List of Approved Spent Fuel Storage Casks: (HI-STAR 100) Addition

NUCLEAR REGULATORY COMMISSION

10 CFR Part 72

RIN 3150-AG17

List of Approved Spent Fuel Storage Casks: (HI-STAR 100) Addition

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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

regulations to add the Holtec International HI-STAR 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 October 4, 1999.

FOR FURTHER INFORMATION CONTACT: Stan Turel, telephone (301) 415-6234,

e-mail [email protected] of the Office of Nuclear Material Safety and

Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-

0001.

SUPPLEMENTARY INFORMATION:

[[Page 48260]]

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 International HI-STAR 100 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 together

with the Safety Analysis Report (SAR) entitled ``HI-STAR 100 Cask

System Topical Safety Analysis Report (SAR), Revision 8.'' 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 International HI-STAR 100 cask system. The NRC

published a proposed rule in the Federal Register (64 FR 1542; January

11, 1999) to add the HI-STAR 100 cask system to the listing in 10 CFR

72.214. The comment period ended on March 29, 1999. Nine comment

letters were received on the proposed rule.

Based on NRC review and analysis of public comments, the 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 International HI-STAR 100 cask system.

The staff has also modified its preliminary SER and has revised the

title of the SAR in the listing of this cask design in 10 CFR 72.214.

The title of the SAR has been revised to delete the revision number

so that in the final rule the title of the SAR is ``HI-STAR 100 Cask

System Topical Safety Analysis Report.'' This revision conforms the

title to the requirements of new 10 CFR 72.248, recently approved by

the Commission.

The proposed CoC has been revised to clarify the requirements for

making changes to the CoC by specifying that the CoC holder must submit

an application for an amendment to the certificate if a change to the

CoC, including its appendices, is desired. This revision conforms the

change process to that specified in 10 CFR 72.48, as recently approved

by the Commission. The CoC has also been revised to delete the proposed

exemption from the requirements of 10 CFR 72.124(b) because a recent

amendment of this regulation makes the exemption unnecessary (64 FR

33178; June 22, 1999). In addition, other minor, nontechnical, changes

have been made to CoC 1008 to ensure consistency with NRC's new

standard format and content for CoCs. Finally, extensive comments were

received from Holtec International and 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. Staff agrees with many of these suggested changes and

has incorporated them into the final documents, as appropriate.

The NRC finds that the Holtec International HI-STAR 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 International HI-STAR 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 International HI-STAR 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

October 4, 1999. 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 nine comment letters on the proposed rule. The

commenters included the applicant, the State of Utah, an individual

member of the public, industry representatives, and several utilities.

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 Direct Final Rule

As part of the proposed rule, the NRC staff requested public

comment on the use of a direct final rulemaking process for future

amendments to the list of approved spent fuel storage casks in 10 CFR

72.214. The direct final rulemaking process is used by Federal

agencies, including the Environmental Protection Agency (EPA) and the

NRC, to expedite rulemaking where the agency believes that the rule is

noncontroversial and significant adverse comments will not be received.

Use of this technique in appropriate circumstances has been endorsed by

the Administrative Conference of the United States (60 FR 43110; August

18, 1995). Under the direct final rulemaking procedure, the NRC would

publish the proposed amendment to the 10 CFR 72.214 list as both a

proposed and a final rule in the Federal Register simultaneously. A

direct final rule normally becomes effective 75 days after publication

in the Federal Register unless the NRC receives significant adverse

comments on the direct final rule within 30 days after publication. If

significant adverse comments are received, the NRC publishes a document

that withdraws the direct final rule. The NRC then addresses the

comments received as comments on the proposed rule and subsequently

issues a final rule.

One commenter supported use of the direct final rule process for

future revisions to the listing in 10 CFR 72.214, stating that it was

imperative that the regulatory process be streamlined when there is no

adverse safety concern. Two commenters were opposed to use of a direct

final rule process stating that a direct final rule would diminish the

public role in commenting on the approval of spent nuclear fuel casks

and thereby the public's ability to affect the outcome of rulemaking

procedures. One of these commenters believed that, given past problems

with the casks, future approval should be subject to adequate and

rigorous public scrutiny.

[[Page 48261]]

Those opposed also believed that 30 days (as would be allowed in a

direct final rule process) is not sufficient time to prepare comments

that may be significantly adverse so as to cause the NRC to withdraw

the published final rule. The two commenters did not believe that an

addition to or revision of the listing is likely to be either

noncontroversial or routine as evidenced by the number of comments they

had on the Holtec HI-STAR 100 proposed rule.

A number of significant adverse comments were received on the NRC's

proposed listing of the Holtec International HI-STAR 100 cask system

which are described in subsequent sections of this notice. Therefore,

it does not appear that the direct final rule approach can be

implemented at this time for additions to the cask listing. The NRC

will reassess this issue in the future after experience with more new

listings to 10 CFR 72.214 has been gained. However, with respect to

amendments to existing CoCs, the NRC anticipates that, except in

unusual cases, the direct final rulemaking process can be used because

the cask design and analysis will have gone through the public comment

process for the initial CoC listing and the revision will be limited to

the subject of the amendment. Unless the NRC has reason to believe that

a particular amendment will be controversial, the NRC plans to use a

direct final rule for amendments to the cask systems in the 10 CFR

72.214 listing. The NRC disagrees that use of the direct final

rulemaking procedure will limit the public's ability to affect the

outcome of the rulemaking. Receipt of a significant adverse comment

will cause the direct final rule to be withdrawn and the comment to be

considered as though received in response to a proposed rule. Further,

the NRC believes that 30 days is a sufficient amount of time in which

to submit a comment on an amendment to the CoC for a listed cask since

most issues related to the cask design will have been resolved in the

rulemaking conducted to place the design on the 10 CFR 72.214 list.

Comments on the Holtec International HI-STAR 100 Cask System

The comments and responses have been grouped into five areas:

general comments, cladding integrity, health impacts, sabotage events,

thermal requirements, and miscellaneous items. Several of the

commenters provided specific comments on the draft CoC, the NRC staff's

preliminary SER, the TSs, and the applicant's Topical SAR. Some of the

editorial comments have been grouped as well as some of the comments on

the drawings in the SAR. To the extent possible, all of the comments on

a particular subject are grouped together. The listing of the Holtec

International HI-STAR 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:

General Comments

Comment No. 1: One commenter asked a number of questions about the

process for review and approval of spent fuel storage cask designs, and

suggested changes to the process.

Response: The NRC finds these comments to be beyond the scope of

the current rulemaking which is focused solely on whether to place a

particular cask design, the Holtec International HI-STAR 100 cask

system, on the 10 CFR 72.214 list.

Comment No. 2: One commenter stated that the cask should be built

and tested before use at reactors, including the loading and unloading

procedures. The commenter objected to the use of computer modeling and

analysis.

Response: The NRC disagrees with the comment. The HI-STAR 100

Storage Cask System Design has been reviewed by the NRC. The basis of

the safety review and findings are clearly identified in the SER and

CoC. Testing is normally required when the analytic methods have not

been validated or assured to be appropriate and/or conservative. In

place of testing, the NRC staff finds acceptable analytic conclusions

that are based on sound engineering methods and practices. NRC accepts

the use of computer modeling codes to analyze cask performance. The

appropriateness of the computer codes and models used by Holtec are

addressed in the SER and Topical SAR. The NRC staff has reviewed the

analyses performed by HOLTEC and found them acceptable. No changes to

the CoC, TSs, SER, or Topical SAR are recommended. These models are

based on sound engineering sciences and processes.

Comment No. 3: One commenter requested that a troubleshooting

manual be prepared that includes information on how many of what type

cask are loaded, where and how long they have been loaded, and on

problems that have occurred, and the solutions. The commenter is

seeking basic information that is periodically updated.

Response: This comment is beyond the scope of this rulemaking.

Cladding Integrity

Comment No. 4: One commenter noted that Holtec's conclusion that

fuel rod integrity will be maintained under all accident conditions is

based on the fact that the HI-STAR 100 system is designed to withstand

a maximum deceleration of 60 g, while a Lawrence Livermore National

Laboratory Report (UCID-21246, Dynamic Impact Effects on Spent Fuel

Assemblies, Chum, Witt, Schwartz (October 20, 1987)) (LLNL Report)

shows that the most vulnerable fuel can withstand a deceleration of 63

g in the most adverse orientation (side drop). The commenter believes

that Holtec and the NRC staff have not demonstrated a reasonable

assurance that the cladding will maintain its integrity because

Holtec's analysis does not take into account the possible increase in

rate of oxidation of cladding of high burnup fuel, and oxidation may

cause the cladding to become effectively thinner, decreasing its

structural integrity and lowering the ``g'' impact force at which fuel

cladding will shatter. With respect to a possible increase in rate of

oxidation of cladding, Holtec has not factored the information in

Information Notice (IN) 98-29, ``Predicted Increase in Fuel Rod

Cladding Oxidation'' (August 3, 1998) into its calculations. The clear

implication of IN 98-29, in the commenter's view, is that the lift

height of the HI-STAR 100 cask must be reduced to lower the ``g''

impact forces on the cladding. Also, the commenter provided a table,

``Effects of Changing Variables in Dynamic Impact Effects on Spent Fuel

Assemblies,'' which the commenter believes shows that the maximum ``g''

impact force, that high burnup fuel with oxidized cladding can

withstand, approaches 45 g.

Response: The NRC disagrees with the comment. Information Notice

98-29 states that high burn-up conditions may increase fuel rod

cladding oxidation. The increased rate of oxidation is a function of

the fuel burn-up and will only affect cladding in high burn-up fuel

applications. In general, fuel with a burn-up exceeding 45,000 MWD/MTU

is considered to be a high burn-up fuel. However, the Holtec HI-STAR

100 Storage Cask System is not authorized to contain fuel with a burn-

up exceeding 45,000 MWD/MTU. Fuel cooling and the average burn-up

approved for the HI-STAR 100 Storage Cask System is: (a) for MPC-24 PWR

assemblies, the fuel burn-up is limited to 42,100 MWD/MTU; and (b) for

MPC-68 BWR assemblies, the fuel burn-up is limited to 37,600 MWD/MTU.

Therefore, the potential for significant amounts of

[[Page 48262]]

oxidized cladding is not a concern for the HI-STAR 100 Storage Cask

System, and the table provided by the commenter regarding the

consequences of significantly oxidized fuel cladding is not relevant to

the approved contents of this cask design.

Comment No. 5: The same commenter stated that Holtec's SAR for the

HI-STAR 100 storage cask relies upon the LLNL report for its estimate

of ``g'' impact force that will damage fuel cladding but that the LLNL

report fails to take into account the increased brittleness of

irradiated fuel assemblies. Because the irradiated fuel assemblies may

have been embrittled, they would also be less resistant to impact.

During the course of a fuel assembly's life, subatomic particle

bombardment, including neutron flux, significantly decreases the

assembly's ductility and increases the assembly's yield stress, thereby

embrittling the fuel assembly.

The HI-STAR 100 design cannot rely on LLNL's analysis, in the

commenter's view, because the LLNL analysis does not account for

irradiation and embrittlement, which lower the impact resistance of the

fuel assemblies. These facts are significant when coupled with the

increased oxidation rate reported in IN 98-29 because increased

oxidation could tangentially cause an increase in cladding

embrittlement. Thus, IN 98-29 compounds the LLNL's error in

disregarding the brittle characteristics of irradiated fuel cladding.

Response: The NRC disagrees with the comment. The LLNL Report, as

referred to, considers the effects of irradiation on cladding. Table 3

of the report delineates irradiated cladding longitudinal tensile tests

on coupon specimens. These test specimens were machined from the

cladding. The effects of irradiation will increase the Young's modulus

and yield stress but decrease the ductility of the cladding. Figure 5

of the report shows that the total elongation values for zircaloy do

not change significantly with strain rate and that the ductility

appears to be independent of the level of the g-loading. Further,

Figure 5 of the report shows that the yield strength is consistently

lower than the tensile strength which suggests that significant margin

exists between yielding of the cladding and gross rupture. The

allowable ``g'' impact force calculation in the report is based on the

yield stress. Thus, the approach that is used in the LLNL Report and

reflected in the SAR is conservative and acceptable.

Comment No. 6: The same commenter stated that Holtec's calculations

rely upon the LLNL report's erroneous assumption that the fuel within

the cladding behaves as a rigid rod. Thus, Holtec merely used a static

calculation for impact analysis versus a dynamic calculation. This

assumption is incorrect, in the view of the commenter. Instead of a

homogenous, rigid rod, the fuel rod consists of fuel pellets stacked

like coins within thin tubing. In any impact scenario, the fuel

assembly acts as a dynamic system with the fuel impacting the inside of

the cladding and creating a greater likelihood of cladding rupture.

Holtec has not shown that the assumption of a rigid rod is

conservative. The thinner cladding due to the increased oxidation

serves to compound this effect because a smaller ``g'' force would be

required to rupture the assembly.

Response: The NRC disagrees with the comment. The assertion that

the fuel rod consists of fuel pellets stacked like coins within thin

tubing is incorrect for irradiated fuels. The fuel pellets are densely

packed inside the fuel tubing, and the effects of irradiation will bond

the pellets to each other and to the fuel cladding. Samples of

irradiated fuel rods have shown that it is indeed nearly impossible to

separate the fuel pellets and the cladding.

It is incorrect to assume the fuel rod acts as a dynamic system

with the fuel pellets impacting the inside of the fuel rod cladding

during an accident drop event. The fuel pellets are densely packed

inside the fuel tube and, for irradiated fuels, the fuel pellets are

bonded together and to the cladding. The LLNL Report discussed above

has conservatively neglected the contributions of the fuel pellets to

fuel rod rigidity. Rather, the report only considers the cladding for

calculating the allowable g-load. It is true that the LLNL Report used

static calculations to derive the allowable g-load equivalent to the

dynamic impact loading. During an accident drop event, the fuel

assembly is subjected to dynamic impact loading and the equivalent

static g-load is determined by a dynamic analysis. The equivalent

static g-load is then shown to be lower than the allowable g-load to

ensure the fuel cladding integrity is maintained. The approach is well

established and acceptable. Therefore, the NRC staff has found Holtec's

accident analysis to be conservative as reflected in SER Chapter 11 and

is therefore acceptable.

Comment No. 7: One commenter stated that the calculated health

impacts under hypothetical accident conditions discussed in Chapter 7

of Holtec's HI-STAR 100 SAR are not 100 percent conservative. Holtec's

original hypothetical design basis accident condition assumed that 100

percent of the fuel rods are nonmechanically ruptured and that the

gases and particulates in the fuel rod gap between the cladding and

fuel pellet are released to the multi-purpose canister (MPC) cavity and

then to the external environment. The accident analysis in the final

version increased the amount of radioactivity to the MPC cavity by 5

orders of magnitude in accordance with NUREG-1536, and would have

placed doses at 100 m over the EPA's limit of 5 rem. An assumed small

leakage rate by the applicant reduced the amount released from the cask

cavity to the environment by more than 5 orders of magnitude. This

design basis accident no longer represents a loss-of-confinement-

barrier accident as originally described.

Response: The NRC disagrees with the comment. The hypothetical

accident dose calculation is appropriate. As discussed in Interim Staff

Guidance (ISG)-5, Rev. 1, ``Normal, Off-Normal, and Hypothetical

Accident Dose Estimate Calculations for the Whole Body, Thyroid, and

Skin,'' the hypothetical accident assumes 100 percent fuel rod failure

within the MPC cavity and release of radioactivity based on factors

from NUREG/CR-6487. The applicant demonstrated that the HI-STAR 100

confinement boundary (MPC) remains intact from all credible accidents.

Therefore, there is not a credible loss-of-confinement-barrier accident

for the HI-STAR 100. The hypothetical accident leakage is

conservatively assumed to be equal to that assumed for normal condition

leakage with corrections for accident pressures and temperatures. The

normal condition leak rate is specified in TS 2.1.1.

The NRC believes that there is reasonable assurance that the

confinement design is adequately rigorous and will remain intact under

the normal and accident conditions identified by the applicant.

Therefore, the design basis change has been found to be conservative

and meets applicable regulations.

Comment No. 8: One commenter requested the criteria for an intact

fuel assembly, the number of pinhole leaks, blisters, hairline cracks,

and crud. The commenter asked if a visual inspection is required and

stated that just performing visual exam was inadequate.

Response: As proof that the fuel to be loaded is undamaged, the NRC

will accept, as a minimum, a review of the records to verify that the

fuel is undamaged, followed by an external visual examination of the

fuel assembly before loading to identify any obvious damage. For fuel

assemblies where

[[Page 48263]]

reactor records are not available, the level of proof will be evaluated

on a case-by-case basis. The purpose of this demonstration is to

provide reasonable assurance that the fuel is undamaged or that damaged

fuel loaded in a storage or transportation cask is confined (canned).

The criteria for intact assembly are defined in TS Section 1.1 as being

fuel assemblies without known or suspected cladding defects greater

than pinhole leaks or hairline cracks and which can be handled by

normal means. Partial fuel assemblies (fuel assemblies from which fuel

rods are missing) shall not be classified as intact fuel assemblies

unless dummy fuel rods are used to displace an amount of water greater

than or equal to that displaced by the original fuel rods.

Radiation Protection

Comment No. 9: One commenter stated that Holtec calculated the

radiation dose to an adult 100 meters from the accident due solely to

inhalation of the passing cloud without considering other relevant

pathways, such as direct radiation from cesium and cobalt-60 deposited

on the ground, resuspension of deposited radionuclides, ingestion of

contaminated food and water, and incidental soil ingestion, and does

not reflect 10 CFR 72.24(m).

Response: The NRC agrees that Holtec calculated the radiation dose

to an adult 100 meters from the accident due solely to inhalation of

the passing cloud and did not consider direct radiation and ingestion.

The NRC staff considers inhalation to be the principal pathway for

radiation dose to the public, and Holtec has followed NRC staff

guidance in making conservative assumptions regarding the source term

and duration of the release. In SER Chapter 10, the NRC staff found

that the radiation shielding and confinement features of the cask

design are sufficient to meet the radiation protection requirements of

10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106. Section 72.106

addresses postaccident dose limits.

When a general licensee uses the cask design, it will review its

emergency plan for effectiveness in accordance with 10 CFR 72.212. This

review will consider interdiction and remedial actions to monitor

releases and pathways based on the chosen site conditions and the

location. Therefore, the pathways identified by the commenter will be

addressed in the general licensee's site specific review.

Comment No. 10: One commenter stated that Holtec has not

specifically calculated potential radiation dose to children, and this

does not meet NRC regulations. Further, the commenter stated that NRC's

methodology for calculating the potential dose to children is

deficient.

Response: The NRC disagrees with the comments. While Holtec did not

specifically calculate potential radiation dose to children, the

international community and the Federal agencies (including EPA and the

NRC) agree that the overall annual public dose limit, from all sources,

should be 1 mSv (100 mrem) which is protective of all individuals. The

purpose of the public dose limit is to limit the lifetime risk from

radiation to a member of the general public. Variation of the

sensitivity to radiation with age and gender is built into the

standards which are based on a lifetime exposure. A lifetime exposure

includes all stages of life, from birth to old age. For ease of

implementation, the radiation standards, that are developed from the

lifetime risk, limit the annual exposure that an individual may

receive. Consequently, the unrestricted release limit of 0.25 mSv (25

mrem), a small fraction of the annual public dose limit, is protective

of children as well as other age groups 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 May 21, 1991,

Federal Register notice (56 FR 23387).

Comment No. 11: One commenter asked if the streaming dose rates

have been measured and if not, will they be measured on the first cask

loading?

Response: There is no NRC regulatory requirement to measure

streaming dose rates at the first cask loading. Further, the applicant

did not provide measured dose rates from cask streaming in its

application because it was not required. The applicant did provide

calculated streaming dose rates in the SAR shielding analysis. The HI-

STAR 100 system is designed to eliminate significant streaming paths,

and each user is required to operate the HI-STAR 100 under a 10 CFR

Part 20 radiological program. NRC has reasonable assurance that the

general licensee's radiological protection and ALARA program will

detect and mitigate exposures from any significant or unexpected

radiation fields for each cask loading.

Comment No. 12: One commenter stated that the applicant should have

performed a specific analysis for off-normal conditions for confinement

analysis and should have included an ``85K'' (Kr-85) dose

calculation to the skin.

Response: The NRC agrees. The applicant should have done an off-

normal condition confinement analysis; however, the off-normal case

dose is approximately a factor of 10 greater than normal dose. The

Holtec normal condition results show acceptable doses when the factor

of 10 is applied for off-normal conditions and have been found

acceptable as reflected in the SER. No additional action is necessary

to meet applicable NRC regulations.

Comment No. 13: One commenter stated that the licensees' report on

specific site doses to the public should be included in the PDR.

Response: The dose for a site-specific location is beyond the scope

of this rulemaking. Licensees are required to meet the dose restriction

in 10 CFR Part 20.

Comment No. 14: One commenter asked for a definition of inflatable

annulus seal. The commenter further questioned the checks and criteria

for surface contamination.

Response: The inflatable annulus seal, which is discussed in

Sections 1.2.2.1, 8.1, and 10.1.4 of the SAR, is designed to prevent

radionuclide contamination of the exterior MPC while the cask is

submerged in a contaminated spent fuel pool. The space between the MPC

and overpack is filled with clean water and is sealed at the top of the

MPC with the inflatable annulus seal. After the seal is removed, the

upper accessible portion of the MPC is examined for contamination to

verify that the seal remained intact during underwater loading. NRC

found the seal description and operation to be acceptable. Each general

licensee will develop site-specific operating procedures that address

the use of the inflatable annulus seal. Each general licensee will also

operate the HI-STAR 100 under a 10 CFR Part 20 radiological protection

program.

Comment No. 15: One commenter suggested that there should be

criteria for the distance of dose measuring mechanism from the cask and

personnel during loading and unloading.

Response: NRC disagrees with this suggestion because NRC

regulations do not specifically require these criteria for dose

measurement. Each general licensee is required to operate the HI-STAR

100 under a 10 CFR Part 20 radiological program and must develop site-

specific operating procedures that include radiological protection dose

surveys that must be conducted during loading and unloading operations.

[[Page 48264]]

Sabotage Events

Comment No. 16: One commenter stated that the current sabotage

design basis is not a bounding accident and that the NRC should

consider the effect of a sabotage event with an anti-tank missile.

There is a lack of a comprehensive assessment of the risks of sabotage

and terrorism against nuclear waste facilities and shipments. The NRC

staff could impose additional conditions on dry storage casks and

Independent Spent Fuel Storage Installations (ISFSIs), e.g., the CoC

could require that an ISFSI be designed with an earthen berm to remove

the line-of-sight.

The commenter stated that since the early 1980s, the NRC has relied

on and poorly interpreted an outdated set of experiments carried out by

Sandia National Laboratory and Battelle Columbus Laboratories that

measured the release of radioactive materials as a result of cask

sabotage. The NRC has never estimated the economic and safety

implications of a sabotage event at a fixed storage facility. Following

the publication of these Sandia study results, the NRC proposed

elimination of a number of safety requirements for shipments of spent

fuel. At least 32 parties submitted more than 100 pages of comments in

response to the notice, to which the NRC never publicly responded. The

NRC suspended action on the rulemaking but inappropriately continues to

use the unrevised conclusions in the proposed rule as a basis for its

policies on terrorism and sabotage of nuclear shipments.

Response: The NRC disagrees with the comment. The NRC reviewed

potential issues related to possible radiological sabotage of storage

casks at reactor site ISFSIs in the 1990 rulemaking that added subparts

K and L to 10 CFR Part 72 (55 FR 29181; July 18,1990). NRC regulations

in 10 CFR Part 72 establish physical protection requirements for an

ISFSI located within the owner-controlled area of a licensed power

reactor site. Spent fuel in the ISFSI is required to be protected

against radiological sabotage using provisions and requirements as

specified in 10 CFR 72.212(b)(5). Further, specific performance

criteria are specified in 10 CFR Part 73. Each utility licensed to have

an ISFSI at its reactor site is required to develop physical protection

plans and install systems that provide high assurance against

unauthorized activities that could constitute an unreasonable risk to

the public health and safety.

The physical protection systems at an ISFSI and its associated

reactor are similar in design features to ensure the detection and

assessment of unauthorized activities. Alarm annunciations at the

general license ISFSI are monitored by the alarm stations at the

reactor site. Response to intrusion alarms is required. Each ISFSI is

periodically inspected by NRC, and the licensee conducts periodic

patrols and surveillances to ensure that the physical protection

systems are operating within their design limits. It is the ISFSI

licensee who is responsible for protecting spent fuel in the casks from

sabotage rather than the certificate holder. Comments on the specific

transportation aspects of the cask system and existing regulations

specifying what type of sabotage events must be considered are beyond

the scope of this rulemaking.

Comment No. 17: One commenter asked whether an evaluation for a

truck bomb sabotage event has been conducted.

Response: The staff has evaluated the effects of a truck bomb

located adjacent to storage casks. 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). Each utility licensed

to have an ISFSI at its reactor site is required to develop physical

protection plans and install a physical protection system that provides

high assurance against unauthorized activities that could constitute an

unreasonable risk to the public health and safety. The physical

protection systems at an ISFSI and its associated reactor are similar

in design to ensure the detection and assessment of unauthorized

activities. Response to intrusion alarms is required. Each ISFSI is

periodically inspected by NRC, and the licensee conducts periodic

patrols and surveillances to ensure that security systems are operating

within their design limits. The NRC believes that the inherent nature

of the spent fuel and the spent fuel storage cask provides adequate

protection against a vehicle bomb, and has concluded that there are no

safety concerns outside the controlled area.

Thermal Requirements

Comment No. 18: One commenter stated that the CoC temperature

limits for the storage cask are deficient because they do not take into

account a minimum pitch or center-to-center distance between casks to

be stored in the ISFSI. Further, Holtec has not performed rigorous

calculations to support the assigned pitch of 12-foot or 4-foot spacing

between casks based on the amount of detail in its nonproprietary

version of its analyses.

Response: The NRC disagrees with the comment. In Section 4.4.1.1.7

of the SAR, Holtec addressed the heat transfer interaction between the

overpacks for a cask array at an ISFSI site. No forced convection was

assumed (e.g. stagnant ambient conditions which would maximize the

interaction heat effect). The applicant further adjusted the heat

transfer in accordance with ANSYS methodology and applied it in the

calculations. Further, in SER Section 4.5.2.1, the NRC staff noted that

the applicant considered in its temperature calculations that multi-

purpose cask baskets were loaded at design basis maximum heat loads,

and systems were considered to be arranged in an ISFSI array and

subjected to design basis normal ambient conditions with insulation.

The NRC staff concluded in the SER that it has reasonable assurance

that the spent fuel cladding will be protected against degradation by

maintaining the clad temperature below maximum allowable limits.

Miscellaneous Items

Comment No. 19: One commenter asked why a coating without zinc was

not required for the VSC-24 cask design. The commenter further

questioned why NRC allowed coatings to be applied to casks because it

will create problems for future DOE waste disposal.

Response: NRC regulations do not prohibit the use of coatings in a

cask design. An applicant must provide information in its safety

analysis report to support use of coatings. The applicant should

describe the near and long term effects of the coatings on systems

important to safety including the benefits and potential impacts of

coating use. Based on the applicant's analysis, the NRC reviews and

assesses the use and adequacy of the coatings. Specific comments

relating directly to VSC-24 are beyond the scope of this rulemaking.

Comment No. 20: One commenter asked why the current HI-STAR 100 is

not an ASME stamped component.

Response: NRC regulations do not require an ASME stamp for a cask.

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.'' Applicant submittals are reviewed to the criteria in

the Standard Review Plan. Cask fabrication activities are inspected by

the licensees and the NRC staff to

[[Page 48265]]

ensure that components are fabricated as designed.

Comment No. 21: One commenter asked a number of questions related

to the Boral and NS-4-FR concerning (1) Whether it has been used ``over

time'' in a cask, (2) the amount of ``creep or slump'' that has

occurred over time, (3) how the testing is conducted, and (4) how the

Boral content is tested in the panels. The commenter further asked if

fabrication is inspected and why no surveillance or monitoring program

is required to check the Boral content.

Response: The questions and comments on the Boral neutron absorber

are addressed in Sections 6.4.2 and 9.1.4 of the SER and Sections

1.2.1.3.1, 6.3.2, and 9.1.5.3 of the SAR. The NRC routinely accepts the

use of Boral as a neutron absorber for storage cask applications, and

it has been used in casks. NRC has approved both storage and

transportation cask designs that use Boral. Section 1.2.1.3.1 of the

SAR describes the historical applications and service experience of

Boral. This information indicates that Boral has been used since the

1950's and used in baskets since the1960's. Several utilities have also

used Boral for nuclear applications such as spent fuel storage racks.

Based on industry experience, no credible mechanism for ``creep or

slump'' of Boral in the cask has been identified.

Sections 1.2.1.3.1 and 9.1.5.3 of the SAR describe the testing

procedures for Boral. Boral will be manufactured and tested under the

control and surveillance of a quality assurance and quality control

program that conforms to the requirements of 10 CFR Part 72, Subpart G.

A statistical sample of each manufactured lot of Boral is tested by the

manufacturer using wet chemistry procedures and/or neutron attenuation

techniques.

The Boral is designed to remain effective in the HI-STAR 100 system

for a storage period greater than 20 years and there are no credible

means to lose the Boral. Further, the NRC accepts the use of NS-4-FR as

a neutron absorber for storage cask applications, and it has been used

in other casks. Therefore, surveillance and monitoring are not needed.

Comment No. 22: One commenter provided a discussion on the VSC-24

design. The issues included materials, the use of coatings, the use of

March Metalfab as a fabricator, calculations being performed when

problems are being solved, testing of soils and pads, and cask handling

temperatures.

Response: These comments are beyond the scope of the current

rulemaking.

Comment No. 23: One commenter asked how the prepossession or

anodization of aluminum surfaces is checked and what the criteria were

for the inspection.

Response: The NRC disagrees that an inspection is necessary. The

only aluminum used in the MPC-24 or MPC-68 is for the Boral neutron

absorbers. Aluminum forms a very thin, adherent film of aluminum oxide

whenever a fresh cut surface is exposed to air or water, becoming

thicker with increasing temperatures and in the presence of water

(Source: ``Corrosion Resistance of Aluminum and Aluminum Alloys,''

Metals Handbook, Desk Edition, American Society for Metals, 1985).

Thus, no inspection or acceptance criteria are necessary.

Comment No. 24: One commenter requested clarification on whether

the helium will be pure and not mixed with krypton or xenon that would

have an effect on internal pressure or temperature. The commenter also

asked whether the helium had to be dry.

Response: Only pure helium will be used to backfill the cask; no

krypton or xenon gasses will be added during backfill. Technical

Specification Table 2-1, Footnote 1, specifies that helium used for

backfill of MPC shall have a purity of 99.995%. Acceptable

helium purity for dry spent fuel storage was defined by R. W. Knoll et

al. at Pacific Northwest Laboratory (PNL) in ``Evaluation of Cover Gas

Impurities and Their Effects on the Dry Storage of LWR Spent Fuel,''

PNL-6365, November 1987. Helium purity is addressed in SAR Section

8.1.4, MPC Fuel Loading, Step 28, and SER Section 8.1.3.

Comment No. 25: One commenter asked whether leakage of gases,

volatiles, fuel fines, and crud was considered credible and whether the

analysis addressed this concern.

Response: The applicant has calculated the postulated annual dose

at 100 meters assuming a realistic leakage rate consistent with ANSI

N14.5 Standard ``Leakage Tests on Packages for Shipment for Radioactive

Materials'' (1997) and has reflected the results in SAR Chapter 7. The

applicant's analysis addresses the commenter's concern, and the

calculated dose had been found to be within regulatory guidelines

(limits) and acceptable to the NRC staff.

Comment No. 26: One commenter was concerned that the cask could

drop or tip over in the loading area of the plant and whether this has

been evaluated. The commenter was also concerned about a drop or tip

over during transfer from the pad or during transport and that all of

the analysis seemed to be for the pad.

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

part of the structural design basis for the HI-STAR 100 cask design.

Holtec described drops and tipover analyses in SAR Section 3.4.9. The

NRC's evaluation of the vendor's analyses is described in SER Sections

3.2.3.1 and 3.2.3.2. The NRC found the results of these analyses to be

satisfactory in that the calculated stresses were within the allowable

criteria of the American Society of Mechanical Engineers (ASME) Code.

Before using the HI-STAR 100 casks, the general licensee must evaluate

the foundation materials to ensure that the site characteristics are

encompassed by the design bases of the approved cask. The events listed

in the comment are among the site-specific considerations that must be

evaluated by the licensee using the cask.

Comment No. 27: One commenter asked whether the design has been

evaluated for a seismic event during loading and unloading.

Response: The HI-STAR 100 casks can only be wet loaded and unloaded

inside the fuel handling facility. Generally, these activities take

place in a segregated under-water cask loading pit which would limit

cask movement during a seismic event. The cask will be supported for a

seismic event during loading and unloading. General procedure

descriptions for these operations are summarized in Sections 8.1 and

8.3 of the SAR. Detailed loading and unloading procedures are developed

and evaluated on a site-specific basis by the licensee using the cask.

Comment No. 28: One commenter questioned whether the method for

cooling has been tested with a real cask.

Response: The NRC regulations and guidance in the Standard Review

Plan require the review and approval of the design criteria. No testing

is required for approval of the design under this current rule. The

cask user is required to perform preoperational testing to determine

the effectiveness of the cooling methods.

Comment No. 29: One commenter questioned whether the manufacturer's

literature for the ``high emissivity'' paint on the overpack had been

evaluated and tested, how the testing was done, and what the results

were. The commenter also questioned whether/how the painted components

were safely stored. The commenter further stated that the paint on the

surfaces of the overpack should be a specified paint, not just a

requirement of ``an emissivity of no less than 0.85.''

Response: The manufacture and application of high-emissivity paints

is

[[Page 48266]]

not a new technology. Several manufacturers provide paints with

specified emissivity ratings. Thermal tests are required to confirm the

heat transfer capabilities of the inner and intermediate shells and

radial channels. Annual cask inspection will check the exterior surface

conditions at which time the paint will be examined and touched up in

local areas as necessary. The NRC does not believe that identifying a

specific brand name of paint is required. There are several suppliers

who manufacture paints with the specified emissivity. The NRC has

reviewed the applicant's analysis and found that paints with an

emissivity greater than 0.85 are acceptable.

Comment No. 30: One commenter questioned the drain down time and

asked how frequently the water is checked. The commenter requested

information on what happens if the MPC can't be vacuum dried

successfully and when the fuel needs to be put back in the spent fuel

pool.

Response: The drain down time is not specified in the TSs but is

part of the vacuum drying procedure. The TSs state that the vacuum

drying must be completed within 7 days. There is not a specific

procedure in the application to monitor the water content; however,

that will be addressed by the cask user on a site-specific basis and is

beyond the scope of this rulemaking. If the drying process is

unsuccessful and the TS requirements cannot be met within 30 days, the

fuel assemblies must be moved from the cask and be placed in the spent

fuel pool.

Comment No. 31: One commenter requested information on the cask

storage array on the pad and the radiation affect from other casks in a

full cask array. The commenter further requested information on how the

applicant/certificate holder/licensee will examine and/or test the HI

STAR 100 and who was actually responsible for the test. The commenter

questioned whether a domed cask cover would be better for runoff and

sky shine concerns.

Response: The applicant performed a shielding analysis that

included a three-by-three cask array (square) model to simulate the

average dose contribution from the center cask, which is partially

shielded by the surrounding periphery casks. This value is applied in

an offsite dose formula used to estimate offsite doses from every cask

in the array. The center-to-center cask pitch was assumed to be 12 feet

in the shielding analyses. Testing of the actual as-installed

configuration will be performed by the cask user and will be evaluated

at that time. Offsite dose estimates for a typical ISFSI array,

including the affects of multiple casks and skyshine, are discussed in

Sections 5.4.3 and 10.4.1 of the SAR. NRC found the dose estimates to

be acceptable. As required in 10 CFR 72.212, each general licensee will

perform a site-specific dose evaluation to demonstrate compliance with

Part 72 radiological requirements. The general licensee will identify

an ISFSI configuration and may elect to use additional engineered

features of its choosing, such as shield walls, a domed cover, or

berms, to ensure compliance with radiological requirements. Section

1.4.7 of Appendix B to the CoC requires that any such engineered

feature be considered important to safety and evaluated to determine

the applicable quality assurance category.

Comment No. 32: One commenter questioned what the criteria were for

the polyester resin ``poured'' into radial channels, how they were

tested, handled and inspected, and whether they had been tested in a

real cask. The commenter questioned whether a ``poured'' neutron shield

was really safe and whether uncontrolled voids caused a problem with

occupational dose requirements. The commenter stated that poured

neutron shields should not be used.

Response: The NRC has reviewed Holtec's application that described

the neutron shielding to be used to meet the requirements of 10 CFR

72.104 and 72.106. The NRC found the Holtec approach acceptable. The

methods for testing, handling, and inspecting installation of the

shielding are beyond the scope of this rulemaking. However, poured

neutron shielding has been successfully used in other cask designs.

Comment No. 33: One commenter stated that appropriate limits for

burnup should be specified in the CoC. The commenter is concerned that

the SAR analysis assumed significantly higher burnups than allowed and

significantly higher initial uranium loading than specified in the

table.

Response: Burnup, cooling time, initial uranium loading, and

initial enrichment are parameters that affect the total source term

(radioactivity) of spent fuel. The applicant's source term analysis

assumed higher uranium loadings and higher burnups than those specified

in TSs of the CoC. Therefore, the radiological source term is

conservative relative to the allowed burnups and uranium loadings.

As discussed in Section 5.2.1 of the preliminary SER, for the same

level of burnup, neutron source terms typically increase as initial

enrichment decreases. Therefore, the source term analysis employed

lower-than-average enrichment values. Based on the SAR analyses,

conditions of the CoC, and other requirements in Parts 20 and 72, the

NRC has determined that minimum enrichment is not warranted as an

additional operating control for the HI-STAR 100. Specific reasons for

this determination include the following: (1) the enrichments bound a

significant portion of spent fuel, and the source terms are calculated

for burnups significantly higher than those allowed in the CoC; (2) the

radiological source terms are adequately controlled in the CoC by

limits on maximum burnup, minimum cooling time, maximum initial uranium

loading, and maximum decay heat; (3) dose rates are controlled in the

CoC by specific dose limits for the top and side of the cask that are

based on values calculated in the shielding analysis; (4) each general

licensee will perform a site-specific dose evaluation to demonstrate

compliance with Part 72 radiological requirements; and (5) each general

licensee will operate the ISFSI under a Part 20 radiological protection

program.

NRC agrees with the comment that the preliminary SER term of ``low

probability'' may not provide definite criteria for general license

cask users regarding limitations on minimum enrichment. Therefore,

Chapter 5 of the SER has been revised to clarify that minimum

enrichment is not an operating control for the HI-STAR 100.

Comment No. 34: One commenter asked what has been considered as

credible ways to lose the fixed neutron poisons.

Response: The NRC staff does not consider the loss of fixed neutron

poisons to be credible after they are installed into the cask because

the poisons are fixed in place and contained.

Comment No. 35: A commenter questioned how the welds of the MPC lid

and closure ring are tested and asked for the acceptance criteria.

Response: Information on the welds is contained in SAR Tables

9.1.1, 9.1.2, and 9.1.3.

Comment No. 36: One commenter asked whether shims are used and

stated that shims or gaps were not acceptable.

Response: There are no shims used in the closure weld of the HI-

STAR 100 casks. The only shims used are located between the canister

and the overpack at basket support locations to provide additional

support for the basket supports. The actual thickness of the shim will

depend on the gaps between the cask and the inside cavity of the

overpack at the basket support locations. Gaps between separate

components such as the cask and the

[[Page 48267]]

overpack are unavoidable and are necessary to ensure that there will be

no physical interferences and to allow free thermal expansions.

Comment No. 37: One commenter stated that all welds should be

monitored unless they have been tested.

Response: NRC accepts welded closure of casks. The regulations do

not require monitoring or testing of welds because there are no

expected degradation mechanisms identified during the cask usage life.

However, both the fabricator and cask user will examine and inspect all

welds as appropriate.

Comment No. 38: One commenter stated that the detailed loading and

unloading procedures developed by each cask user should be put in the

PDR.

Response: Loading and unloading procedures are site-specific issues

not required for design approval and are beyond the scope of this

rulemaking.

Comment No. 39: One commenter asked how long before an ultrasonic

testing examination is conducted should the equipment be calibrated.

Response: Comments on the site-specific examination techniques and

associated calibration are beyond the scope of rulemaking for the HI-

STAR 100 system.

Comment No. 40: One commenter was concerned over the possibility

that the bolts could rust and crack over time or become brittle and

crack because water, ice, and frost could get into the bolt holes over

the years.

Response: The NRC disagrees with this concern over the integrity of

the bolting material. The 54, 1\5/8\-inch-diameter, closure plate bolts

are made from ASME SB-637-N07718 material per SAR BM-1476. N07718, a

nickel-chromium alloy, does not become brittle at colder temperatures.

N07718 is a high strength, corrosion resistant material used in

applications with a temperature range from -423 deg.F (-253 deg.C) to

1300 deg.F (704 deg.C) (Source: Inconel Alloy 718, Inco Alloys

International, fourth edition, 1985). This material will not rust,

unlike carbon steels in corrosive environments. In addition, the

material retains significant ductility down to -320 deg.F (-196

deg.C) as shown by impact test results (Source: Inconel Alloy 718,

Table 27). Therefore, the NRC has no concerns about the bolting

material.

Comment No. 41: One commenter asked what type of radiographic exam

is applicable and where it would be conducted.

Response: SAR Tables 9.1.1, 9.1.2, and 9.1.3 describe which

radiographic exams are to be performed and when they are required to be

performed.

Comment No. 42: One commenter disagreed with allowing the use of a

penetrant test in lieu of volumetric examination on austenitic

stainless steels because flaws in these are ``not expected'' to exceed

the thickness of the weld head. The commenter believes that volumetric

welds should be required because if you don't know for sure the real

size of the actual weld, how can you accept a certain flaw size? The

commenter asked how the permanent record is kept and stated that black

and white photographs should be used as a permanent record.

Response: NRC disagrees with this comment. The NRC position on

inspection of closure welds is contained in ISG-4, ``Cask Closure Weld

Inspections.'' Actual cask welds are examined in accordance with site-

specific procedures that are beyond the scope of rulemaking for the HI-

STAR 100 system. Nondestructive Examination (NDE) methods are specified

in accordance with Section III ``Rules for Construction of Nuclear

Power Plant Components,'' and Section V ``Nondestructive Examination,''

of the ASME Code and are already described in SAR Tables 9.1.1, 9.1.2,

and 9.1.3. 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.

Comment No. 43: One commenter believed that the marking material

for the casks should be designated and that the mark needed to be

permanent.

Response: NRC agrees with the comment. The storage marking

nameplate is made from a 4-inch by 10-inch, 14-gauge Type 304 stainless

steel sheet and welded to the outside of the HI-STAR 100 Overpack.

Lettering will be etched or stamped on the plate. Details are shown in

SAR Drawing 1397, Sheet 4 of 7, and described in SER Section 9.1.6. The

nameplate will provide appropriate cask identification that will last

well beyond the design life of the HI-STAR 100 system. No nonpermanent

marking will be used.

Comment No. 44: One commenter requested information on ``rupture

disc replacements,'' how they are tested for replacement, what the time

criteria are, and what is considered a rupture.

Response: The rupture disc is located in the neutron shield tank of

the HI-STAR 100 casks. The purpose of the rupture disc is to limit

pressure build-ups to a precalculated level within the neutron shield

tank during the fire accident condition. When the pressure build-up

exceeds the precalculated design pressure, the disc will rupture to

relieve the pressure. The rupture disc is tested and certified by the

manufacturer. There is no regulatory requirement for the replacement of

rupture discs. The SAR has arbitrarily set a replacement schedule for

every 5 years to assure functionality.

Comment No. 45: One commenter asked if the casks are checked in

winter for ice and snow loads or ice around the base and if the pads

will be kept clean.

Response: Casks are designed for the worst ice and snow loads

possible. Ice build-ups around the cask base are not allowed, and the

pad will be kept clean. Site-specific procedures will address these

items.

Comment No. 46: One commenter questioned if there was an evaluation

for a plane crash, with a fuel fire, into a cask or full cask array

conducted and whether there is a stipulation as to putting a pad in an

area where planes regularly fly.

Response: Before using the HI-STAR 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 cask as

required by 10 CFR 72.212(b)(3). The licensee's site evaluation should

consider the effects of nearby transportation and military activities.

Generally, a cask's inherent design will withstand tornado missiles and

collision forces imposed by light general aviation aircraft (i.e.,

1500-2000 pounds) that constitute the majority of aircraft in operation

today. The events listed in the comment are among the site-specific

considerations that must be evaluated and are beyond the scope of this

rulemaking.

Comment No. 47: One commenter questioned why Holtec stated that the

HI-STAR 100 could be part of the final geologic disposal system.

Response: The NRC is not reviewing this design for use in a final

geologic disposal system, but only for interim storage under Part 72.

Comment No. 48: One commenter asked where the MPC shell weld is

located and if the pocket trunnions at the bottom of the overpack have

been analyzed specifically for tipovers and falls.

Response: The MPC shell has multiple welds located both

longitudinally on the side of the MPC and circumferentially on the top

and bottom of the MPC. The pocket trunnions at the bottom overpack have

been analyzed by the applicant for tipovers and falls. The NRC reviewed

the design for normal, off-normal, and

[[Page 48268]]

accident conditions, and found it acceptable.

Comment No. 49: One commenter stated that the lifting and pocket

trunnions should be checked over the years for cracking or brittleness

and for debris accumulation and should be kept ready for use over the

years.

Response: The NRC agrees with this comment. As shown in SAR Table

9.2.1, lifting trunnion and pocket trunnion recesses are visually

inspected before the next handling operation after HI-STAR 100 casks

are placed on the ISFSI pad. The trunnion material has been evaluated

for brittle fracture and found to be satisfactory for the operating

temperature range. In addition, the trunnions are load tested in

accordance with ANSI N14.6, ``American National Standard for

Radioactive Materials--Special Lifting Devices for Shipping Containers

Weighing 10000 Pounds (4500 kg) or More.'' Thus, there is no credible

reason to suspect undetected cracking or brittleness. The pocket

trunnion recess is closed by a pocket trunnion plug during storage.

There is no possibility of animal and bird access and nesting in the

recess.

Comment No. 50: One commenter requested information on the criteria

for the critical flaw size.

Response: The criteria for critical flaw size are included in ISG

No. 4, ``Cask Closure Weld Inspections.'' The NRC review determined

that Holtec's proposed methodology is consistent with this ISG.

Comment No. 51: One commenter asked how subcontractors are to be

audited and inspected.

Response: This comment is beyond the scope of this rulemaking.

Comment No. 52: One commenter believed that the first cask for each

utility should be tested at a full heat load and asked what is meant by

the ``First System In Place'' requirement.

Response: The heat transfer characteristics of the cask system will

be recorded by temperature measurements for the first HI-STAR 100

systems (MPC-24 and MPC-68) placed into service with a heatload greater

than or equal to 10 kW. An analysis shall be performed by the cask user

that demonstrates that the temperature measurements validate the

analytical methods and the predicted thermal behavior described in

Chapter 4 of the SAR.

The cask user will perform validation tests for each subsequent

cask system that has a heat load that exceeds a previously validated

heat load by more than 2 kW (e.g., if the initial test was conducted at

10 kW, then no additional testing is needed until the heat load exceeds

12 kW). No additional testing is required for a system after it has

been tested at a heat load greater than or equal to 16 kW.

The cask user will provide a letter report to the NRC in accordance

with 10 CFR 72.4 summarizing the results of each of these validation

tests. Cask users may also satisfy these testing and reporting

requirements by referencing validation test reports submitted to the

NRC by other cask users with identical designs and heat loads.

Comment No. 53: One commenter asked how much water is to be drained

under the MPC lid before welding and how the temperature enters into

the calculations.

Response: Chapter 8 of the SAR directs the operators to pump

approximately 120 gallons of water from the MPC before commencing

welding operations. The water level is lowered to keep moisture away

from the weld region. Under these conditions, ample water remains

inside the MCP to maintain cladding temperatures well below their short

term limits. This operating condition has been evaluated by the NRC.

The resulting temperature increase is much less than any previously

analyzed accident condition might produce.

Comment No. 54: One commenter asked how lifting height should be

verified and stated that the height should be recorded.

Response: The maximum lifting height maintains the operating

conditions of the Spent Fuel Storage Cask (SFSC) within the design and

analysis basis. It is the general licensee's responsibility to limit

the SFSC lifting height to allowable values. The lift height

requirements are specified in TS LCO 2.1.7 for the vertical and

horizontal orientations. Surveillance requirements require verification

that SFSC lifting requirements are met after the SFSC is either

suspended or secured in the transporter and prior to moving the SFSC

within the ISFSI.

Comment No. 55: One commenter questioned how the MPC closure ring,

lid, vent, and drain covers are removed during unloading and what

precautions are taken.

Response: The specific procedures for removal of the closure ring,

lid, vent, and drain covers are to be developed by the cask user. These

procedures will be evaluated by the licensee and by the NRC during

inspections to address adequacy and implementation and, therefore, are

beyond the scope of this rulemaking.

Comment No. 56: One commenter questioned that if the MPC gas

temperature is not met, what additional actions are required and have

they been evaluated (TS B3.1.8-3)?

Response: The NRC staff has evaluated this condition. The TSs

require that if the MPC gas temperature is exceeded during unloading,

no additional operational actions may be conducted until the

temperature is restored to below the TS limit.

Comment No. 57: One commenter asked if ``dry'' unloading operations

are considered.

Response: A dry unloading operation was not requested or explicitly

described in the SAR and thus is not currently allowed for the HI-STAR

100 system and is beyond the scope of this rulemaking.

Comment No. 58: One commenter questioned if crud disposal is a

problem and how it can be mitigated.

Response: Dispersal of crud is beyond the scope of this rulemaking

and is a site-specific issue. Experience with wet unloading of some

fuel types after transportation has involved handling significant

amounts of crud. However, the NRC notes that the HI-STAR generic

unloading procedures mitigate crud dispersal. As discussed in Section

8.3.1 of the SAR, these procedures include gas sampling of the MPC

internal atmosphere and specific cool-down steps. Each cask user will

develop additional site-specific unloading procedures based on its

radiological protection program to further address and mitigate crud

dispersal.

Comment No. 59: The applicant made comments relevant to the helium

backfill pressure of the cask. After discussions with the NRC staff,

Holtec withdrew this comment during a telephone conversation on 5/7/99.

Response: Not applicable.

Comments on Proposed TSs

Upon review of the public comments received on the proposed TSs for

the HI-STAR-100 Storage Cask, particularly comments received from EXCEL

Corporation and the Holtec Users Group, the NRC staff has determined

that several structural changes to the TSs were in order. These changes

result in a clearer set of TSs and move the TSs from the new generation

of dual-purpose cask systems toward a standardized format.

Comment No. 60: It was suggested that controlling the bases for the

TSs as part of the CoC would result in administrative burdens to all

involved. These bases are not controlled as part of power reactor

licenses.

Response: The NRC staff agrees. Therefore, the bases have been

relocated to an appendix to the SAR.

[[Page 48269]]

Comment No. 61: A number of commenters also raised concerns with

the inclusion of the extensive fuel specifications (formerly Section

2.0) and a very lengthy design specification section (formerly Section

4.0).

Response: The NRC staff agrees that placement of much of this

information in the TSs is unwarranted. Therefore, much of the

information regarding fuel specifications and some of the design and

codes information were moved from the TSs to a separate appendix to the

CoC. However, the NRC staff did maintain some of the information

regarding requirements for bases controls by adding it to a revised

Section 3.0, ``Administrative Controls and Programs,'' of the TSs.

Upon consideration of public comments and further consideration

within the NRC, the NRC staff has determined that the structure of TS

Section 2.1, ``SFSC INTEGRITY,'' did not provide appropriately clear

guidance. Therefore, the NRC staff has revised this section of the TSs

to reflect a more logical and focused approach. The number of limiting

conditions for operations (LCOs) in this section has been reduced to

four. The NRC staff believes that this will enhance the usefulness of

the TSs.

Comment No. 62: One commenter stated that if surface contamination

exceeds 2200 dpm/100 cm2 from gamma and beta emitting sources, and

smearable contamination limits cannot be reduced to acceptable levels,

the TSs require actions up to and including removal of the MPC from the

HI-STAR 100 overpack after removing the spent fuel from the MPC. The

commenter stated that the proposed Skull Valley ISFSI in Utah does not

have facilities for decontaminating casks and, therefore, these TSs

could not be met.

Response: The NRC agrees in part. The revised version of the TSs

(TS 2.2.2) requires verification that removable contamination is within

limits during loading operations and provides up to 7 days to restore

the contamination within limits. The specifications no longer list MPC

or spent fuel removal actions. Further, comments on the proposed site-

specific Skull Valley ISFSI currently under review are beyond the scope

of this rulemaking. Decontamination requirements will be reviewed as

part of the site-specific licensing provisions under Part 72 Subpart B

for the Skull Valley ISFSI.

Comment No. 63: One commenter stated that the definition of

``TRANSPORT OPERATIONS'' needs to be revised to reflect that the drop

analysis is not limited to drops from the transporter, and that lifting

of a cask with other devices is not prohibited. The commenter

recommended similar changes to the definition of ``LOADING OPERATIONS''

and ``UNLOADING OPERATIONS.''

Response: The NRC disagrees. The definitions of the three terms in

question do not prohibit lifting of a cask with other devices (the

revised note in TS 2.1.3 clarifies this issue), nor do the definitions

affect the lifting requirements contained in TS 2.1.3.

Comment No. 64: One commenter stated that it would increase the

standardization of the TSs by relocating the explanatory information of

the defined terms in TS Section 1.0 to the TS Bases.

Response: The NRC disagrees with the comment. The terms defined in

TS Section 1.0 are important in the understanding of the TS

requirements. These definitions need to be contained within the TSs.

This practice is consistent with the standard TSs developed for the

U.S. nuclear power reactors.

Comment No. 65: One commenter stated that in Examples 1.3-2 and

1.3-3, the word ``action'' should be capitalized.

Response: The NRC agrees. The word ``action'' has been capitalized.

Comment No. 66: One commenter recommended the removal of portions

of Table 2.1-1 and all of Table 2.1-2 and Table 2.1-3 from the TSs.

Response: The NRC agrees, in part, that this information should be

moved. This design information is crucial to the conclusions reached by

the NRC staff in its SER; therefore, the design information contained

in these tables has been relocated (and renumbered) to a separate

appendix to the CoC, along with other critical design information.

Comment No. 67: One commenter recommended a change to the format of

the Titles of Tables 2.1-1, 2.1-2, 2.1-3, and 2.1-4.

Response: The NRC agrees with the comment. The format has been

changed.

Comment No. 68: One commenter recommended a wording change in TS

Section 3.0 from ``not applicable to an SFSC'' to ``not applicable.''

Response: The NRC agrees with this comment and has made the

indicated change.

Comment No. 69: One commenter stated that there is no need to

create two specifications for TS 3.1.1, MPC Cavity Vacuum Drying

Pressure, and TS 3.1.2, OVERPACK Annulus Vacuum Drying Pressure. In

addition, the commenter indicated there is no need to create two

specifications for TS 3.1.5, MPC Helium Leak Rate, and TS 3.1.6,

OVERPACK Helium Leak Rate.

Response: The NRC agrees with the comment. Section 2.1 of the TSs

has been revised based on these and similar comments received to

combine these TSs.

Comment No. 70: One commenter stated that the frequency of SR

3.1.7.1 should be revised because, as written, the frequency would

apply only when a cask is being moved to or from the ISFSI and would

not apply at other times, such as when moving casks within the ISFSI.

However, the drop analysis applies any time the cask is suspended. The

frequency should be revised similar to ``Prior to movement of an

SFSC.''

Response: The NRC agrees with the comment. The frequency of SR

3.1.7.1 has been revised.

Comment No. 71: One commenter recommended that TS Sections 4.1 and

4.2 be eliminated because they contain no unique information.

Response: NRC agrees with the comment. Sections 4.1 and 4.2 have

been eliminated.

Comment No. 72: One commenter recommended relocating the

information contained in TS Sections 4.3 and 4.5 to the SAR, and

recommended eliminating TS Section 4.4, stating that this section is a

duplication of existing regulatory requirements.

Response: The NRC agrees in part. The NRC staff agrees that these

sections do not belong in the TSs. This design information has been

relocated to Appendix B to the CoC. The NRC staff disagrees with the

commenter's proposal to eliminate or relocate these sections to the

SAR. The NRC has relocated these sections to Appendix B to the CoC due

to the importance of the design information contained in these

sections. The NRC staff also disagrees with the comment that TS Section

4.4 is a duplicate of existing regulations, since this section contains

the acceptance criteria for the site-specific design parameters.

Comment No. 73: A commenter recommended relocating the information

contained in TS Sections 4.6 and 4.8 to an Administrative Controls

chapter due to their content and relocating Section 4.7 to the SAR

because it is a one-time administrative task.

Response: The NRC agrees in part. The NRC staff agrees that these

sections belong in the administrative section of the TSs and has placed

this information in a new TS Chapter 3.0, ``Administrative Controls and

Programs.'' The NRC staff disagrees with the commenter on the proper

location of Section 4.7 (now TS Section 3.2), because it is established

NRC staff

[[Page 48270]]

practice to place important administrative requirements, even one-time

requirements, in the TSs.

Comment No. 74: A commenter stated that TS 3.1.8 contains conflicts

because the APPLICABILITY statement, and the COMPLETION TIME when the

condition is not met, are the same statement. The commenter further

recommended that because of its complexity and rarity of its use, this

specification be eliminated and the information specified in the SAR.

Response: The NRC agrees in part. The NRC agrees with the first

point. TS 2.1.4 has been rewritten to remove this conflict. The NRC

staff disagrees with the second point and considers this information

important to the proper operation of the cask system. Further, the

changes made to this section resolve concerns regarding its complexity.

Comment No. 75: One commenter recommended relocating the figure

attached to TS 3.2.1 to the TS Bases, because the purpose of the figure

is to show where dose measurements should be taken.

Response: The NRC disagrees with this comment. This figure, now

attached to TS 2.2.1, is an integral part of the proper implementation

of this TS and assures that the dose measurements will be taken at the

proper locations.

Comment No. 76: The commenter stated that the TSs do not comply

with 10 CFR 72.44(d) that requires TSs on radioactive effluents.

Response: The NRC agrees with this comment. TS Section 3.0 has been

revised to incorporate the requirements of 10 CFR 72.44(b).

Comment No. 77: One commenter recommended that within TS Section

1.1, the definition for ``Intact Fuel Assembly'' should be revised to

state `` * * * an amount of water greater than or equal to * * *,''

adding the term ``greater than or'' to allow greater flexibility with

respect to dummy rod sizing.

Response: The NRC agrees with the comment and has revised the

definition.

Comment No. 78: One commenter recommended that within TS Table 2.1-

1, Item II.B should be reworded for clarification because the current

wording could be misinterpreted by users that intact fuel assemblies

are required to be loaded into damaged fuel containers.

Response: The NRC agrees with the comment. The table, which has

been relocated to Appendix B, has been revised.

Comment No. 79: One commenter requested clarification of TS Section

4. As written, the text does not require a written report of the

results of the first measurements, only ``each cask subsequently loaded

with a higher heat load.'' NRC's intent to require a written report for

the first temperature measurements is not clear. The commenter further

stated that it is not clear what ``calculation'' is being referred to

in the last two sentences, whether it is the original design

calculation or a new calculation generated from the test. The commenter

further recommended the addition of ``decay heat'' after ``lesser'' and

before ``loads'' in the last line.

Response: The NRC agrees with these comments, except for the

recommendation to add the phrase ``decay heat,'' which the NRC

considers unnecessary. TS Section 3.3 has been revised to clarify the

reporting requirements and the calculational comparison required by

this TS condition.

Comment No. 80: One commenter recommended some editorial changes to

revise TS Bases 2.2.2 and 2.2.3 to clarify that 10 CFR 72.75 has

additional reporting requirements that may need to be met independent

of these TS requirements.

Response: The NRC agrees with the comment. A reference to 10 CFR

72.75 has been added to Appendix B to the CoC.

Comment No. 81: One commenter recommended adding a new definition

for fuel building to the TSs.

Response: The NRC agrees with the comment. A definition for fuel

building has been added to the TSs.

Comment No. 82: One commenter recommended editorially revising TS

LCO 3.1.7, ``SFSC Lifting Requirements'' and the related bases to

clarify the applicability. The revision is necessary because the LCO is

not intended to be applicable while the transport vehicle is in the

fuel building or when the cask is secured on a railcar or heavy haul

trailer because the cask is not being lifted.

Response: The NRC agrees with the comment. TS 2.1.3 has been

revised accordingly.

Comment No. 83: One commenter recommended a revision to TS Tables

2.1-2 and 2.1-3, Note 1, for the purposes of clarification and to allow

for manufacturer tolerances.

Response: The NRC agrees with the comment. The recommended changes

to the tables have been made. The table has been relocated to Appendix

B of the CoC.

Comment No. 84: One commenter recommended the revision of TS Table

3-1, Item 1.c, to change the lower helium tolerance to 10 percent

because the smaller tolerances were associated with convection heat

transfer, for which no credit is taken in the application.

Response: The NRC agrees with the comment and has revised

renumbered TS Table 2-1.

Comment No. 85: One commenter recommended that TS 4.3.1 be revised

to allow for changes to codes and standards because it would provide

both the vendor and the NRC the flexibility to add exceptions/

alternatives to the code without amending the certificate.

Response: The NRC agrees with the comment. Section 1.3.2 of

Appendix B has been revised accordingly.

Comment No. 86: The applicant recommended in TS Section 4.4.6, the

revision of the soil effective modulus of elasticity from

``6,000psi'' to ``28,000 psi.'' In addition, the

commenter recommended an acceptable method for licensees to comply with

the soil modulus limit.

Response: The NRC agrees with the comment. The information has been

added to Appendix B to the CoC.

Comment No. 87: One commenter recommended the addition of a third

option to TS LCO 3.1.7 and Bases B3.1.7 (or elsewhere in the TSs) that

allows general licensees to calculate site-specific lifting

requirements based on the site-specific pad design and associated drop/

tipover analyses.

Response: The NRC agrees with the comment. TS LCO 2.1.3 has been

revised to add this option.

Comment No. 88: One commenter believed that the 48-hour time limit

within TSs 3.1.1 through 3.1.6 is overly restrictive.

Response: The NRC agrees with this comment in part. Accordingly,

the NRC has reviewed the time limit in each applicable TS. Some of the

time limits have been extended to provide for a controlled, deliberate

response to the LCO condition.

Comment No. 89: One commenter recommended the deletion of the

Design Features, Section 4.6, Training Module, and Section 4.7, Pre-

Operational Testing and Training Exercise because the review of the

training program is required by 10 CFR 72.212(b)(6) and the TS

duplicates the requirement in the regulation.

Response: The NRC agrees in part. The NRC agrees that there is

duplication in the TSs and the regulatory requirements. Accordingly, TS

3.1 (previously Section 4.6) has been modified to reference the general

licensee's systematic approach to training. However, the NRC staff

believes that listing the training exercises as a specific requirement

for proper cask operation is appropriate to

[[Page 48271]]

be included in the TSs, and it has been maintained.

Comment No. 90: One commenter recommended adding ``diesel'' before

``fuel'' in TS Section 4.4.5 and in SER Sections 3.1.2.1.8, 4.3.4, and

4.4.3.4 for clarification.

Response: The NRC agrees conceptually with the comment. TS Section

4.4.5 (now 1.4.5 of Appendix B) and SER Sections 3.1.2.1.8, 4.3.4, and

4.4.3.4 have been revised to refer to combustible transporter fuel.

Comments on the Draft CoC

Comment No. 91: Two commenters recommended that CoC Condition 10 be

revised to be consistent with 10 CFR 72.48 for the cask design and

operating procedures. Another commenter stated that Condition 10 was

not clear.

Response: The NRC agrees with the comments. The applicable CoC

condition has been revised to delete the prescriptive controls for

making changes to the cask design and operating procedures. The

condition now reflects 10 CFR 72.48 as recently approved by the

Commission.

Comment No. 92: Two commenters recommended that a Bases Control

Program be added to the TSs or CoC.

Response: The NRC disagrees with the comment. The proposed TS bases

are part of the SAR. Because 10 CFR 72.48 provides a change process for

the SAR for control of the bases, there is no need to incorporate this

program into the CoC or TSs.

Comment No. 93: One commenter requested information on the status

of a petition for rulemaking on the change process in 10 CFR 72.48.

Response: This comment is beyond the scope of this rulemaking.

Comment No. 94: One commenter stated that the description of the

attachment to the CoC was in error.

Response: The NRC agrees with this comment. The description has

been corrected.

Comments on the NRC Staff's SER

Comment No. 95: One commenter asked a question about what is meant

by the statement included in the NRC SER in Section 9.3 related to the

examination and/or testing of the HI-STAR 100 by the applicant/

certification holder/licensee.

Response: The SER refers to Section 9.1 of the applicant's SAR.

This section summarizes the scope and acceptance criteria for the HI-

STAR 100 test program. It includes fabrication and nondestructive

examinations, weld inspecting, structural and pressure tests, leakage

tests, component tests, and shielding and integrity testing and

controls. The SAR or SER does not specify which entity must perform

each test. This is because some tests are performed during fabrication,

while others can only be performed after installation. The quality

assurance programs implemented by the fabricator, certificate holder,

or applicant with appropriate oversight will ensure that these SAR

specified tests are completed and are effective. Further, the NRC

inspection program also verifies on a sampling basis that tests and

surveillances are conducted as required.

Comment No. 96: One commenter recommended revising the last

sentence of the first paragraph of SER Section 3.1.2.1.6 to read: ``The

design-basis earthquake accelerations are assumed to be applied at the

top of the ISFSI concrete pad with the resulting inertia forces applied

at the HI-STAR 100 mass center.''

Response: The NRC agrees with the comment. The SER has been

revised.

Comment No. 97: One commenter recommended in SER Section 3.1.4.4,

in the first paragraph, the replacement of ``* * * the fabricator is an

accredited facility by the ASME for nuclear fabrication work holding

``N'' and ``NPT'' stamps, * * *'' with ``* * * the HI-STAR 100 System

is designed in accordance with the ASME Code, as clarified by the

exceptions to the Code listed in TS Table 4-1.''

Response: The NRC agrees with the comment. The SER has been

revised. Note that the table is now in Appendix B.

Comment No. 98: One commenter recommended that in SER Section 6.3,

the word ``minimum'' be replaced with ``maximum'' in the third sentence

of the first full paragraph to match the analysis.

Response: The NRC agrees with the comment. The SER has been revised

to correct the error.

Comment No. 99: One commenter stated that SER Section 8.1.4, which

discusses the evaluation of welding and sealing procedures, should be

revised to recognize the option of performing manual welding of the MPC

lid closure weld in accordance with a user's as low as reasonably

achievable (ALARA) practices.

Response: The NRC disagrees with the comment. As discussed in

Sections 8.1 and 10.1 of the SAR, the use of the Automated Weld System

provides justification that the HI-STAR 100 is designed in accordance

with Part 72 radiological requirements and ALARA objectives consistent

with Part 20. However, the intent of the proposed SER revision is

already implied in Section 8.1.2 of the SER that states: ``Each cask

user will need to develop detailed loading procedures that incorporate

the ALARA objectives of their site-specific radiation protection

program.'' Therefore, each user can develop site-specific operating

procedures based on ALARA objectives that would include the use of

manual welding and make changes to the SAR in accordance with 10 CFR

72.48.

Comment No. 100: One commenter recommended that SER Section 8.3.1,

which discusses the evaluation of cooling, venting, and reflooding

during cask unloading operations, should be revised to allow the option

of a once-through purge in lieu of the closed-loop cooling system.

Response: The NRC disagrees with this comment. An amendment

application with a specific design and supporting analysis for a once-

through helium cooling system would be required for NRC review and is

beyond the scope of this rulemaking.

Comment No. 101: One commenter noted that a more appropriate method

to implement the thermal test for the overpack had been accepted by the

NRC for the HI-STAR 100 transportation cask and recommended this method

be used for this cask design. Appropriate changes were recommended to

be made to the SER and SAR.

Response: The NRC agrees that this method should be included in the

SAR for the HI-STAR 100 storage cask. Appropriate changes have been

made to Section 9.1.6 of the SAR and Chapter 9 of the SER.

Comment No. 102: The applicant submitted numerous editorial

comments on the SAR, SER, and CoC. Comments were intended as

clarification, restoration of deleted information, grammatical

corrections, corrections to text, to maintain consistency between

documents, typographical corrections, format changes, and to correct

terminology. These editorial changes do not change the design of the

cask or supporting analysis.

Response: The NRC agrees with many of the editorial comments

suggested by Holtec International. The SAR, SER, and CoC have been

revised to address the comments as appropriate.

Comments on the Applicant's Topical SAR

Note: In response to comments received, a number of changes to

the SAR were made by Holtec International, as discussed below.

Comment No. 103: One commenter proposed a revision to the language

in Section 8.0 of the SAR to clarify that users will have some

flexibility to use

[[Page 48272]]

procedures and equipment suitable for site-specific needs and

capabilities.

Response: The NRC agrees with the suggested editorial changes. The

changes to the SAR have been made.

Comment No. 104: One commenter recommended some editorial changes

within SAR Section 4.4, because the wording in Subsection 4.1.1.15 may

be erroneously interpreted to mean that the chilled helium delivered to

the MPC cavity to cool the internals prior to flooding the cavity with

water must be at 100 deg.F. The commenter stated that the text of the

SAR requires clarification to permit each cask user's cooldown system

to be engineered with the flexibility to cool MPCs containing fuel with

varying levels of decay heat production.

Response: The NRC agrees with the comment. The SAR has been

revised.

Comment No. 105: In SAR Section 1.5, Drawings 1399, Sheet 3, and

BM-1476, and in Drawing Section ``N-N,'' one commenter recommended the

addition of four threaded holes spaced 90 degrees apart as a personnel

dose reduction enhancement. The new holes would allow the personnel

attaching the shield to work in an area of lesser exposure to radiation

within the same time frame. The effect of the shield attachment will

remain the same.

Response: The NRC agrees with the comment. Drawings 1399 and BM-

1476 have been revised to reflect the change.

Comment No. 106: One commenter suggested that in SAR Revision 10,

the drawings in Chapter 1 be revised to match those approved by the NRC

in the transportation SAR.

Response: The NRC agrees with the comment. Seven drawings in SAR

Section 1 have been revised to match those in the transportation SAR.

Although four drawings have not been revised to match the

transportation SAR, this is acceptable to the NRC staff because they

reflect storage design features.

Comment No. 107: In the SAR, one commenter (the applicant)

recommended changing Section 6.1 by replacing ``(20 deg.C-100 deg.)''

with ``(i.e., water density of 1.000 g/cc)'' and delete ``(20 deg.C

assumed)'' to more accurately describe the assumption made in the

analyses.

Response: The NRC agrees. The SAR has been revised as suggested by

the commenter.

Comment No. 108: The applicant suggested a number of changes to the

drawings for the HI-STAR 100 Storage Cask. These changes did not

require a change to the supporting design analyses.

Response: The NRC agrees that the changes to the drawings were

appropriate and do not result in any changes to the supporting design

analyses. The SAR drawings have been revised in accordance with the

suggested changes.

Comment No. 109: The applicant suggested using Magnetic Particle

Examination in lieu of Liquid Penetrant Examination for the overpack

weld examination and recommended changes to the associated drawing

notes.

Response: The NRC agrees with this suggested change. The NRC agrees

that resolution of this comment will involve a change to the drawings

which will mean that drawings referencing this examination shall be

different for the storage and transportation certificates. These

differences are not significant because the staff finds Magnetic

Particle Examination to be equally acceptable to Liquid Penetrant

Examination. Appropriate changes to the drawings have been made.

Comment No. 110: The applicant suggested a clarification for the

sequence for the hydrostatic testing and helium leakage testing during

fabrication of the overpack.

Response: The NRC agrees with the suggested change. The SAR has

been revised accordingly.

Comment No. 111: As it relates to the Radiography and Heat

Treatment requirements for the containment boundary of the HI-STAR

overpack, the applicant requested that post weld heat treatment (PWHT),

after completing nondestructive examination, be used for all overpack

containment boundary welds which require an exception from the ASME

code.

Response: The NRC agrees. The SAR and Appendix B to the CoC have

been modified appropriately.

Comment No. 112: The applicant suggested a revision to the drawings

in the SAR to reflect the localized thinning tolerance in the

containment shell.

Response: The NRC staff agrees with the suggested revision.

However, the applicant did not provide the suggested changes in its

final revisions to the SAR. The initial drawings remain acceptable.

Comment No. 113: One commenter (the applicant) recommended that

changes to Technical Specification Table 4-1, MPC Enclosure Vessel and

Lid, should be made to replace ``and sufficient intermediate layers to

detect critical wild flaws'' with ``and at least one intermediate PT

after approximately \3/8\ inch weld depth.'' The commenter also

recommended the deletion of ``Flaws in austenitic stainless are not

expected to exceed the bead''. The commenter further recommended

several changes to the SER as follows: SER Section 8.1.4 should be

changed to add ``(or optional multi-layer PT examination),'' after

``ultrasonic examination (UT)''; the SER should recognize that users

may choose to perform the MPC void-to-shell weld manually; and SER

Section 11.4.1.3.1 should be reworded to read ``examined using UT or

multi-layer PT techniques,'' instead of ``volumetrically examined using

UT.''

Response: The NRC agrees and notes that the applicant's comments

with respect to TS Table 4-1 have been superseded by its latest

revision to the SAR. Changes have been made to Table 1-3 to Appendix B.

The SER has been revised as recommended.

Summary of Final Revisions

The NRC staff modified the listing for the Holtec International HI-

STAR 100 cask system within 10 CFR 72.214, ``List of approved spent

fuel storage casks,'' with respect to the title of the SAR as well as

the CoC and its two appendices, the TSs, and the Approved Contents and

Design Features. The NRC staff has also modified its SER.

Agreement State Compatibility

Under the ``Policy Statement on Adequacy and Compatibility of

Agreement State Programs'' approved by the Commission on June 30, 1997,

and published in the Federal Register on September 3, 1997 (62 FR

46517), this rule is classified as compatibility Category ``NRC.''

Compatibility is not required for Category ``NRC'' regulations. The NRC

program elements in this category are those that relate directly to

areas of regulation reserved to the NRC by the Atomic Energy Act of

1954, as amended (AEA), or the provisions of Title 10 of the Code of

Federal Regulations. Although an Agreement State may not adopt program

elements reserved to NRC, it may wish to inform its licensees of

certain requirements via a mechanism that is consistent with the

particular State's administrative procedure laws, but does not confer

regulatory authority on the State.

Finding of No Significant Environmental Impact: Availability

Under the National Environmental Policy Act of 1969, as amended,

and the Commission's regulations in Subpart A of 10 CFR part 51, the

NRC has determined that this rule is not a major Federal action

significantly affecting the quality of the human environment and

therefore an environmental impact statement is not required. This final

rule adds an additional cask to the list of

[[Page 48273]]

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 Stan

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

Regulatory Commission, Washington, DC 20555, telephone (301) 415-6234,

e-mail [email protected].

Paperwork Reduction Act Statement

This final rule does not contain a new or amended information

collection requirement subject to the Paperwork Reduction Act of 1995

(44 USC 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-STAR 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

rulemaking 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 rule. Therefore, a backfit analysis is not required.

List of Subjects in 10 CFR Part 72

Criminal penalties, Manpower training programs, Nuclear materials,

Occupational safety and health, Reporting and recordkeeping

requirements, Security measures, Spent fuel.

For the reasons set out in the preamble and under the authority of

the Atomic Energy Act of 1954, as amended; the Energy Reorganization

Act of 1974, as amended; and 5 U.S.C. 553; the NRC is adopting the

following amendments to 10 CFR part 72.

[[Page 48274]]

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 1008 is added to

read as follows:

Sec. 72.214 List of approved spent fuel storage casks.

* * * * *

Certificate Number: 1008

SAR Submitted by: Holtec International

SAR Title: HI-STAR 100 Cask System Topical Safety Analysis Report

Docket Number: 72-1008

Certification Expiration Date: (20 years after final rule effective

date)

Model Number: HI-STAR 100

Dated at Rockville, Maryland, this 23rd day of August, 1999.

For the Nuclear Regulatory Commission.

William D. Travers,

Executive Director for Operations.

[FR Doc. 99-23075 Filed 9-2-99; 8:45 am]

BILLING CODE 7590-01-P