFR

Part 72 have established physical

protection and security requirements for

an ISFSI located in the owner-controlled

area of a licensed power reactor site.

Comment E.24: One commenter stated

that Conditions 1a and 1b of the CoC

should both state that the cask system

has two transfer casks.

Response: The NRC disagrees with the

comment. Condition 1b of the certificate

of compliance specifies that there are

two types of transfer cask options: the

125-ton HI–TRAC and the 100-ton HI–

TRAC. It is not necessary to repeat that

information in Condition 1a.

Comment E.25: One commenter stated

that there should be a drawing of the

damaged fuel container in the CoC

because the structure is not explained.

Response: The NRC disagrees with

this comment. A drawing of the

damaged fuel container is included in

Chapter 1 of the SAR and is available to

the public. This level of detail is not

necessary in the CoC.

Comment E.26: One commenter asked

what the screens are made of, how the

screens are attached, if the screens can

deteriorate or come loose over time, and

what happens if the screens fall out.

Response: As shown on the drawings

in Chapter 1 of the SAR, the damaged

fuel container, including the screen, is

constructed of stainless steel. The

damaged fuel container is an additional

structural component that will make the

MPC fuel basket even stronger. The

screen is placed between two steel

plates welded together with a 0.06 inch,

25252 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

continuously 360 degree, all around

fillet weld. It is not considered credible

for the screens to fall off or fail.

However, if a screen failed, there would

be no release of radioactive material

because the MPC is sealed. Small

amounts of loose debris in the MPC

have been considered during unloading

operations, as described in SAR Section

8.3.4.

Comment E.27: One commenter stated

that damaged fuel and intact fuel should

not be placed in the same cask because

it can cause potential problems in

unloading.

Response: The NRC disagrees with the

comment. Damaged fuel can be stored

safely with undamaged fuel. If damaged

fuel is stored with undamaged fuel, then

CoC, Appendix B, Section 2.1.1.c

requires all fuel assemblies in the cask

to meet the more restrictive heat

generation requirements for the

damaged fuel. Additionally, damaged

fuel must be loaded into damaged fuel

containers to enable safe handling

during cask loading and unloading

operations.

Comment E.28: One commenter asked

what the basis is for putting the hotter

fuel in the center of the cask. The

commenter also asked if the doses

would be accurate if the lower dose fuel

is placed at the periphery positions. The

commenter stated that it would be better

to have a more even heat and dose

distribution in the MPC and asked if

dose was more important than the heat.

Response: The design of the HI–

STORM cask considered both the

thermal and radiological effects of the

fuel. If one assumes the same

enrichment and burnup (time that the

fuel was left in the reactor to produce

power), the fuel that is left longer to

decay in the spent fuel pool (e.g.,

‘‘cooler fuel’’) will generate less heat

and high-energy radiation than the fuel

that is removed sooner from the pool

(e.g., ‘‘hotter fuel’’). For the method

used in the HI–STORM design, cooler

fuel assemblies are stored on the

periphery of the cask for two reasons.

First, the ‘‘cooler fuel’’ assemblies have

lower allowable peak clad temperature

limits and the temperature of the

assemblies on the periphery is cooler.

Second, storing the ‘‘cooler fuel’’ on the

periphery of the MPC provides some

additional radiological shielding from

the hotter fuel assemblies in the center.

Comment E.29: One commenter asked

if BPRAs and thimble plugs could be

stored in the cask. The commenter

stated that they should not be because

they add weight.

Response: BPRAs and thimble plugs

have not been analyzed for this cask

system and, therefore, are not

authorized for storage in the HI–STORM

100 at this time.

Comment E.30: One commenter asked

what the cask transfer station is and

whether it had been designed yet. The

commenter asked if it is constructed of

reinforced concrete. The commenter

stated that more explanation was

necessary and that a drawing should be

included. The commenter asked how

and why an impact limiter is used. The

commenter asked what the basis is and

if an evaluation had been completed.

The commenter was concerned with the

use of terms ‘‘if’’ and ‘‘shall be

designed’’ because this implies the CTF

hasn’t been designed. The design should

have specific criteria.

Response: The term ‘‘cask transfer

station’’ in CoC, Appendix B, Section

3.5.1, is a typographical error and has

been corrected to ‘‘cask transfer

facility.’’ A cask transfer facility (CTF) is

a facility used for transferring the MPC

between the transfer cask and the

overpack. The CTF does not include 10

CFR Part 50 controlled structures such

as the fuel handling building or reactor

building. The NRC disagrees that a

drawing of the CTF or more design

details are necessary. The HI–STORM

100 Cask System is approved for use

under the general license provisions of

10 CFR Part 72. Therefore, the cask may

be used in any nuclear power reactor

site licensed under 10 CFR Part 50,

provided that the site parameters are

enveloped by the cask design bases. The

specific design and operation of the CTF

will be dictated by site-specific needs.

Because of the varied needs of each

reactor site, the NRC found it

impractical and unnecessary to review

and approve a specific CTF design,

including the specific materials of

construction. The NRC reviewed and

approved the criteria for the design,

construction, and operation of the CTF.

These criteria are specified in CoC,

Appendix B, Section 3.5, and SAR

Section 2.3.3.1.

The impact limiter is a possible CTF

design feature whose function would be

a defense-in-depth measure because the

lifting equipment used in the CTF must

be designed to preclude a drop. As

discussed in the response to comment

J.14, the specific requirement for an

impact limiter has been eliminated

because other methods may be available

to prevent a canister breach in case of

a canister drop during transfer

operations.

Comment E.31: One commenter stated

that we should clearly state what the

CTF is and make sure that every detail

of the procedure is carefully analyzed

because it is vague.

Response: The CTF is defined in SAR

Section 2.3.3.1 and in CoC, Appendix B,

Section 1.0. Detailed design and

operational requirements for the CTF

are also specified in SAR Section

2.3.3.1, as well as in CoC, Appendix B,

Section 3.5. Under the provisions of 10

CFR 72.212 and CoC Condition 2, each

licensee that elects to use a CTF must

develop written procedures for

operating the CTF. These procedures are

subject to NRC review during

inspection. As required by TS 5.2.h, the

licensee must conduct a dry run training

exercise, prior to first use of a CTF, to

demonstrate that the procedures can be

conducted safely and successfully.

Comment E.32: One commenter

recommended that Design Drawing

1495, Sheets 4 and 6 and Design

Drawing BM–1575, Sheet 2 be revised.

Response: The NRC agrees with the

comment. These changes correct

drafting errors or provide a level of

flexibility that is acceptable to the NRC

staff. The SAR drawings have been

revised accordingly.

F. Welds

Comment F.1: One commenter asked

how use of the trunnions puts stress on

the weld at the water jacket.

Response: Use of the pocket trunnion

does not put any stress on the water

jacket. The seal weld between the

pocket trunnion and water jacket shell

is for retaining the water inside the

water jacket. The pocket trunnion is

attached to the outer transfer cask shell

by full penetration welds all the way

around the trunnion. When the pocket

trunnion is used, the force is transferred

to this weld and not to the seal weld on

the water jacket. The other type of

trunnion on the transfer cask is the

lifting trunnion. The lifting trunnion is

not connected to the water jacket and,

therefore, puts no stress on the water

jacket.

Comment F.2: One commenter asked

how the welds are checked to be

leakproof and whether water can enter

the trunnion.

Response: All the structural welds

have to be examined and inspected

according to the applicable ASME code.

The welded joint is an integral part of

the structure and is leak proof. Because

the trunnions are made of solid steel,

water cannot leak into them.

Comment F.3: One commenter stated

that the lid and closure ring of the MPC

should be full penetration welds and

should be ultrasonic tested (UT) as this

is the basis for qualification as a

redundant seal.

Response: The NRC disagrees with

this comment. Full penetration welds

are unnecessary from the structural and

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25253

containment boundary requirements of

the design. Employing unnecessarily

heavy welds leads to fabrication

problems such as excessive warpage. UT

of heavy section, full or partial

penetration, austenitic stainless steel

welds to ASME Code acceptance criteria

is not feasible with current technology.

The redundant seal concept is based

upon the use of two welds forming the

leak barrier, not the inspection method.

With redundant welds, one weld could

leak and the second still provide leak

tight integrity.

Comment F.4: One commenter stated

that just because there is no known

plausible, long-term degradation

mechanism to cause seal welds to fail

doesn’t mean that the welds won’t fail.

Response: The NRC disagrees with

this comment. The NRC staff has

examined the plausible mechanisms

that would cause failure of the seal

welds and has determined that those

mechanisms are inoperative under

normal service and design accident

conditions for HI–STORM. This gives

the NRC staff reasonable assurance that

the welds will not fail under design

basis normal, off-normal, and accident

conditions.

Comment F.5: One commenter asked

how lid welds are removed in unloading

and stated that the procedures should be

in the documents before the casks are

loaded.

Response: The NRC disagrees with

this comment. Welded cask lids may be

removed by one of several methods. The

method of removal and the detailed

procedures (as opposed to the general

procedures of the SAR) are the

responsibility of the ISFSI licensee,

subject to NRC review and inspection.

The TSs require ISFSI licensees to

perform lid removal method

demonstrations on full-size mock-ups as

part of their pre-operational testing and

training exercises. The NRC staff has

reviewed and inspected several methods

and their associated procedures.

Inclusion of such detailed procedures in

the SAR is unnecessary.

Comment F.6: One commenter asked

what happens if the water jacket welds

leak water.

Response: The resultant dose rate for

an assumed loss of the water jacket is

addressed in Section 5.1.2 of the SAR.

The analysis indicates that the off-site

dose at 100 meters will be below the 5

rem accident limit in 10 CFR 72.106.

Comment F.7: One commenter stated

that the penetrant test (PT) is

unacceptable, that the criteria for layers

and time are ‘‘wishy washy,’’ and that

PT tests should not be allowed.

Response: The NRC disagrees with

this comment. The commenter has not

specified why PT is unacceptable or

why it should not be allowed. PT is a

Code accepted examination method.

The progressive PT technique is used

and accepted in the nuclear industry

when a volumetric examination by

means such as UT is impractical. UT is

unsuitable for heavy section austenitic

stainless steel welds.

The basis for the structural lid weld

examination methods is documented in

the NRC’s Interim Staff Guidance-4,

Revision 1, that allows the use of a

multi-layer (i.e., progressive) PT

examination in lieu of a volumetric

examination.

Comment F.8: One commenter stated

that welds need to be checked carefully.

Response: The NRC agrees with the

comment. Welds are important which is

why they are examined and inspected

according to the applicable ASME code.

Comment F.9: One commenter stated

that the leak testing procedure used to

demonstrate MPC closure cannot be

performed as described and that

performance of the test is not generally

consistent with ANSI N14.5–1997,

‘‘Leakage Tests on Packages for

Shipment.’’ Consequently, containment

of the radioactive material to the stated

criteria cannot be demonstrated. The

principal reason provided by the

commenter is that the nominal

concentration of helium in air is 5 parts

per million. This atmospheric

concentration masks any leakage from

the MPC using the specified test

conditions. In addition, the commenter

noted that there is no direct reference to

definitions, equations, formula,

methodology, or criteria of the standard

in the text. The commenter further

noted that when terminology from the

standard is given, it is (for the most part)

used incorrectly.

Response: The NRC disagrees with the

comment. These welds are multipass

stainless steel welds that are dye

penetrant examined multiple times

during the weld process. The multiple

dye penetrant examinations provide

reasonable assurance of high integrity

welds that will retain the inert gas and

prevent leakage of radioactive material

into the environment. The leakage

testing of these welds provides

additional insight into the leak-tightness

of these welds.

The NRC staff has reasonable

assurance that the leakage test

procedure outlined in SAR Section 8.1

can be performed as described provided

that appropriate equipment is used and

the leak test method is properly

qualified. The leakage testing for the lid

is to be performed with a sniffer type

probe; however, the test method for the

port and drain covers is not specified in

the SAR. There are test methods

discussed in Appendix A of ANSI N–

14.5 that could be used to perform the

port and drain cover leakage testing.

Detailed procedures are developed by

the user who is responsible for ensuring

that the TS limits are met and therefore,

confinement is adequately maintained.

The leakage testing will require a

demonstration that the detector can

identify an appropriate calibrated leak

in the presence of background helium.

Although sniffer probes detect discrete

leaks rather than an integrated leakage

rate, the typical sniffer probe sensitivity

of 10¥8 provides reasonable assurance

that the TS leakage rate limit of 5×10¥6

will not be exceeded.

As stated in the SAR, the leak testing

will be performed in accordance with

ANSI N14.5. It is not necessary to

include any more level of detail in the

SAR; therefore, no change to the SAR is

necessary. Appropriate detail will be

included in the site procedures.

The SAR has been changed to use

terminology consistent with the TS and

the 1997 revision of ANSI N14.5. The

terminology was changed from std cc/s

to atm cc/s. SAR section 7.3.3 justifies

the use of the units atm-cc/sec. Also, the

SAR was revised to delete the

sensitivity of the detector. The

sensitivity will be addressed in the

detailed site procedures.

G. Structural

Comment G.1: One commenter stated

that all of the accident level events and

conditions listed in the SER,

particularly a transfer cask handling

accident or sabotage, should be

evaluated for structural analysis in a

jamming condition on top of the

overpack.

Response: All design basis normal,

off-normal, and accident events have

been evaluated in the structural

analyses and are discussed in Section 3

of the SER. This includes an evaluation

of the transfer cask under a 42-inch

horizontal drop during transfer

operations. A horizontal drop from a

greater height is not considered because

the horizontal lifting height limit for the

transfer cask is 42 inches. The

evaluation shows that the HI–TRAC

meets all structural requirements and

there is no adverse effect on the

confinement, thermal, or subcriticality

performance of the contained MPC.

As discussed in the response to C.15,

vertical drop of a transfer cask is not

considered credible because vertical

lifting of a loaded transfer cask must be

performed with structures and

components designed to prevent a cask

drop. Also, as discussed in the response

to E.5, jamming is not considered to be

25254 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

credible because of the design of the HI–

STORM system. The threat of sabotage

is beyond the scope of this rule and is

discussed in the response to C.8.

Comment G.2: One commenter stated

that bending of the web and pushing of

the flanges possibly accompanied by

some local weld failures sounded

feasible and may not result in limited

deformation as assumed. The

commenter asked if a full size cask had

been tested in a drop or tipover.

Response: The channels attached to

the inner shell of the overpack are not

classified as important-to-safety and

serve no structural purpose.

Deformation of the channels, whether

limited or complete collapse, does not

affect retrievability of the MPC. On the

contrary, the deformation of these

channels due to a tipover accident

absorbs energy which reduces the

deceleration loadings to the MPC and

provides a greater opening in the

overpack during retrieval. NRC

regulations do not require full size

testing of casks. The applicant can

choose the method of analysis.

Computer analyses have been performed

to determine the responses of a cask in

drop and tipover accidents.

Comment G.3: One commenter

questioned how the structural analysis

conducted for the 125-ton HI–TRAC

transfer cask could bound the 100-ton

version and indicated that the 100-ton

version needs its own analysis.

Response: All the structural analyses

and evaluations of the 125-ton transfer

cask were repeated for the 100-ton

transfer cask. However, the analytical

results of the 125-ton transfer cask are

greater than that of the 100-ton transfer

cask. Therefore, the structural analysis

of the 125-ton transfer cask bounds the

100-ton transfer cask.

Comment G.4: One commenter asked

what would be the consequences of the

deformation of the outer shell and lead

and water jacket from a missile,

particularly if the transfer cask was on

top of the concrete shell.

Response: The HI–TRAC transfer cask

is always held by the handling system

while in a vertical orientation

completely outside of the fuel building.

Therefore, considerations of instability

due to a tornado missile strike are not

included in the evaluation. However, a

structural evaluation of the damage to

the HI–TRAC transfer cask from an

intermediate missile strike and a large

missile strike is performed. The

evaluation shows that the outer shell

and the water jacket would not

experience any plastic deformation and

will not adversely affect the

retrievability of the MPC.

H. Materials

Comment H.1: One commenter

questioned why carbon steel was used

for the inner and outer plate instead of

stainless steel because of the concern

over corrosion. The commenter also

asked if the carbon steel was coated.

Response: The materials used in the

fabrication of the cask are described in

Chapters 1 and 3 of the SAR and

discussed in Section 3.3 of the NRC

SER. These materials have been found

acceptable because they meet the

requirements for their respective

applications in the cask system. They

are suitable for the expected loading and

storage in wet and dry environments,

including corrosion and galvanic effects.

There is no requirement for designers to

select materials from a given class, e.g.

stainless steels.

The carbon steel used in the overpack

is protected from corrosion by an

industrial epoxy coating commonly

used for the protection of steel.

Comment H.2: One commenter stated

that one alloy should be specified for

cask fabrication instead of allowing a

choice because if later problems

develop, there are fewer variables.

Response: The NRC disagrees with the

comment. The materials used in casks

are selected on the basis of the needed

properties. Allowing a choice of more

than one material or alloy for fabrication

is acceptable provided that each of the

options has the appropriate properties.

The materials chosen for use in the

Holtec HI–STORM 100 design have a

long history of favorable performance in

the nuclear industry.

Comment H.3: One commenter

questioned why plain concrete is not

included in NUREG–1536 and why an

exemption was being given to allow

plain concrete since reinforced concrete

is stronger.

Response: No exemption was given to

allow plain concrete to be used for

structural components. The plain

concrete in the HI–STORM 100

overpack is for shielding only and is not

a structural component of the overpack.

The reinforced concrete included in

NUREG–1536 is for concrete structures

(concrete components that provide

structural strength) only. The HI–

STORM 100 overpack is a welded steel

structure, not a concrete structure.

Comment H.4: One commenter asked

if the NRC has reviewed the

manufacturers direction for the

carboline 890 and thermaline 450

coatings. The commenter asked how the

coatings are used and applied, and if

they will wash or flake off in pool water,

making the water cloudy.

Response: The NRC staff has reviewed

the manufacturer’s technical

information for the coatings mentioned.

Both coatings are standard coatings

employed in industry for immersion

service and are applied using common

industry tools and techniques. No

performance problems would be

expected during intended service.

Comment H.5: One commenter asked

if the carbon steel caused reactions that

could create loading or unloading

problems such as reaction products

clogging venting or draining equipment

with crud or flakes or making the pool

water cloudy.

Response: Carbon steel exposed to the

cask loading environment produces very

fine particulates that do not clog

equipment. Turbidity that may arise

from corrosion of uncoated carbon steel

can be controlled with appropriate

water treatment equipment.

Comment H.6: One commenter asked

if temperature or coatings on the

channel could affect the fit. The

commenter also asked if flaking of the

coating could clog a channel slide or if

corrosion in the channels could cause

problems in unloading.

Response: The effects of temperature

on the channels have been calculated

and do not affect the fit. Each coat of the

epoxy paint applied to the exposed

surfaces of the inner components of the

overpack is, at maximum, 0.008 inches

thick. Two coats result in a maximum

diametral reduction in inside diameter

of 0.032 inches. This reduction will not

affect the fit. Both the interior and the

exterior of the channels are coated to

prevent corrosion.

Comment H.7: One commenter asked

if aging was factored into the analysis of

the pad and stated that the specific site

should be evaluated for a full cask array.

Response: Concrete is resistant to

environmental conditions, including air

pollution and moisture. Therefore, the

NRC staff expects no significant

degradation of the pad during the

licensed lifetime of the ISFSI facility.

Each proposed site is subject to a

specific evaluation to ensure that the

design parameters satisfy site-specific

conditions. In addition, cask users are

responsible for inspecting and

maintaining the pad, and for ensuring

that significant degradation is not

occurring over time.

Comment H.8: One commenter asked

what the condition of the concrete is

right under the shell and expressed

concern that the concrete could crack

where nobody would see damage

needing repair.

Response: As discussed in the

response to E.3, the main function of the

concrete encased between the steel

shells in the HI–STORM 100 overpack

is shielding. The structural strength of

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25255

the HI–STORM 100 overpack is

provided by the inner and outer carbon

steel shells. Cracking of the concrete

would not have a significant impact on

the cask’s ability to meet the regulatory

dose limits. There is no credible

mechanism for the concrete to undergo

any significant damage. Thus,

inspection of the concrete is not

necessary.

Comment H.9: One commenter asked

if concrete expanded or released water

or gas when it is superheated.

Response: Concrete contains some

traces of free water. If the water is

heated, it will evaporate. Concrete will

expand upon heating and contract upon

cooling. The amount is governed by the

temperature. These expansions/

contractions are reversible and not

permanent. There are no significant

effects of expansions/contractions that

would occur even if the temperature

went considerably beyond the design

temperature parameters.

Comment H.10: One commenter asked

how the bottom face affects the

supporting surfaces (heat, radiation,

weight, stress, pressure etc.).

Response: As listed in Table 4.4.9 of

the SAR, the temperature of the bottom

lid plate at normal conditions is 183°F.

That temperature will not have an

adverse effect on the concrete. Radiation

will have minimal impact on the

concrete pad due to the shielding

provided by the pedestal. The weight,

stress, and pressure from the cask

bottom have no adverse effect upon the

pedestal or slab because they are

specifically designed to support all the

loads due to the casks.

Comment H.11: One commenter asked

how the gas and liquid media that

escapes from the damaged fuel

container interacts with other materials

in the MPC and if they can cause

problems.

Response: The materials of the cask

have been selected to be compatible

with any constituent or reaction product

of the fuel.

I. Thermal

Comment I.1: One commenter asked if

hot spots in the cladding could cause

lead to sag in the transfer cask if the

inner canister is in place and the

temperature is close to the boiling point

of the water pack.

Response: Hot spots in the cladding

would not result in sagging or melting

of the lead. The bounding calculation

performed by Holtec assumed all the

fuel assemblies were at the design basis

limit (hottest assemblies). The bounding

rod cladding temperature occurs at the

center of the MPC and does not have a

direct impact on the lead. The

assemblies on the periphery of the MPC

are significantly cooler because they are

located near the cooler surface of the

MPC. Table 11.2.8 in the SAR provides

the results from a calculation that

assumes no water in the water jacket.

These results bound the impact of

boiling in the water pack. Based on

those results, it can be concluded that

the lead temperature remains well

below the melting temperature.

Comment I.2: One commenter asked

what happens if the water in the transfer

pack boils and the steam pressure builds

up, and stated that this situation should

be evaluated.

Response: As the pressure builds up,

the pressure is relieved through a safety

valve. As water is removed through the

safety valve, the temperature of the

water remains at the saturation

temperature. The case of water boiling

in the HI–TRAC water jacket is bounded

by the event that assumed no water in

the water jacket. This event leads to a

temperature in the water jacket that is

higher than the saturation temperature

of the water. The impact of loss of water

in the water jacket is summarized in

Table 11.2.8 of the SAR.

Comment I.3: One commenter asked

how, during normal conditions, the

temperature of the outer shell could be

higher than the temperature of the

concrete because the carbon steel would

breathe less than the concrete, causing

the heat to be retained in the concrete.

The commenter also asked how the

temperature of the concrete could be

measured since it is encased in the

carbon steel.

Response: The question raised by the

commenter is not clear. The temperature

of the concrete is higher than the

temperature of the outer shell under

normal conditions. Reviewing Table

4.4.9 in the SAR, the temperature at the

overpack outer shell is not higher than

the concrete cross sectional average

temperature. The temperature

distribution through the overpack under

normal conditions is listed in Table

4.4.9 of the SAR (e.g., 149 °F for the

concrete and 131 °F for the outer shell).

With regard to the question of

measuring the temperature of the

concrete, the applicant does not

measure the temperature of the

concrete. Bounding calculations are

used to assure that the concrete

temperature limits will not be exceeded.

Comment I.4: One commenter asked

how the pad reacts to the bottom plate

of a cask from a temperature differential

standpoint. The commenter asked if the

pad would crack and sink under each

cask and form a concave area that could

then collect moisture. The commenter

further asked if the collected moisture

could boil and if the moisture could

cause the bottom plate to rust.

Response: The heat transfer between

the bottom plate of the overpack and the

concrete pad is modeled in the thermal

computer code for the HI–STORM cask

system. As listed in Table 4.4.9 of the

SAR, the temperature of the bottom lid

plate at normal condition is 183 °F. That

energy is transmitted to the concrete

pad down to the ground, which is at the

normal soil annual average temperature

of 77 °F. Therefore, the concrete will not

experience temperatures above boiling

(no superheating will occur). In the

winter, the concrete will not reach

freezing temperatures below the cask

because it generates heat. If the concrete

reaches or exceeds boiling or freezing

temperatures, there is no detrimental

effect on the strength or condition of the

concrete. The pad is specifically

designed to support the weight of the

casks without any cracking or sinking of

the pad. The bottom plate of the cask is

stainless steel and will not rust.

Comment I.5: One commenter

questioned the basis and validity of

simulating the heat effect of adjacent

casks radiating heat back to an interior

cask and if an analysis of the real

situation had been conducted.

Response: The impact of radiation

heat transfer from neighboring casks

was calculated in the HI–STORM

thermal evaluations. The method used

by the applicant was to assume that all

of the radiated heat is reflected back to

the cask. This modeling assumption is

equivalent to assuming that the cask

was totally encircled by other casks. In

reality, less heat will be radiated back to

the cask; therefore, the calculations

bounded the effects of neighboring

casks. The impact of neighboring casks

was shown to be minimal. The NRC

staff does not require validation of the

analytic method with actual

experimental data.

Comment I.6: One commenter asked

why the analysis assumed that the soil

below the overpack was at a constant

temperature because the casks could

cause hot spots.

Response: The analyses did model the

hot spots below the overpack. The

computer simulation of the overpack

modeled the concrete pad that the

overpack is placed on and the

temperature of the soil below the

concrete pad. The soil is one of several

paths for heat to leave the cask. The

most significant path for heat

dissipation is through the air passage

between the MPC and the overpack. The

applicant used the highest annual

average soil temperature found in the

USA. The purpose for using the highest

average temperature for the soil and air

25256 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

in the thermal analyses is to

demonstrate fuel retrievability and that

the cladding is protected during storage

against degradation that leads to gross

ruptures (10 CFR Part 72.122). One

acceptable method for demonstrating

that the cladding will not undergo gross

rupture is to place a limit on the

allowable cladding temperature such

that reasonable assurance exists that the

cladding will not significantly degrade.

A report by the Pacific Northwest

National Laboratory, PNL–6189, dated

May 1987, provides one acceptable

approach for establishing a temperature

limit. The PNL method is conservative

when compared to the maximum

allowable degradation permitted in Part

72 of the regulations. This method, in

conjunction with the maximum annual

average temperature, solar heating (e.g.,

insulation), analytic assumptions, etc.,

provide reasonable assurance that the

requirements of Part 72 will be met.

Comment I.7: One commenter asked

why an exception was allowed for

exceeding the short term temperature

limit for the fire accident scenario and

stated that an exception should not be

allowed.

Response: The American Concrete

Institute (ACI) establishes temperature

criteria for concrete. One, but not the

only, acceptable demonstration that the

concrete overpack will maintain its

intended function is to meet the

temperature criteria in ACI 349.

However, as stated in the NRC staff’s

Standard Review Plan (NUREG–1536),

‘‘a small amount of exterior concrete

spalling may result from a fire, the

application of fire suppression water,

rain on heated surfaces or other hightemperature condition. The damage

from these events is readily detectable,

and appropriate recovery or corrective

measures may be presumed. Therefore,

the loss of such a small amount of

shielding material is not expected to

cause a storage system to exceed the

regulatory requirements in 10 CFR

72.106 and, therefore, need not be

estimated or evaluated in the SAR. The

NRC accepts that concrete temperatures

may exceed the temperature criteria of

ACI 349 for accidents if the

temperatures result from a fire.’’ The

Holtec analysis demonstrated that the

amount of concrete that exceeds the ACI

temperature limit is very limited and

would not pose a significant safety

hazard.

Comment I.8: One commenter asked

for the basis of using an average

temperature of the gas in the gap and

plenum of the limiting rod and

questioned the validity of the

assumption.

Response: The purpose for evaluating

the average of the gas temperature in the

fuel rod is to calculate the pressure

within the fuel rod. The computer code

used in the analysis calculates the

temperature profile of the fuel rod, but

does not calculate the corresponding

pressure for that rod. To calculate the

pressure, the average temperature of the

gas is calculated and from the ideal gas

law, the corresponding pressure is

established.

Comment I.9: One commenter asked

what is in the water used for forced

water circulation under wet transfer of

the fuel from the spent fuel pool to the

location for vacuum drying and if the

water could chemically affect other

materials in the cavity. The commenter

asked how fast the water flows, if steam

could be formed, and if the water could

physically affect other materials in the

cavity, movement of rods, flaking of

paint, etc.

Response: The licensee can either use

demineralized water or water from the

spent fuel pool. Neither demineralized

nor spent fuel pool water would

adversely interact with the system. The

flow rate of the water is based on the

heat output of the fuel assemblies and

is a site-specific issue.

Comment I.10: One commenter asked

what the water chiller is used for and

what material is used as the chilling

medium.

Response: The water chiller is used as

the heat sink for cooling the helium

inside the MPC to below 200°F. The

type of water chiller used is a sitespecific issue and not part of this

rulemaking activity.

Comment I.11: One commenter asked

for specific criteria that defines

clearance around the cask for cooling

purposes instead of stating a reasonable

amount. The commenter also asked how

close other heat sources may be located

and what is considered to be a

significant heat source.

Response: The actions identified by

the commenter are only valid when a

breakdown occurs in the helium coolers

(LCO 3.1.3). Section B3.1.3 of the

technical specification bases states that

‘‘if the TRANSFR CASK is located in a

relatively open area such as a typical

refuel floor, no additional actions are

necessary.’’ However, a licensee may

elect to perform the cooling with the

cask located in a pit or vault. This is a

site-specific activity. The bases identify

three acceptable options for ensuring

adequate heat transfer for the

TRANSFER CASK. The user may

develop other alternatives on a sitespecific basis, considering actual fuel

loading and decay heat generation

within the cask. One of the options is to

fill the annulus between the MPC and

the TRANSFER CASK with water. The

second option is to remove the

TRANSFER CASK from the pit or vault

and place it in an open area such as the

refueling floor with a reasonable amount

of clearance around the cask and not

near a significant source of heat. The

third option is to supply nominally

1000 SCFM of ambient air to the space

inside the confined space (e.g., pit or

vault). With respect to defining an

acceptable distance, the licensee could

use the analyzed event of 15 feet centerto-center storage spacing that

corresponds to a four foot clearance.

Smaller clearances would also be

acceptable, given the heat load rating of

the cask and ambient conditions. With

regard to defining a significant heat

source, this is a site specific

consideration. For example, if the plant

is using a bank of radiant heaters near

the cask, then an evaluation needs to be

performed to ensure that those heaters

pose no adverse impact on the cask.

These options are only guidelines to an

LCO that a user would have to consider.

Comment I.12: One commenter asked

how cool air was provided to the space

inside the vault at the bottom of the

overpack. The commenter stated that

this needs to be planned out ahead of

time for ALARA considerations and

equipment availability.

Response: This is a site-specific issue

and not part of this rulemaking activity.

The NRC staff agrees with the comment

that the user needs to plan this activity

considering ALARA and equipment

availability.

Comment I.13: One commenter stated

that fuel should be adequately cooled

before it goes into the transfer cask.

Response: The NRC agrees with the

comment that adequate planning is

needed when performing cask cooldown

and reflooding. The purpose of the

analyses performed in the SAR is to

maintain the integrity of the fuel. The

requirements on the burnup and

minimum cooling time serve that

purpose.

Comment I.14: One commenter asked

if the temperature of the helium

accurately reflects the internal

temperature of the MPC and stated that

this should be tested.

Response: The exit temperature of the

helium reflects the conditions of the

fuel rods. After the helium temperature

is reduced below 200°F, the bulk of the

fuel will be at low temperatures,

minimizing the potential for excessive

steaming. Reflooding of a canister has

been demonstrated without pre-cooling

the helium. No additional tests are

needed.

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25257

Comment I.15: One commenter

objected to the addition of ethylene

glycol solution to the demineralized

water in the water jacket to prevent

freezing and asked where this had been

tested. The commenter also asked why

the antifreeze was used, how the

solution would mix, how the NRC

knows it will work, what types of effects

it could have on the inside of the water

jacket and the channel walls, if it would

add weight, and how it is added to the

water if the jacket is welded shut.

Response: Ethylene Glycol is the

chemical name for ordinary antifreeze.

Adding antifreeze to the water jacket,

located on the outside of the HI–TRAC

transfer cask, is an option if the user

elects to move a loaded MPC in cold

weather, down to 0°F. The use of

antifreeze in the water jacket does not

add appreciable weight to the HI–TRAC

cask. Although the water jacket is a

welded system, openings are designed

to add and remove water from the water

jacket. Antifreeze has been used in

many applications to keep water from

freezing. The NRC staff believes that the

industry has ample experience with

antifreeze that additional testing and

validity is not necessary. Mixing of the

antifreeze is a site-specific issue that

will ensure that the proper amount of

antifreeze is added to prevent the water

from freezing at temperatures down to

0°F.

Comment I.16: One commenter

recommended the addition of a note to

Tables 4.4.20 and 4.4.21 of the SAR to

provide clarification for the heat loads.

Response: The NRC disagrees with the

comment. SAR Tables 4.4.20 and 4.4.21

refer to loading the MPC with uniformly

aged fuel assemblies emitting heat at the

design basis maximum rate. Section

4.4.2 identifies these assemblies as the

limiting design basis fuel assemblies.

Comment I.17: One commenter stated

that Holtec’s use of a two-by-four block

array to be equivalent to an infinite

array assumed a center-to-center

distance between casks of 18.6 feet. The

commenter stated that this equivalency

determination between an infinite array

and a two-by-four array is invalid where

the differences in cask spacing do not

meet the 18.6-feet center-to-center

assumption underlying the analysis.

The commenter noted that the PSF

facility design uses a 15-foot center-tocenter distance. The commenter stated

that any CoC issued for this cask system

must address this shortcoming.

Response: The NRC disagrees with the

comment. First, the PFS facility is

outside the scope of this rulemaking

activity. Second, as identified in Table

1.4.1 of the SAR, the analysis of a 2 by

N array was performed using a pitch of

13.5 feet, not 18.6 feet. The 18.6-feet

pitch is used for a square array. When

calculating an equivalent hydraulic

diameter for the square array and taking

into account that the center-to-center

spacing of neighboring casks between

the pads is 38 feet, as described in

Figure 1.4.1 of the SAR, the hydraulic

diameters for the two cases (square array

versus 2 by N array) is the same.

Comment I.18: One commenter stated

that thermal interaction of casks through

radiative heat transfer should be

considered. The commenter also stated

that the assumption that individual

casks will not interfere with cooling air

supply of each other may not be correct.

Response: The NRC agrees that

neighboring casks can have an influence

on each other. However, these

influences have a second order impact

on the results. The analyses performed

by Holtec did credit radiation heat

transfer between the neighboring casks.

A bounding calculation was performed

with an ambient temperature of 125°F.

That calculation accounted for heat

reflected by the hot concrete pad and

heat generated by neighboring casks.

Although not required by the NRC

staff’s review of the SAR submittal,

Holtec, in response to other inquiries,

performed a sensitivity study to

quantify the impact of neighboring casks

and the impact of the sun heating the

concrete pad.

The impact of increasing the spacing

between casks by a factor of five in the

radial direction resulted in a decrease in

the peak cask surface temperature of 16

oF for an ambient temperature of 100°F

and 17°F for an ambient temperature of

125°F. The impact on peak clad

temperature resulted in a decrease of

6°F for an ambient temperature of 100°F

and a decrease of 8°F for an ambient

temperature of 125°F. Because the peak

clad temperature is on the order of

760+°F, the impact of neighboring casks

is minimal.

Comment I.19: One commenter stated

that the temperature of the reflecting

boundary should be taken as the

temperature of the cask in interaction

with the other casks and not the

temperature of an isolated cask.

Response: The NRC agrees that one

method for calculating the impact of

neighboring casks is to model the

neighboring casks in the array. Another

acceptable method, that was used by the

applicant, is to model the limiting

(highest temperature) cask and assume

that all the radiation it emits is reflected

back. This analysis bounds the amount

of radiation that neighboring casks can

impose on the center cask. This

bounding analysis is acceptable. As

noted in the response to comment I.18,

above, the impact of neighboring casks

is minimal, given the significant

margins between the allowable

temperatures and the bounding

calculated temperatures.

Comment I.20: One commenter stated

that the Holtec model does not appear

to take into account that the heating of

the concrete pad is likely to diminish

the ‘‘chimney effect’’ of the intake and

outlet vents. The commenter stated that

if Holtec had taken this effect into

account, the calculated temperature

would be higher in Revision 9 of the

SAR.

Response: In a response to other

inquires, Holtec performed calculations

to quantify the effect of concrete pad

heating on the cask performance. For

the bounding 125°F ambient

temperature event, neglecting the heat

reflected by the pad resulted in a

reduction of cask surface temperature of

10°F and a reduction in peak clad

temperature of 6°F. These temperature

differences illustrate that the concrete

pad has negligible impact on the cask.

Comment I.21: One commenter stated

that ambient temperature should be

defined due to the importance of the

term. The commenter noted that

ambient temperature is an important

assumption in the thermal calculations

and an important design element in the

CoC. The commenter stated that the

gross oversimplification of the concept

of ambient temperature renders the

Holtec thermal analysis completely

useless. The commenter noted that

Holtec assumes that the ambient

temperature at the intake and outlet

vents is the same; however, the

temperature at ground level will be

significantly higher than it will be some

distance above due to the ground

absorbing solar energy. The commenter

stated that a desert may have a surface

temperature of 180°F, much higher than

the 80°F assumed by Holtec as an intake

temperature. This would reduce the

effective buoyancy and air velocity

through the cooling ducts and result in

a higher fuel cladding temperature.

Response: The NRC disagrees with the

comment. The thermal response of a

cask is very slow. This is due to the

large mass of the system. An analogy

can be reached by observing the

buildings constructed in the desert.

Massive concrete is used to maintain the

indoor temperatures at reasonable

conditions where air conditioners do

not exist. The temperature in those

regions fluctuates over each day. For

these structures, an estimate of the

average conditions can be assessed by

assuming a bounding average daily

temperature. Holtec used such a

method. In addition, the method

25258 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

assumed the maximum solar heating

specified in 10 CFR Part 71 averaged

over a 24-hour period. Holtec used

bounding assumptions approved in the

NRC staff’s SER.

Comment I.22: One commenter stated

that NRC should have reviewed the

inputs and outputs of the FLUENT

calculation. The commenter also stated

that the NRC should have conducted an

independent analysis and validation of

the thermal model employed by Holtec.

The commenter stated that the HI–STAR

analysis cannot be extrapolated to the

HI–STORM cask because the casks are

constructed of different materials, with

different methods of heat dispersion.

The commenter stated that the NRC

performed a superficial review and had

abdicated its role as independent

regulator and should not issue a CoC for

the HI–STORM 100 cask system because

there is no lawful basis.

Response: The NRC disagrees with the

comment. The NRC staff’s review of the

HI–STORM system was not superficial.

This is clearly demonstrated by the NRC

staff’s requests for additional

information, the applicant’s many

revisions to the SAR to address NRC

staff concerns, and commitments made

by the applicant as outlined in

Appendix 12B of the SAR. The NRC

staff does not perform independent

confirmatory calculations for every

analysis submitted in an application,

nor does the NRC staff routinely review

the inputs and outputs of the computer

calculations without cause. Independent

analyses that duplicate the extensive

computer calculations performed by an

applicant may, at times, be performed

for various reasons. Some reasons

include, but are not limited to, concern

that a major error exists in the

calculations; allegations that

calculations were improperly

performed; use of new modeling

techniques not previously reviewed by

the NRC staff; crediting heat transfer

mechanisms not previously reviewed by

the NRC staff; concern that the margin

in a complex analysis is small; and

concern that little conservatism exists in

the modeling approach.

For the HI–STORM application, the

NRC staff reviewed the basic

assumptions used in the calculations, as

identified in the SAR and in the NRC’s

requests for additional information. A

detailed review of every number is not

warranted. As for performing

independent analysis and validation,

the NRC staff was able to reach its safety

findings without the need for such

calculations. The need for these

calculations is case specific, as

addressed above. For HI–STORM, the

applicant used computer codes that are

employed by the NRC and have been

found acceptable. The applicant

demonstrated its knowledge of the code

by benchmarking its methodology with

a full-scale spent fuel cask instrumented

with thermocouples, validating its

thermal model and providing reasonable

assurance that its analysts have good

working knowledge of the code to

perform the required calculations. The

NRC staff’s review of the applicant’s

methods and assumptions indicate

ample margin and conservatism in the

analyses.

The HI–STORM application review

process was conducted under NRC

policy and guidance, and as required by

the regulations in 10 CFR Part 72.

Regarding the reference to the HI–STAR

analysis in Section 4.5.4 of the

preliminary SER, the NRC staff intended

to indicate that it was aware that

Holtec’s use of the FLUENT code had

been previously found acceptable for

the HI–STAR application. This

reference was not intended to imply that

the NRC staff relied on the HI–STAR

calculations or the prior evaluation in

its evaluation of the HI–STORM cask.

Section 4.5.4 of the SER has been

modified to clarify the description of the

NRC staff’s review. Also, to better

illustrate the NRC staff’s review of the

applicant’s submittal, Section 4 of the

SER was supplemented with additional

information.

Comment I.23: One comment

indicated that the SER states that the

ambient temperature under normal

conditions must be less than 80°F. In

addition, the commenter believed

Holtec assumed that the ambient

temperature at the inlet and outlet vents

is the same and did not consider

warming of the air by heat generated by

neighboring casks and the concrete pad.

The commenter stated that calculations

indicated that a desert may have a

surface temperature of 180°F and that

the temperature 0.5 m above the ground

would be 130°F.

Response: The NRC disagrees with the

comment. The SER does not state that

the ambient temperature under normal

conditions must be less than 80°F. The

applicant evaluated the cask conditions

with an annual average ambient

temperature of 80°F. The use of an

annual average ambient temperature is

used in conjunction with the method

described in a Pacific Northwest

National Laboratory report PNL–6189.

The method provides one acceptable

means for obtaining reasonable

assurance that the requirements in 10

CFR Part 72 will be met. These

requirements include protecting the

cladding from degradation that leads to

gross ruptures and designing the storage

system to allow ready retrieval of the

spent fuel or high-level radioactive

waste. With respect to the 180°F surface

temperature in the desert, the SAR

assumptions used in the 125°F ambient

temperature calculation credits solar

heating (also referred to as solar

insolation) and heat generated by the

casks. Holtec calculated a concrete pad

surface temperature of 206°F

(surrounding the concrete overpack), an

ambient temperature just above the inlet

vent of the overpack of 136°F, and a

concrete temperature at the outlet vent

of the overpack of 182°F. The NRC staff

finds that the Holtec calculation

adequately models the thermal

responses of the cask and its

environment.

J. Technical Specifications

Comment J.1: One commenter asked

for clarification on the conditions for

use and the TSs, and if they could be

changed without an amendment.

Response: The conditions for cask use

are specified in the CoC, and includes

Appendix A (TSs) and Appendix B

(Approved Contents and Design

Features). These conditions cannot be

changed without an amendment to the

certificate.

Comment J.2: One commenter stated

that the Use and Application section of

the TSs is confusing and allows too

much flexibility for completion times

and frequencies, and that the TSs

should be simple to understand and

done on time.

Response: The NRC disagrees with the

comment. The Section 1.0, ‘‘Use and

Application’’ of the HI–STORM 100 TSs

are modeled on the Improved Standard

Technical Specifications (ISTS) for

power reactors. The ISTS were

developed as the result of extensive

technical meetings and discussions

between the NRC staff and the nuclear

power industry in the early 1990s in an

effort to improve clarity and consistency

of the power reactor TSs and to make

them easier for operators to use. The

most likely users of the HI–STORM 100

Cask System TSs are power reactor

licensees familiar with the format of the

ISTS. The NRC staff believes that the

format of the HI–STORM 100 TSs will

make them easier for operators to use

and will help to achieve consistency

between power reactor and spent fuel

dry cask storage TSs. The NRC staff

disagrees that there is too much

flexibility for completion times and

frequency. The NRC staff believes that

the specific wording of the TSs clearly

specifies the allowable time to complete

a required action and the frequency of

any surveillance requirements.

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25259

Comment J.3: One commenter

objected to the use of the term

‘‘TRANSPORT’’ in TS 3.2.2 and

indicated that movement to the pad

should be used because this CoC is for

storage only.

Response: The NRC disagrees with the

comment. The term TRANSPORT

OPERATIONS is specifically defined in

Section 1.1 of the Technical

Specifications and includes all activities

involved in moving a loaded overpack

or transfer cask to and from the ISFSI

pad. Further clarification of the term is

not warranted.

Comment J.4: One commenter stated

that ‘‘Each’’ should be in large letters in

LCO 3.2.2. The commenter also asked

why all the removable contamination is

not removed instead of setting a limit.

Response: The NRC disagrees with the

comment. The capitalization of ‘‘each’’

is consistent with the format of the TSs.

As discussed in the TS Bases, Section

B.3.2.2, the contamination limits for the

transfer cask are established from

guidance in NRC IE Circular 87–01. The

limits are based on minimum level of

activity that can be routinely detected

under a surface contamination control

program using direct survey methods.

These limits are consistent with levels

that prevent the spread of

contamination to clean areas and are

significantly less than the levels

associated with significant occupational

exposure.

Comment J.5: One commenter stated

that the dry run should be conducted in

sequence and not an alternate step

sequence as permitted by TS 5.2.

Response: The NRC disagrees with the

comment. The dry runs are performed

in discrete functional areas to

demonstrate the ability to perform

certain activities as anticipated. The

order of performance of the functional

areas, for the purpose of a dry run, is not

directly pertinent to a demonstration of

a user’s capability. The operating

procedures and technical specifications

already control, as necessary, functional

areas that must be performed

sequentially for safe storage. The NRC

staff considers it important to allow the

cask user the necessary flexibility to

allocate the appropriate resources and

oversight to the performance of dry runs

thatmay involve performing and

concentrating on certain activities that

would be out of sequence with a cask

loading.

Comment J.6: One commenter asked

why no lifting height limit was

established for the vertical orientation of

the transfer cask in TS 5.5 and stated

that there should be a limit established.

Response: In the SAR, the design

basis drop event analysis is based on the

horizontal lifting height of 42 inches.

Therefore, TS 5.5 only specifies the

lifting height of the horizontal lifting

limit. TS 5.5.c permits vertical lifting of

loaded transfer cask to any height

necessary to perform cask handling

operations, including the MPC transfer.

However, the lifts must be made with

structures and components designed to

prevent a drop and in accordance with

the criteria specified in CoC, Appendix

B, Section 3.5 and SAR Section 2.3.3.1.

Therefore, a vertical lift height limit was

not established.

Comment J.7: One commenter asked if

the diamond-shaped water rod

mentioned in note 10 of TS Table

2.1–3 had been completely analyzed.

Response: The shape (geometry) of

water rods that are part of the fuel

assembly, is considered in the

evaluation.

Comment J.8: One commenter

recommended deleting the words ‘‘For

OVERPACKS with installed temperature

monitoring equipment’’ at the beginning

of the second option under SR 3.1.2.1

because users should have the option of

using temporary equipment.

Response: The NRC agrees with the

comment. Temperature monitoring, a

surveillance option permitted in SR

3.1.2.1, could be conducted with either

temporary or permanently installed

equipment. The term ‘‘installed’’ could

be interpreted as a requirement that the

temperature monitoring equipment be

permanently fixed. Therefore, the

beginning of the second option under

SR 3.1.2.1 has been reworded as

follows: ‘‘For OVERPACKS with

temperature monitoring equipment’’

(i.e., the word ‘‘installed’’ has been

deleted).

Comment J.9: One commenter

recommended several miscellaneous

editorial changes to the appendices to

the CoC.

Response: The NRC agrees with the

comment. The appendices to the CoC

have been revised to correct

typographical errors and incorporate

minor editorial changes.

Comment J.10: One commenter

recommended that Items 5.2.f and 5.2.j

in Section 5 of the TSs be revised to

insert the phrase ‘‘(for which a mock-up

may be used)’’ at the end of the items

for consistency with SAR Section

12.2.2.

Response: The NRC agrees with the

comment. Items 5.2.f and 5.2.j of the

TSs have been revised to indicate that

a mock-up may be used for those

specific dry-run evolutions.

Comment J.11: One commenter

recommended that item 5.5.c in Section

5 of the TSs be revised to replace the

words ‘‘and MPC’’ with ‘‘or

OVERPACK’’ because some utilities

plan to implement an MPC transfer

scheme that requires temporary lifting

of the loaded OVERPACK above its lift

height limit.

Response: The NRC disagrees with the

comment. There are no evaluations,

equipment design criteria, or other

information in the SAR that support

lifting a loaded overpack above its lift

height limit.

Comment J.12: One commenter

recommended revising the definitions of

DAMAGED FUEL ASSEMBLY and

PLANAR–AVERAGE INITIAL

ENRICHMENT in Section 1 of Appendix

B to the CoC to reflect the evolution of

these terms and for consistency with

those in the HI–STAR 100 CoC.

Response: The NRC agrees with this

comment. The CoC, Appendix B,

Section 1 has been revised to reflect the

new definitions.

Comment J.13: One commenter

recommended revising CoC, Appendix

B, Section 3.4.6.c to replace the

specified yield strength with the

equivalent ASTM Grade specification.

Response: The NRC agrees with the

comment in part. Storage pad design is

a site-specific issue that needs to be

addressed in the cask user’s 10 CFR

72.212 evaluation. CoC, Appendix B,

Section 3.4.6.c lists the design

parameters for the storage pads. It is not

a list of components for fabrication. By

using the specific ASTM Grade

specification as recommended by the

commenter, namely, ASTM A615, Grade

60, the designer of the pad will not have

the flexibility to choose other

reinforcing steels that could also be

used (e.g. ASTM A616 or A617, Grade

60, etc.). To allow flexibility for the

design and still ensure adequate

reinforcement in the pad CoC,

Appendix B, Section 3.4.6.c has been

changed to state that reinforcement shall

be 60 ksi yield strength ASTM material.

Comment J.14: One commenter

recommended eliminating the

requirement for impact limiters at the

cask transfer facility contained in CoC,

Appendix B, Item 3.5.2.2.

Response: The NRC staff assumes that

the commenter’s reference to Section

3.5.2.2 is a typographical error because

the requirement for an impact limiter is

in CoC, Appendix B, Item 3.5.2.1. The

NRC agrees in part with the comment.

The specific requirement for an impact

limiter has been eliminated from CoC,

Appendix B, Section 3.5.2.1.4. The NRC

determined that this requirement is too

restrictive because other methods may

be available to prevent a canister breach

in the event of a canister drop during

transfer operations. Instead, Item

3.5.2.1.4 has been revised to require that

25260 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

the CTF be designed, constructed, and

evaluated to ensure that if the MPC is

dropped during an inter-cask transfer

operation, its confinement boundary

would not be breached.

However, the NRC disagrees with the

underlying reason for the comment

which is: Because a single failure proof

crane (or equivalent) is required in the

CTF, the design features to mitigate the

consequence of a drop should not be

necessary. The NRC staff acknowledges

that the use of a single-failure proof

crane precludes the possibility of a

heavy load drop event. The requirement

for a mitigating feature in the CTF

design is a defense-in-depth measure

that is consistent with the overall

philosophy and approach of NUREG–

0612. This philosophy encompasses an

intent to prevent as well as to mitigate

the consequences of postulated

accidental load drops. The NRC staff

notes that, even with a single-failure

proof crane, NUREG–0612 still imposes

a requirement for a safe load travel path

‘‘to minimize the potential for heavy

loads, if dropped, to impact irradiated

fuel in the reactor vessel and in the

spent fuel pool, or to impact safe

shutdown equipment.’’ The NRC staff

views the mitigating feature in the CTF

as a defense-in-depth measure

equivalent to the safe load path. Its

function is to protect the MPC

confinement boundary and the integrity

of the spent fuel in the MPC in case of

a postulated drop.

Comment J.15: One commenter

recommended that CoC, Appendix B,

Item 3.5.2.1.4 be clarified to indicate

that the acceptance criterion for the

impact limiter also applies to the use of

mobile cranes.

Response: The NRC agrees with the

comment. As discussed in the response

to J.14, CoC, Appendix B, Item 3.5.2.1.4

has been revised to require that the CTF

be designed and evaluated to ensure

that if the MPC is dropped during an

inter-cask transfer operation, its

confinement boundary would not be

breached. Section 3.5.2.1.4 has also

been revised to specify that this

requirement and acceptance criterion

apply to both stationary and mobile

cranes.

Comment J.16: One commenter

recommended that CoC, Appendix B,

Item 3.5.2.1.4 be revised to clarify the

scope of drops that require evaluation in

designing the impact limiter.

Response: The NRC agrees with the

comment. CoC, Appendix B, Item

3.5.2.1.4 has been revised to clarify that

the potential drops that require

evaluation are those that may occur

during inter-cask transfer operations.

Comment J.17: One commenter

recommended that the TSs be removed

from Appendix 12.A of the SAR because

they are included in the appendices to

the CoC.

Response: The NRC agrees with the

comment. The TSs have been removed

from the SAR.

K. Miscellaneous

Comment K.1: One commenter

expressed approval that movement

could be conducted at 0°F and above.

Response: No response is necessary.

Comment K.2: One commenter stated

that the HI–TRAC transfer cask must be

as safe as the HI–STORM overpack if it

is to be used outside the reactor security

fence.

Response: The NRC staff reviewed the

HI–TRAC transfer cask and determined

that, like the HI–STORM overpack, it

will perform its intended safety

functions under the design basis

normal, off-normal, and accident events.

It should be noted that the CoC

authorizes use of the HI–TRAC only

within the owner-controlled areas of a

licensed power reactor.

Comment K.3: One commenter asked

if the inner canister could be dropped

through, if water could spill out of the

overpack, and if the water helped to

disperse the fuel particles.

Response: During a canister transfer

operation, the transfer cask is placed on

top of the storage overpack. The canister

is then lowered through the bottom of

the transfer cask into the overpack. It is

unlikely that a canister drop would

occur during this operation because the

canister must be lifted with equipment

(i.e., a single failure proof crane or

equivalent) that are designed to prevent

a drop. In addition, the overpack

contains only traces of water that is part

of the concrete material and the canister

is dry during cask transfer operations.

Comment K.4: One commenter

questioned the assumption that the HI–

TRAC remains static because there are

a number of man-made or natural causes

that could put it in motion, drop,

tipover, roll, etc.

Response: The HI–TRAC is required

to be independently secured on top of

the overpack during the transfer of the

MPC.

Comment K.5: One commenter asked

when the measuring equipment (for

checking tolerances) is calibrated.

Response: The timing of calibration at

the fabricator’s facility is beyond the

scope of this rule. However, the

implemented QA program at the

fabricator’s facility provides reasonable

assurance that the measuring equipment

for checking tolerances of fabrication

will be appropriately calibrated.

Comment K.6: One commenter asked

if the restraint of 11 inches in vertical

height for overpack handling would

actually preclude a corner drop

situation. The commenter asked how a

corner drop could be initiated, such as

a defective trunnion or lifting lug, etc.

Response: The 11-inch restriction on

lifting height for the overpack was

calculated to ensure that deceleration

loading to the loaded MPC would not

exceed the design criteria for the

confinement boundary of the basket. A

tipover of the overpack cannot occur if

the baseplate is limited to 11 inches

above a receiving surface.

Comment K.7: One commenter asked

what happens to the inside of the cask

during a horizontal drop of 50 inches.

Response: The effect of a 50-inch

horizontal drop of a cask was not

evaluated because the horizontal lifting

height limit for the transfer cask is 42

inches. The 50-inch carry height

specified in the SER was a

typographical error and has been

corrected to 42 inches. There is no effect

on the confinement function of the MPC

as a result of a horizontal drop of 42

inches. The structural evaluation shows

that all stresses are within allowable

values and that the confinement

boundary integrity of the MPC is not

impaired.

Comment K.8: One commenter

requested that the SER define what is

meant by cladding oxide thickness on

page 4–1 of the SER.

Response: The NRC disagrees with the

comment. Cladding oxide thickness is a

measure of corrosion at the clad surface.

As water interacts with the zirconium

clad, the zirconium can interact with

the oxygen molecules to create

zirconium oxide (ZrO2). The

terminology is commonly used in the

spent fuel storage arena and a definition

in the SER is not necessary.

Comment K.9: One commenter asked

why the internal rod pressure is

assumed to remain the same. The

commenter asked how the gas behaves

in a dry cask and if it can leak from

pinhole leaks and hairline cracks over

the storage period. The commenter

further asked how the lower pressure in

the rods affects the analysis and heat

transfer.

Response: The internal rod pressure is

derived from the initial gas inserted

during fabrication plus the fission

product gases that develop during

power production within the reactor

core. In a closed system (e.g., the fuel

pin), the pressure is a function of the

gases in the fuel rod and the average

temperature of the gas. As the decay

heat decreases with time, so does the

temperature and the pressure.

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25261

Therefore, the rod temperature does not

remain the same. This is similar to

inflating a balloon with hot air and

placing the balloon in the refrigerator.

As the gases cool, the pressure

decreases, as is implied by the smaller

diameter of the balloon. The lower

pressure reduces the stress on the

cladding and permits a higher allowable

temperature limit. If the rod experiences

a pinhole leak or a hairline crack, the

gases inside the rod will mix with the

helium gas in the cask and reduce the

internal pressure within the rod.

Reduction of the internal fuel rod

pressure results in added assurance that

the cladding will remain stable because

the internal pressure will have

equilibrated with that of the cask. The

gases from the fuel pin mix with the

gases in the cask and decreases the

thermal conductivity of the helium,

while at the same time increasing the

density of the gas. The analyses for

accident conditions incorporate the

impact of reduced conductivity of the

helium gases. This impact is reduced

when crediting cooling that results from

natural circulation of the gases inside

the cask. The use of a maximum

allowable temperature limit provides

assurance that the fuel pins will remain

intact throughout the storage period. For

conservatism, the applicant assumed

that 1 percent of the cladding

experiences a leak under normal

conditions, a 10-percent leak under offnormal conditions, and a 100-percent

leak under accident conditions.

Comment K.10: One commenter asked

what cask design was tested at INEEL

(page 4–3 of the SER).

Response: Several full scale cask

designs were tested at the Idaho

National Engineering and

Environmental Laboratory. The cask

used by Holtec to validate the FLUENT

computer code was the TN–24P. The

heat output of the cask was 23 kW. The

NRC staff found the FLUENT computer

code acceptable for calculating the

thermal response of a spent fuel cask.

Comment K.11: One commenter

expressed concern over water and

debris going into cracks on the pad and

then freezing and thawing causing

concrete upheaval and subsequent cask

tipover.

Response: Issues related to cask

storage pad will be addressed in the

cask user’s evaluation under 10 CFR

72.212 and is beyond the scope of this

rule.

Comment K.12: One commenter asked

how moisture and pollution in the air

could affect the casks and pad over time

and if the pad would ever need to be

replaced.

Response: The cask can withstand the

ambient environmental conditions over

its 20-year license period with no

significant degradation. The adequacy of

the pad must be addressed by the cask

users in their 10 CFR 72.212 evaluation

and is beyond the scope of this rule.

Cask users are responsible for inspecting

and maintaining the pad. With

appropriate maintenance, air pollution

or moisture would not cause significant

degradation to the pad.

Comment K.13: One commenter asked

if both the helium and fission gases

created the pressure inside the rods and

for an explanation of the fission gases.

The commenter also asked why only 30

percent of the fission product gas was

assumed to be released instead of 100

percent because over time 100 percent

would likely leak out.

Response: Fission gases are

byproducts of uranium splitting in a

reactor. These include gases such as

hydrogen, krypton, and iodine. The

gases are contained inside the fuel rod.

Data have shown that a conservative

estimate of 30 percent of the gases

generated inside the fuel pellet can

escape to the gap that exists between the

fuel pellet and the cladding. This is a

conservatively large number used for

calculating dosage. Experimental data

has shown this number to be

significantly less. The rest of the gases

are trapped inside the fuel pellet.

Therefore, assuming that 100 percent of

the gases are released from the fuel

pellet is not realistic. Helium gas is

added to the MPC to keep the

environment inside the cask inert so it

does not promote corrosion and to help

cool the fuel by transferring heat from

the fuel rods to the wall of the cask. The

impact of helium gas on the pressure

within a fuel rod is not as significant as

the temperature of the gas within the

fuel rod.

Comment K.14: One commenter asked

what is in the water of the water jacket

and if the water could affect the carbon

steel channels or get into the pool

through a weld crack or leak and affect

the pool. The commenter also asked

how hot the water and the lead get, and

if the water could cause pressure

buildup in the channels.

Response: The water used in the

water jacket is demineralized water as is

used in the loading pool, but without

boron addition because the boron is

unnecessary for loading/unloading

operations. Carbon steel corrodes very

slowly in demineralized water; thus, its

effect may be ignored for the durations

experienced in loading operations. If the

cask is to be loaded in cold weather,

antifreeze may be added to the jacket

water. Antifreeze contains an inhibitor

to prevent corrosion. There would be no

significant effect if the jacket water or

water with antifreeze leaked into the

pool. With regard to the water and lead

temperatures and pressure buildup in

the water jackets, see the response to

comment I.2.

Comment K.15: One commenter asked

if there was a recent study on cladding

degradation from creep cavitation.

Response: Studies on cladding

degradation were performed several

years ago. These studies led to the

development of analytic methods to

calculate the maximum allowable peak

clad temperature limits. A report

developed by the Pacific Northwest

National Laboratory (PNL) in May 1987,

PNL–6189, ‘‘Recommended

Temperature Limits for Dry Storage of

Spent Light Water Reactor Zircaloy-Clad

Fuel Rods in Inert Gas’’ provides an

acceptable method for assessing

cladding temperature limits.

Comment K.16: One commenter stated

that the 100-ton transfer cask should not

be included in the certification because

it is site-specific and not made the same

as the 125-ton cask.

Response: NRC disagrees with the

comment. The 100-ton and 125-ton

transfer cask designs have been

evaluated and found to meet the

regulatory requirements of 10 CFR Part

72. The 100-ton transfer cask design is

not considered site-specific and is

approved under this rule for use by any

general licensee as part of the HI–

STROM 100 system as described in the

SAR. Section 2.0.3 of the SAR provides

guidance regarding site-specific ALARA

objectives that should be considered by

each user when using either transfer

cask design.

Comment K.17: One commenter asked

what does reasonable assurance mean in

Section 5.1.2 of the SER regarding

acceptability of the shielding design

criteria.

Response: The finding in Section

5.1.2 is intended to mean that the NRC

staff believes that the dose rate criteria

presented in the SAR are acceptable

values and that a cask system operating

at these values can meet the applicable

radiological requirements of 10 CFR

Parts 20 and 72. The SAR subsequently

demonstrates that the dose rates

calculated for the HI–STORM system

meet the regulatory requirements.

Comment K.18: One commenter asked

if the MOX (mixed oxide) fuel was

covered by the sabotage report. The

commenter asked if MOX fuel had been

tested and verified to be safe for this

design. The commenter further

questioned how the NRC could include

MOX fuel in the SER evaluation and

stated that storage of MOX fuel should

25262 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

not be allowed by the certification. The

commenter also asked how we know

that storage of MOX fuel will work as

expected because it has not yet been

tested in Canada.

Response: The sabotage report is

beyond the scope of this rule. However,

the design and physical characteristics

of a MOX fuel assembly are very similar

to those of a uranium fuel assembly. The

primary difference is the fuel pellet

constituents and its effects on the

radiological source term. Testing of

MOX fuel is also beyond the scope of

this rule.

The HI–STORM design was evaluated

for storage of the MOX fuel assemblies

listed in the Appendix B to the CoC

using computer codes and models. In

lieu of testing, the NRC finds analytic

conclusions that are based on sound

engineering methods and practices to be

acceptable. Testing is only required if

the analytic methods have not been

validated or assured to be appropriate

and/or conservative. The NRC staff

reviewed the applicant’s analyses and

found them acceptable. The basis of the

safety review and findings are identified

in the SER and the CoC.

Comment K.19: One commenter asked

if all the analysis was based on the 100-

ton transfer cask or did HI–STORM 100

refer to something else.

Response: The shielding analysis

presented in the SAR evaluated both the

100-ton and 125-ton transfer cask

designs as part of the HI–STORM 100

cask system.

Comment K.20: One commenter asked

how the NRC could base its evaluation

on historical statements when reference

documents indicate Inconel impurity

may be higher than 1000 ppm. The

commenter further asked what the

historical statements were and how we

know if the statements are valid.

Response: The applicant’s analysis of

cobalt impurities are discussed in

Section 5.2.1 and 5.2.3 of the SER. The

applicant showed that the cobalt

impurity value of 1000 ppm assumed in

the shielding analyses was appropriate

based on industry data and analyses of

post-irradiation cooling of older fuel

types that may have had higher cobalt

impurities for the HI–STORM 100 cask

system. As discussed in Section 5.2.1 of

the SAR, historical statements included

industry data gathered by the applicant

from utilities and vendors.

Cobalt impurities were not necessarily

controlled for older fuel designs.

However, the applicant showed that the

post-irradiation cooling time that is

inherent to these older fuel types

significantly reduces the HI–STORM

100 dose rates. Therefore, the effects of

higher impurities are mitigated. Based

on historical knowledge of recent cobalt

reduction programs, the decay effects on

older fuel, and its own independent

evaluations, NRC has reasonable

assurance that the historical statements

referenced in the application are used

appropriately for the HI–STORM 100.

Furthermore, each cask user will

operate the HI–STORM 100 under a 10

CFR Part 20 radiological protection

program and will be required to verify

dose rates that are specified in the TSs.

This defense-in-depth approach will

mitigate potential hardware activation

anomalies and ensure compliance with

radiological requirements.

Comment K.21: One commenter asked

if the steel transport overpacks could be

reused, how contaminated the

overpacks would be after use, the

number of times an overpack could be

reused, and if they would be checked

after each use.

Response: This comment that

concerns the HI–STAR steel transport

overpack, is beyond the scope of this

rule on the Holtec HI–STORM 100 cask

system.

Comment K.22: One commenter was

pleased that the NRC had evaluated

uneven flooding.

Response: No response is necessary.

Comment K.23: One commenter asked

about the chance of one of the screens

being damaged or loosened in unloading

and the debris floating out with the

cooling water into the pool.

Response: The damaged fuel

container that is placed in the MPC is

stainless steel and is designed to retain

damaged fuel and debris in a safe

configuration under all normal, offnormal, and accident conditions. The

damaged fuel container also provides a

means to safely handle the damaged fuel

and debris during loading and

unloading. It is not considered credible

that the screens will fall off or fail.

However if a screen failed, there would

be no release of radioactive material

during storage since the MPC is sealed.

Consideration of loose debris during

unloading is addressed in SAR Section

8.3 which outlines the MPC unloading

operations in a spent fuel pool and

specifically considers loose debris in the

MPC. Additionally, the spent fuel pool

filtration system would capture any

debris that remained in the pool.

Comment K.24: One commenter asked

why the volume of water removed from

the cask is recorded and why this is not

done for other cask designs.

Response: The purpose of recording

the volume of water removed from the

canister is to identify the open volume

in the canister. This open volume is

used to calculate the amount of helium

to be added to the cask following

vacuum drying. The procedure and

equation used for this procedure is

discussed on page 8.1–21 in the HI–

STORM SAR. The comment concerning

other cask designs is beyond the scope

of this rule.

Comment K.25: One commenter stated

that a detailed procedure on mitigating

the possibility of fuel crud particulates

dispersal should be included in the

documents and that the procedure

should not be site-specific.

Response: NRC disagrees with the

comment. The generic unloading

procedures for the HI–STORM 100

system are designed to mitigate crud

dispersal. However, each cask user will

need to develop detailed unloading

procedures that incorporate the ALARA

objectives of its site-specific radiation

protection program. NRC expects the

cask user to consider the specific

characteristics of its fuel, including crud

phenomena, when developing these

procedures.

Comment K.26: One commenter asked

how the utilities are required to

document that they will not lift the

overpack any higher than 11 inches and

that the receiving surface hardness does

not exceed that analyzed in the SAR.

The commenter stated that the criteria

should be clarified and which surface

should be indicated.

Response: The receiving surface is the

top of the storage pad as clearly stated

in Sections 3.4.2 and 11.2.3.2 of the SER

and described in Section 3.4.10 of the

SAR. Users of the HI–STORM 100

system are required to meet Appendices

A and B of the CoC that list the design

parameters for surface hardness and the

restriction for lifting height.

Furthermore, the cask users are required

to develop detailed written operating

procedures. The restriction on lifting

height must be incorporated into the

operating procedures subject to NRC

inspection.

Comment K.27: One commenter stated

that Condition 8 should remain in the

CoC.

Response: The NRC agrees with the

comment. Condition 8 has not been

removed from the CoC. Under Condition

8, Certificate holders who wish to make

changes to the CoC, including

Appendices A and B, must submit an

application for amendment of the

Certificate.

Comment K.28: One commenter asked

how upending/downending of the

transfer cask affected the water in the

neutron shield, how the licensee knows

the shield is full, what happens to the

contents of the cask when the position

changes, what are the stresses and

pressures, and if the debris in damaged

fuel containers goes through the screen.

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25263

Response: The structural, shielding,

and confinement functions of the

transfer cask are not affected during

movement of the cask. The neutron

shield will normally be filled through

the drain valve at the bottom of the

water jacket and is considered full as

water exits the vent port at the top of the

water jacket. The vent plug is then

installed to retain the water in the

jacket. During the upending and

downending of the transfer cask, water

remains within the neutron shield and

fuel debris remains within the

confinement boundary of the MPC. The

structural evaluation in the SAR showed

all the stresses and pressures to remain

within allowable values.

Comment K.29: One commenter stated

that exceptions to the codes should not

be allowed and that the NRC should

demand full code requirements.

Response: The NRC disagrees with

this comment. Exceptions (alternatives)

to the ASME Code specifications may be

granted by the NRC staff on a case-bycase basis. During the NRC staff review

of a proposed alternative, the applicant

must demonstrate that the proposed

alternative to the Code satisfies one of

the following criteria: (1) The alternative

provides an acceptable level of quality

and safety, or, (2) compliance with a

specific Code requirement would result

in hardship or unusual difficulty

without a compensating increase in the

level of quality or safety.

Comment K.30: One commenter stated

that videos should not be used as a

permanent record. The commenter

stated that black and white photos and

negatives should be used and that the

negatives should be kept in museum

qualified storage. The commenter asked

what method is best to document weld

integrity and how the records are stored.

The NRC should have specific criteria

for record keeping requirements.

Response: The NRC disagrees with the

comment. The NRC’s regulations do not

explicitly require specific criteria for

record keeping to document weld

integrity by the applicant. A permanent

record of completed welds will be made

using video, photographic, or other

means that can provide a retrievable

record of weld integrity. As per

accepted industry practice, the record is

typically in color format, in order to

capture the red dye typically used for

PT examinations. The general licensee’s

QA program will specify the types of

records and how the records are to be

stored.

Comment K.31: One commenter stated

that even if the overpack baseplates,

shell, pedestal shell, and radial plates

have large margins of safety in the

design, they should still be examined to

code.

Response: Holtec has committed to

inspect the welds of the overpack

baseplate to the shell, pedestal shell,

and radial plates under ASME Code

Section V, Article 9. Weld inspection

acceptance criteria meet the

requirements in ASME Section III,

Subsection NF–5360.

Comment K.32: One commenter asked

why a mobile lifting device is used and

why it is not required to meet the

requirements of NUREG–0612, Section

5.1.6(2) for new cranes. If a new crane

is necessary to meet the requirements,

the utilities should get one and not be

allowed to lower requirements.

Response: A mobile lifting device is

an alternative option to a stationary

lifting device that may be used in a CTF.

The decision to use either a mobile or

stationary lifting device would be made

by the cask users and would be based

on their plant’s site-specific needs.

NUREG–0612, Section 5.1.6(2) specifies

that new cranes should be designed to

meet NUREG–0554, ‘‘Single-FailureProof Cranes for Nuclear Power Plants.’’

These requirements are not applicable

to mobile lifting devices which are not

single-failure-proof; therefore, mobile

lifting devices are exempted from this

particular requirement in NUREG–0612.

To ensure that the mobile lifting device

has the equivalent level of safety as a

single-failure-proof crane, additional

conditions in CoC, Appendix B,

Sections 3.5.2.2.1, 3.5.2.2.2, and

3.5.2.2.4 were imposed.

Comment K.33: One commenter stated

that a discussion on the cask transfer

facility should be included in the SER,

and that the public should not have to

read the SAR to understand the generic

design. The commenter requested that

this part of the cask transfer facility be

resubmitted with a complete clear

design with specific criteria.

Response: The NRC disagrees with the

comment. SER Section 1.1 discusses the

CTF in a level of detail appropriate for

an SER. The detailed design and

operating criteria for the CTF are given

in SAR Section 2.3.3.1. This satisfies 10

CFR 72.24, which requires that the SAR

contain information on structures,

systems, and components important to

safety in sufficient detail for the NRC

staff to make its regulatory finding.

Repeating this information in the SER is

not necessary. The NRC disagrees that

cask transfer facility should be

resubmitted with a complete clear

design with specific criteria. The

specific criteria for the CTF are already

given in CoC, Appendix B, Section 3.4,

and SAR Section 2.3.3.1. As discussed

in the response to E.30, NRC found it

unnecessary to approve a specific CTF

design.

Comment K.34: One commenter

recommended that Section 3.5.7 of the

SER be revised to reflect that transport

of the HI–TRAC transfer cask in the

vertical orientation is permitted. The

comment also recommended that ‘‘50

inches’’ be changed to ‘‘42 inches’’ to be

consistent with TS Table 5–1.

Response: The NRC agrees with the

comment. The SER has been modified to

reflect that transport of the HI–TRAC

transfer cask in the vertical orientation

is permitted. The horizontal lifting

height per TS Table 5–1 will be

corrected to 42 inches to correct the

typographical error.

Comment K.35: One commenter

recommended that Section 9.1.2.2.b of

the SER be revised to delete ‘‘(either to

the fuel pool or the site licensee’s offgas system)’’ because users may or may

not have these systems at their plants.

Response: The NRC agrees with the

comment. It is up to the cask users to

develop the specific procedures for

venting the MPC and to determine the

appropriate location under their plant’s

waste gas handling system design and

radiation protection program. Section

9.1.2.2.b of the SER has been modified

as recommended.

Summary of Final Revisions

As a result of the NRC staff’s response

to public comments, or to rectify issues

identified during the comment period,

TSs 5.2.f and 5.2.j have been modified

(see comment J.10). The NRC staff has

also updated the CoC, including

Appendix B, and has removed the bases

section from the TSs attached to the CoC

to ensure consistency with NRC’s format

and content. The NRC staff has also

modified its SER. In addition, the NRC

staff has modified the rule language by

changing the word ‘‘Certification’’ to

‘‘Certificate’’ to clarify that it is actually

the Certificate that expires.

Agreement State Compatibility

Under the ‘‘Policy Statement on

Adequacy and Compatibility of

Agreement State Programs’’ approved by

the Commission on June 30, 1997, and

published in the Federal Register on

September 3, 1997 (62 FR 46517), this

rule is classified as compatibility

Category ‘‘NRC.’’ Compatibility is not

required for Category ‘‘NRC’’

regulations. The NRC program elements

in this category are those that relate

directly to areas of regulation reserved

to the NRC by the Atomic Energy Act of

1954, as amended (AEA), or the

provisions of Title 10 of the Code of

Federal Regulations. Although an

Agreement State may not adopt program

25264 Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations

elements reserved to NRC, it may wish

to inform its licensees of certain

requirements via a mechanism that is

consistent with the particular State’s

administrative procedure laws, but does

not confer regulatory authority on the

State.

Finding of No Significant

Environmental Impact: Availability

Under the National Environmental

Policy Act of 1969, as amended, and the

Commission’s regulations in Subpart A

of 10 CFR Part 51, the NRC has

determined that this rule is not a major

Federal action significantly affecting the

quality of the human environment and

therefore, an environmental impact

statement is not required. This final rule

adds an additional cask to the list of

approved spent fuel storage casks that

power reactor licensees can use to store

spent fuel at reactor sites without

additional site-specific approvals from

the Commission. The environmental

assessment and finding of no significant

impact on which this determination is

based are available for inspection at the

NRC Public Document Room, 2120 L

Street NW. (Lower Level), Washington,

DC. Single copies of the environmental

assessment and finding of no significant

impact are available from Merri Horn,

Office of Nuclear Material Safety and

Safeguards, U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

telephone (301) 415–8126, e-mail

mlh1@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.

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 Holtec

International HI–STORM 100 cask

system to the list of NRC-approved cask

systems for spent fuel storage in 10 CFR

72.214. This action does not constitute

the establishment of a standard that

establishes generally-applicable

requirements.

Regulatory Analysis

On July 18, 1990 (55 FR 29181), the

Commission issued an amendment to 10

CFR Part 72. The amendment provided

for the storage of spent nuclear fuel in

cask systems with designs approved by

the NRC under a general license. Any

nuclear power reactor licensee can use

cask systems with designs approved by

the NRC to store spent nuclear fuel if it

notifies the NRC in advance, the spent

fuel is stored under the conditions

specified in the cask’s CoC, and the

conditions of the general license are

met. In that rule, four spent fuel storage

casks were approved for use at reactor

sites and were listed in 10 CFR 72.214.

That rule envisioned that storage casks

certified in the future could be routinely

added to the listing in 10 CFR 72.214

through the rulemaking process.

Procedures and criteria for obtaining

NRC approval of new spent fuel storage

cask designs were provided in 10 CFR

Part 72, Subpart L.

The alternative to this action is to

withhold approval of this new design

and issue a site-specific license to each

utility that proposes to use the casks.

This alternative would cost both the

NRC and utilities more time and money

for each site-specific license.

Conducting site-specific reviews would

ignore the procedures and criteria

currently in place for the addition of

new cask designs that can be used under

a general license, and would be in

conflict with NWPA direction to the

Commission to approve technologies for

the use of spent fuel storage at the sites

of civilian nuclear power reactors

without, to the maximum extent

practicable, the need for additional site

reviews. This alternative also would

tend to exclude new vendors from the

business market without cause and

would arbitrarily limit the choice of

cask designs available to power reactor

licensees. This final rule will eliminate

the above problems and is consistent

with previous Commission actions.

Further, the rule will have no adverse

effect on public health and safety.

The benefit of this rule to nuclear

power reactor licensees is to make

available a greater choice of spent fuel

storage cask designs that can be used

under a general license. The new cask

vendors with casks to be listed in 10

CFR 72.214 benefit by having to obtain

NRC certificates only once for a design

that can then be used by more than one

power reactor licensee. The NRC also

benefits because it will need to certify

a cask design only once for use by

multiple licensees. Casks approved

through rulemaking are to be suitable

for use under a range of environmental

conditions sufficiently broad to

encompass multiple nuclear power

plants in the United States without the

need for further site-specific approval

by NRC. Vendors with cask designs

already listed may be adversely

impacted because power reactor

licensees may choose a newly listed

design over an existing one. However,

the NRC is required by its regulations

and NWPA direction to certify and list

approved casks. This rule has no

significant identifiable impact or benefit

on other Government agencies.

Based on the above discussion of the

benefits and impacts of the alternatives,

the NRC concludes that the

requirements of the final rule are

commensurate with the Commission’s

responsibilities for public health and

safety and the common defense and

security. No other available alternative

is believed to be as satisfactory, and

thus, this action is recommended.

Small Business Regulatory Enforcement

Fairness Act

In accordance with the Small

Business Regulatory Enforcement

Fairness Act of 1996, the NRC has

determined that this action is not a

major rule and has verified this

determination with the Office of

Information and Regulatory Affairs,

Office of Management and Budget.

Regulatory Flexibility Certification

In accordance with the Regulatory

Flexibility Act of 1980 (5 U.S.C. 605(b)),

the Commission 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 Holtec International. 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

Federal Register / Vol. 65, No. 84 / Monday, May 1, 2000 / Rules and Regulations 25265

rule. Therefore, a backfit analysis is not

required.

List of Subjects in 10 CFR Part 72

Administrative practice and

procedure, Hazardous waste, Nuclear

materials, Occupational safety and

health, Penalities, Radiation protection,

Reporting and recordkeeping

requirements, Security measures, Spent

fuel, Whistleblowing.

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 adopting the following amendments

to 10 CFR part 72.

PART 72—LICENSING

REQUIREMENTS FOR THE

INDEPENDENT STORAGE OF SPENT

NUCLEAR FUEL AND HIGH-LEVEL

RADIOACTIVE WASTE

1. The authority citation for Part 72

continues to read as follows:

Authority: Secs. 51, 53, 57, 62, 63, 65, 69,

81, 161, 182, 183, 184, 186, 187, 189, 68 Stat.

929, 930, 932, 933, 934, 935, 948, 953, 954,

955, as amended, sec. 234, 83 Stat. 444, as

amended (42 U.S.C. 2071, 2073, 2077, 2092,

2093, 2095, 2099, 2111, 2201, 2232, 2233,

2234, 2236, 2237, 2238, 2282); sec. 274, Pub.

L. 86–373, 73 Stat. 688, as amended (42

U.S.C. 2021); sec. 201, as amended, 202, 206,

88 Stat. 1242, as amended, 1244, 1246 (42

U.S.C. 5841, 5842, 5846); Pub. L. 95–601, sec.

10, 92 Stat. 2951 as amended by Pub. L. 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 Section 72.214, Certificate of

Compliance 1014 is added to read as

follows:

§ 72.214 List of approved spent fuel

storage casks.

* * * * *

Certificate Number: 1014.

SAR Submitted by: Holtec

International.

SAR Title: Final Safety Analysis

Report for the HI–STORM 100 Cask

System.

Docket Number: 72–1014.

Certificate Expiration Date: June 1,

2020.

Model Number: HI–STORM 100.

Dated at Rockville, Maryland, this 12th day

of April, 2000.

For the Nuclear Regulatory Commission.

Frank J. Miraglia, Jr.,

Acting Executive Director for Operations.

[FR Doc. 00–10393 Filed 4–28–00; 8:45 am]

BILLING CODE 7590–01–P

DEPARTMENT OF ENERGY

Office of Energy Efficiency and

Renewable Energy

10 CFR Part 420

[Docket No. EE–RM–96–402]

RIN 1904–AB01

State Energy Program

AGENCY: Office of Energy Efficiency and

Renewable Energy, DOE.

ACTION: Final rule.

SUMMARY: Today the Department of

Energy (DOE or Department) adopts an

interim final rule published on August

24, 1999 revising the regulations for its

State Energy Program. Because there

were no comments received in response

to the program’s interim final rule, that

rule is being adopted as a final rule

without change.

EFFECTIVE DATE: May 31, 2000.

FOR FURTHER INFORMATION CONTACT:

Thomas P. Stapp, Office of Building

Technology, State and Community

Programs, Department of Energy, Mail

Stop 5E–080, EE–42, Forrestal Building,

1000 Independence Avenue, SW,

Washington, DC 20585, (202) 586–2096.

SUPPLEMENTARY INFORMATION:

I. Introduction and Description of the

Program

II. Review Under Executive Order 12866

III. Review Under Executive Order 12988

IV. Review Under Executive Order 13132

V. Review Under the Paperwork Reduction

Act

VI. Review Under the National

Environmental Policy Act

VII. Review Under the Small Business

Regulatory Enforcement Fairness Act of

1996

VIII. Review Under the Unfunded Mandates

Reform Act of 1995

IX. Review Under the Regulatory Flexibility

Act

X. The Catalog of Federal Domestic

Assistance

I. Introduction and Description of the

Program

On August 24, 1999, the Department

published in the Federal Register an

interim final rule (64 FR 46111) revising

the regulations for its State Energy

Program (SEP or program). This rule

provides for the possibility of certain

activities being funded under the

Special Projects part of the program that

are not permitted under the formula

grant part of the program. The rule also

provides for the specification of any

Special Projects funding limitations by

the sector specific program offices

providing the Special Projects funding,

and clarifies the applicability of Subpart

B to the formula grant part of the

program and of Subpart C to the Special

Projects part of the program.

The program provides formula grants

to States for a wide variety of energy

efficiency and renewable energy

initiatives, and, in years when funding

is available, may also offer financial

assistance for a number of State-oriented

competitively awarded Special Projects

activities with funding contributed by

the Office of Energy Efficiency and

Renewable Energy’s End-Use Sector

Programs. Special Projects have been

funded in every fiscal year since SEP

was established in 1996. DOE expects

the Special Projects part of SEP to

continue in future years.

Among the goals of the SEP Special

Projects activities are to assist States to:

accelerate deployment of energy

efficiency and renewable energy

technologies; facilitate the acceptance of

emerging and under utilized energy

efficiency and renewable energy

technologies; and increase the

responsiveness of Federally funded

technology development efforts to

private sector needs.

The interim final rule published on

August 24, 1999 announced a 30-day

public comment period that closed on

September 23, 1999. We received no

comments regarding the changes made

under 10 CFR part 420, and those

changes are made final. Therefore, this

rule is adopted as it was published in

the program’s interim rule on August

24, 1999 (64 FR 46111).

II. Review Under Executive Order

12866

Today’s regulatory action has been

determined not to be a significant

regulatory action under Executive Order

12866, ‘‘Regulatory Planning and

Review,’’ 58 FR 51735 (October 4, 1993).

Accordingly, this action was not subject

to review under the Executive Order by

the Office of Information and Regulatory

Affairs (OIRA).