Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices 47227

Name of applicant; date of application, date received, application

number

Description of material

Country of

origin Material type Total

(kilograms)qty End use

Diversified Scientific Services ....

July 23, 1997

July 25, 1997

IW004

Radioactive waste in the form of

liquid products.

15,000,000 For processing into solids & return to Canada.Canada

Diversified Scientific Services ....

July 23, 1997

July 25, 1997

XW002

Radioactive waste in the form of

solids.

15,000,000 Rtn of waste after processing .... Canada.

Dated this 28th day of August 1997 at

Rockville, Maryland.

For the Nuclear Regulatory Commission.

Ronald D. Hauber,

Director, Division of Nonproliferation,

Exports and Multilateral Relations Office of

International Programs.

[FR Doc. 97–23699 Filed 9–5–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–282, 50–306 and 72–10]

Northern States Power Company,

Prairie Island Nuclear Plant and Prairie

Island Independent Spent Fuel Storage

Installation, Issuance of Director’s

Decision Under 10 CFR 2.206

Notice is hereby given that the

Director, Office of Nuclear Reactor

Regulation, has issued a Director’s

Decision concerning a Petition dated

May 28, 1997, filed by the Prairie Island

Indian Community (Petitioners) under

Section 2.206 of Title 10 of the Code of

Federal Regulations (10 CFR 2.206). The

Petition requested that the NRC (1)

determine that Northern States Power

Company (the licensee) violated the

requirements of 10 CFR 72.122(l) by

using its Materials License No. SNM–

2506 for an Independent Spent Fuel

Storage Installation (ISFSI) prior to

establishing conditions for safely

unloading the TN–40 dry storage

containers; (2) suspend Materials

License No. SNM–2506 for cause under

10 CFR 50.100 until such time as all

significant issues in the unloading

process, as described in the Petition,

have been resolved, the unloading

process has been demonstrated, and

until an independent third-party review

of the TN–40 unloading procedure has

been conducted; (3) provide Petitioners

an opportunity to participate fully in the

reviewing of the unloading procedure

for the TN–40 cask, hold hearings and

allow Petitioners to participate fully in

these and any other procedures initiated

in response to the Petition; and (4)

update the Technical Specifications for

the Prairie Island ISFSI to incorporate

mandatory unloading procedure

requirements.

The Director of the Office of Nuclear

Reactor Regulation has determined that

the Petition should be denied for the

reasons stated in the ‘‘Director’s

Decision Under 10 CFR 2.206’’ (DD–97–

18), the complete text of which follows

this notice. The decision and documents

cited in the decision are available for

public inspection and copying in the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

Minneapolis Public Library, Technology

and Science Department, 300 Nicollet

Mall, Minneapolis, MN.

A copy of this decision has been filed

with the Secretary of the Commission

for the Commission’s review in

accordance with 10 CFR 2.206(c). As

provided therein, this decision will

become the final action of the

Commission 25 days after issuance

unless the Commission, on its own

motion, institutes a review of the

decision within that time.

Dated at Rockville, Maryland, this 29th day

of August 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

Director’s Decision Under 10 CFR 2.206

I. Introduction

On May 28, 1997, the Prairie Island

Indian Community filed a Petition

pursuant to Section 2.206 of Title 10 of

the Code of Federal Regulations (10 CFR

2.206) requesting that the U.S. Nuclear

Regulatory Commission (NRC) take

action to accomplish the following:

1. Determine that Northern States

Power (NSP) violated the requirements

of 10 CFR 72.122(l) by using its

Materials License No. SNM–2506 for an

Independent Spent Fuel Storage

Installation (ISFSI) prior to establishing

conditions for safely unloading the TN–

40 dry storage containers;

2. Suspend Materials License No.

SNM–2506 for cause under 10 CFR

50.100 until such time as all significant

issues in the unloading process, as

described herein [the Petition], have

been resolved, the unloading process

has been demonstrated, and until an

independent third-party review of the

TN–40 unloading procedure has been

conducted;

3. Provide Petitioners an opportunity

to participate fully in the reviewing of

the unloading procedure for the TN–40

cask, hold hearings and allow

Petitioners to participate fully in these

and any other procedures initiated in

response to this Petition; and

4. Update the Technical

Specifications (TS) for the Prairie Island

ISFSI to incorporate mandatory

unloading procedure requirements.

The Petition has been referred to me

pursuant to 10 CFR 2.206. The NRC

letter dated June 27, 1997, to Byron

White, on behalf of the Petitioners,

acknowledged receipt of the Petition

and provided the NRC staff’s

determination that the Petition did not

require immediate action by the NRC. A

notice of receipt was published in the

Federal Register on July 3, 1997 (62 FR

36085).

On the basis of the NRC staff’s

evaluation of the issues and for the

reasons given below, the Petitioners’

requests are denied.

II. Background

On October 19, 1993, the NRC issued

Materials License No. SNM–2506 to

NSP (the licensee) to allow storage of

spent nuclear fuel in TN–40 dry storage

casks, designed by Transnuclear

Incorporated, at the ISFSI located at the

Prairie Island Nuclear Plant. No spent

nuclear fuel was allowed to be loaded

into a storage cask at Prairie Island until

several preoperational license

conditions were satisfied. Among the

preoperational license conditions were a

required training exercise (dry-run) of

the loading, handling, and unloading

activities for the TN–40 casks and the

implementation of written procedures

describing the actions to be taken during
47228 Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices

operation, off-normal, and emergency

conditions associated with the Prairie

Island ISFSI. The NRC issued TS

defining operating limits, surveillance

requirements, design features, and

administrative controls as Appendix A

to Materials License No. SNM–2506.

A report dated April 20, 1995,

submitted by the licensee to the NRC

pursuant to 10 CFR 72.82(e), provided

the results of the preoperational tests

that were required to be performed by

the licensee before loading of spent fuel

into a TN–40 cask. On May 11, 1995, the

NRC granted a schedular exemption to

the provision of 10 CFR 72.82(e) that

requires licensees to submit the

preoperational test results at least 30

days before receipt of spent fuel into the

ISFSI. The basis for the exemption was

the fact that the NRC staff had reviewed

cask fabrication records, observed

portions of the preoperational test

activities as they occurred, and

completed its review of the report

submitted on April 20, 1995. On May

12, 1995, the licensee began loading

spent fuel assemblies into a TN–40 cask.

The licensee subsequently placed the

cask, and casks loaded since that time,

onto the storage pad within the Prairie

Island ISFSI.

NRC regulations include a

requirement that an ISFSI be designed

to provide for the ready retrieval of

spent fuel or high-level radioactive

waste for further processing or disposal.

This regulation, 10 CFR 72.122(l),

provides as follows:

Retrievability. Storage systems must be

designed to allow ready retrieval of spent

fuel or high-level radioactive waste for

further processing or disposal.

Certain events or conditions could

warrant removing a TN–40 cask from

the Prairie Island ISFSI and returning it

to the spent fuel pool and unloading the

stored fuel assemblies. In addition to the

regulatory requirements in 10 CFR

72.122(l) pertaining to retrieval of the

fuel assemblies for further processing or

disposal, the TS for the Prairie Island

ISFSI requires the licensee to take

corrective actions in response to those

design-basis events or conditions that

may challenge the integrity of the

storage cask or the cladding of the spent

fuel assemblies. For example, Section

2.3, ‘‘Maximum Cask Lifting Height,’’

Section 3/4.3, ‘‘Maximum Helium Leak

Rate,’’ and Section 3/4.5, ‘‘Maximum

Cask Surface Temperature,’’ of the TS

include provisions for unloading of a

TN–40 storage cask in response to the

specified events or conditions.

NRC regulations in 10 CFR Part 72

require that the design of the storage

system and the procedures implemented

by specific licensees support the

unloading activity, whether it is being

performed to allow further processing or

disposal of the spent fuel or it is

required as part of the response to an

unplanned event or condition, while

preventing gross ruptures of the fuel

cladding in order to prevent operational

safety problems. Unloading procedures

should also include contingencies in

case fuel cladding has degraded during

storage such that additional measures

are necessary to address increased

radiological hazards during the

unloading process.

NRC regulations, facility licenses, and

NRC-approved quality assurance

programs require licensees to establish

and maintain a formal process for the

preparation and issuance of procedures

and changes thereto. NRC assessments

of licensee procedures are generally

conducted as part of the NRC’s

inspection program. In this instance, the

major procedures pertaining to dry cask

storage activities at Prairie Island,

including the procedure for unloading a

cask, were reviewed by the NRC staff

during a special inspection conducted

from January 24 through May 11, 1995.

In addition to the review of the

licensee’s facility and procedures, the

NRC inspectors observed preoperational

testing that the licensee was required to

perform before loading casks with spent

fuel assemblies. The inspection findings

are documented in NRC Inspection

Report 50–282/95002; 50–306/95002;

72–10/95002(DRP), dated June 30, 1995.

The NRC inspectors identified several

instances in which the procedures for

dry cask storage activities that the

licensee had in place at the beginning of

the inspection, including the procedures

for loading and unloading of the TN–40

casks, did not ensure compliance with

the requirements of the license.

Although the inspectors were able to

verify that the licensee corrected the

identified procedural deficiencies

during the course of the inspection, a

Notice of Violation was issued to the

licensee for failing to satisfy Criterion V

of Appendix B to 10 CFR Part 50, which

for activities affecting quality, requires

the preparation and adherence to

procedures appropriate to the

circumstances. In addition, the

inspectors identified weaknesses in the

licensee’s initial performance in

overseeing the activities of the cask

vendor and in overall planning for dry

cask storage activities. However, on the

basis of the licensing reviews and

inspection findings, documented in

Inspection Report 50–282/95002; 50–

306/95002; 72–10/95002(DRP), the NRC

staff concluded that as of May 1995, the

licensee had corrected the identified

deficiencies and was ready to safely

load and, if necessary, unload spent

nuclear fuel in TN–40 casks.

In July 1995, the NRC staff issued an

action plan for dry cask storage to

manage the resolution of a variety of

technical and process issues that were

identified during the licensing reviews

and inspections completed for the first

several ISFSI facilities. An item related

to the loading and unloading of dry

storage casks was added to the action

plan, in part to ensure that the

importance of the unloading procedures

was emphasized to licensees and

technical issues related to unloading

problems were resolved. Addition of an

item pertaining to unloading was

deemed prudent because the staff

observed that some unloading

procedures implemented by licensees

neglected to consider contingencies and

assumptions related to possible fuel

degradation, gas sampling techniques,

cask design issues, radiation protection

requirements, and the thermal-hydraulic

behavior of a cask during the process of

cooling and filling it with water from

the spent fuel pool.

To implement the action plan, the

NRC staff formed a working group to

identify issues associated with loading

and unloading processes for dry storage

casks and to propose means of

informing the industry and the NRC

staff of those issues. The working group

considered industry experiences,

concerns identified during reviews and

inspections, and other issues related to

loading and unloading procedures. The

working group completed its reviews in

April 1996. The concerns related to

unloading procedures reviewed by the

working group were found to involve

either (1) isolated occurrences that had

been adequately resolved by sitespecific corrective actions or (2) generic

issues that were addressed by

incorporating remedial measures into

ongoing staff activities, such as the

preparation of revised inspection

procedures or other guidance

documents.

To fulfill some of the goals included

in its dry cask storage action plan, the

NRC staff has emphasized the

importance of unloading procedures

and shared observations with licensees

using or considering dry cask storage

during opportunities such as the Spent

Fuel Storage and Transportation

Workshop held in May 1996 and

meetings with individual licensees. The

staff revised inspection procedures to

specifically instruct NRC inspectors to

review unloading procedures developed

by licensees and to identify those issues

that warrant particular attention.

Guidance included in NRC Inspection
Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices 47229

Procedure 60855, ‘‘Operation of an

ISFSI,’’ issued February 1, 1996, states:

For unloading activities, attention should

be paid to how the licensee has prepared to

deal with the potential hazards associated

with that task. Some potential issues may

include: the radiation exposure associated

with drawing and analyzing a sample of the

canister’s potentially radioactive atmosphere;

steam flashing and pressure control as water

is added to the hot canister; and filtering or

scrubbing the hot steam/gas mixture vented

from the canister, as it is filled with water.

Similar guidance was included in

NUREG–1536, ‘‘Standard Review Plan

for Dry Cask Storage Systems,’’ issued in

January 1997. Application of the revised

guidance ensures that recent and future

reviews will address the adequacy of

unloading procedures developed by

licensees. The staff also issued NRC

Information Notice 97–51, ‘‘Problems

Experienced with Loading and

Unloading Spent Nuclear Fuel Storage

and Transportation Casks,’’ dated July

11, 1997, to inform licensees of

operating experiences and problems

encountered with the loading and

unloading of storage and transportation

casks for spent nuclear fuel.

To address those ISFSIs that began

operation before the improvements in

the NRC’s review and inspection

guidance, the staff performed audits or

inspections of those licensee programs

for which the inspection record did not

document whether the unloading

procedures adequately addressed the

major issues included in the action

plan. Regarding Prairie Island, the staff

reviewed the available information and

determined that additional reviews or

inspections were not necessary because

the assessment of the unloading

procedure performed as part of the

inspection documented in NRC

Inspection Report 50–282/95002; 50–

306/95002; 72–10/95002(DRP)

adequately addressed the concerns

included in the NRC action plan.

III. Discussion

The Petition requests four actions by

the NRC on the basis of the contention

that the unloading procedure

implemented by the licensee was

inadequate and, therefore, the licensee

violated the NRC regulation requiring it

to have the ability to readily retrieve

spent fuel or high-level radioactive

waste for further processing or disposal.

Item 1: Determine That the Licensee

Violated 10 CFR 72.122(l)

In support of the Petition’s contention

that the licensee violated NRC

requirements, the Petitioners claim that

the procedure to unload a TN–40 cask

at Prairie Island has not been adequately

evaluated or tested because neither the

NRC nor NSP has completely

demonstrated that a TN–40 dry cask can

be unloaded after it has remained on the

storage pad for a number of years. The

Petitioners state that their request is

supported by the fact that the

preoperational test results for the Prairie

Island ISFSI were submitted to the NRC

on the day before the unloading

procedure was approved by the

licensee’s Operations Committee. The

Petitioners also express concern that

only portions of the licensee’s

unloading procedure were tested during

the required preoperational tests and,

therefore, the tests did not provide

assurance that an unloading can be done

safely. In addition, the Petitioners state

that procedures for unloading a cask

should address specific concerns

regarding failed fuel recovery and

possible contamination of the spent fuel

pool, venting of radioactive gases,

functional checks of radiation

monitoring and ventilation systems, and

the build-up of steam when water is

pumped into the cask as part of the

unloading process.

As previously mentioned, cask

designs and associated procedures are

required to support the unloading of the

spent fuel assemblies either to support

further processing or disposal or in

response to an unplanned event or

condition that may challenge the

integrity of the storage cask or the

cladding of the spent fuel assemblies.

Although the NRC staff agrees with the

Petitioners’ premise that actually

unloading a storage cask would likely

result in licensees learning lessons that

could result in additional enhancements

to unloading procedures, the staff does

not agree that an actual demonstration

of the unloading procedure at Prairie

Island is warranted. In addition to the

staff’s review of the procedure for

unloading a TN–40 cask at Prairie

Island, reasonable assurance that the

TN–40 casks can be safely unloaded is

provided by a variety of experiences

related to the use and storage of

radioactive materials. These experiences

include the dry-run exercises that were

performed to verify key aspects of

unloading procedures for the TN–40

cask; related research sponsored by the

commercial nuclear industry, the U.S.

Department of Energy, and the NRC;

actual loading and unloading of

transportation casks; loading of storage

casks; handling of spent fuel assemblies

under various conditions; and

performing relevant maintenance and

engineering activities associated with

reactor facilities.

Regarding the Petitioners’ concerns

pertaining to the dates of the submittal

of preoperational tests and the approval

of the licensee’s unloading procedure,

the NRC staff identified this discrepancy

in Inspection Report 50–282/95002; 50–

306/95002; 72–10/95002(DRP). The

administrative controls included in the

TS for the Prairie Island ISFSI require

that the Operations Committee review

and approve procedures and changes

thereto. The approval of the Operations

Committee is usually the last step in the

process for preparing or revising a

procedure. The fact that the Operations

Committee approved the procedure

shortly after submittal of the

preoperational test results and before

fuel loading satisfied the preoperational

license condition to implement written

procedures before loading spent nuclear

fuel into a TN–40 cask. This matter does

not, therefore, represent a violation of

NRC requirements or introduce

concerns pertaining to the technical

adequacy of the unloading procedure.

The Petitioners identified several

concerns pertaining to the lack of

specific guidance in the unloading

procedure to address a scenario in

which significant fuel degradation

occurs during storage. The NRC staff

agrees with the Petitioners that such a

scenario would complicate the

unloading process by requiring

additional measures and precautions to

limit the release of radioactive materials

from the cask into parts of the reactor

facility and nearby environs. The

licensee’s unloading procedure includes

a step to sample the atmosphere within

the cask cavity to test for radioactive

and flammable gases before venting the

cask cavity and loosening the bolts

securing the cask lid. Following the

analysis of the gas sample, the licensee’s

unloading procedure includes a hold

point to allow personnel to determine

whether additional steps or precautions

are warranted. While acknowledging

many of the Petitioners’ legitimate

concerns regarding the potential

difficulties in retrieving failed fuel from

dry storage casks, the NRC staff has

concluded that licensees need not be

required to incorporate specific

guidance into the normal unloading

procedure to address this unlikely

situation. The staff’s conclusion is

based, in part, on the fact that the

required compensatory actions and

precautions needed to address such

situations may vary significantly,

depending on the actual results from the

analysis of the gas sample. Requiring the

licensee to include contingencies or

steps in the unloading procedure to

address the unlikely event of failed fuel

may unnecessarily complicate and delay

the unloading of fuel assemblies that
47230 Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices

have remained intact during storage. On

the basis of licensees’ experiences in

developing and implementing plans to

address the problem of fuel assemblies

damaged during reactor operations, in

handling radioactive wastes of various

forms, and in resolving other

comparable problems, the NRC staff has

confidence that licensees could, if

necessary, develop a plan to retrieve

damaged fuel from a storage cask while

minimizing the radiological

consequences to plant workers and the

general public. In addition to the

general confidence of the NRC staff that

the technical problems associated with

retrieving failed fuel could be overcome,

requirements for planning and

executing such an activity are contained

in the licenses issued for the Prairie

Island ISFSI and the Prairie Island

Nuclear Generating Plant, and NRC

regulations in 10 CFR Parts 20, 50, and

72. The NRC staff has, therefore,

accepted gas sampling and defined hold

or decision points before breaching the

cask confinement boundary as an

adequate means to address concerns

pertaining to the unlikely degradation of

fuel assemblies during storage.

The specific issues raised by the

Petitioners to support their claim that

the licensee’s unloading procedure is

deficient are addressed below.

(a) Failed Fuel Considerations

As previously discussed, the NRC

staff has accepted that procedures

developed by licensees to support

unloading of dry storage casks do not

need to address the retrieval of failed

fuel provided that measures to detect

possible fuel degradation and a defined

hold point for determination of possible

compensatory actions are appropriately

placed within the subject procedures.

As documented in NRC Inspection

Report 50–282/95002; 50–306/95002;

72–10/95002(DRP), the licensee had

originally failed to incorporate a step in

the unloading procedure for taking a

gaseous sample from the cask in order

to ensure that fuel degradation had not

occurred during storage. However, in

response to the findings of the NRC

inspectors, the licensee incorporated

sampling of the cask atmosphere and a

hold point for deliberation into the

unloading procedure and the revised

procedure was in place before spent

nuclear fuel was loaded into a TN–40

cask. The NRC staff has found that this

action, in combination with the

requirement that spent fuel assemblies

loaded into TN–40 casks be free of gross

cladding defects, provides reasonable

assurance that the licensee will not

unknowingly breach the confinement

boundary of a cask containing failed

fuel. In the unlikely event that the

gaseous sample indicates that spent fuel

assemblies have degraded during

storage, the unloading procedure

instructs the licensee’s Operations

Committee to add steps or precautions

to the procedure in order to minimize

the radiological consequences of

retrieving the failed fuel. The NRC staff

has found this approach to be

acceptable and does not require the

licensee’s normal unloading procedure

to include contingency actions to

address the possible release of

radioactive materials to parts of the

reactor facility, including the spent fuel

pool, that may occur if fuel assemblies

degrade during storage. The NRC staff

believes, however, that the Petitioners

have identified valid concerns regarding

the potential recovery of fuel assemblies

that have unexpectedly degraded during

storage. As previously mentioned, the

staff believes that the regulations and

licenses issued by the NRC require the

licensee to address these and other

problems that may occur in the unlikely

event that fuel assemblies that have

degraded during storage need to be

unloaded from dry storage casks.

(b) Venting of Radioactive Gases

The possible need to vent radioactive

gases from a cask is among the issues

that the licensee would need to address

if the required sampling of the

atmosphere within a cask indicates that

the spent fuel assemblies have

experienced unanticipated degradation

during storage. As with the concern

regarding the contamination of the spent

fuel pool, the need to vent the cask

while minimizing the radiological

consequences of unloading a cask

containing failed fuel is an issue that the

licensee would need to address before

revising the procedure and proceeding

with the unloading process. In addition

to ensuring that the unloading activity

results in occupational doses and doses

to members of the public that are as low

as is reasonably achievable (see 10 CFR

20.1101), the licensee would need to

perform the venting of a cask containing

failed fuel in accordance with the

Prairie Island Nuclear Generating Plant

Facility Operating Licenses, associated

TS, and applicable regulations.

(c) Radiation Monitors

The Petitioners contend that the

unloading procedure must include a

‘‘stop-check’’ to verify that ventilation

systems and radiation monitors are

functioning before the venting of a cask

is performed. Although agreeing with

the Petitioners’ general premise that

prerequisites to preforming procedures

should include establishing confidence

in the tools and equipment being used,

the NRC staff notes that during the

anticipated unloading of spent nuclear

fuel that has not degraded during

storage, special ventilation or radiation

monitoring equipment beyond that

specified in the licensee’s unloading

procedure and radiation protection

program is not required. The unloading

procedure requires the involvement of

radiation protection personnel and the

activity must be controlled in

accordance with the licensee’s radiation

protection program, which includes

provisions for the maintenance and

calibration of radiation detectors.

Although the venting process is not

expected to need ventilation systems

equipped with filters and radiation

monitors, the spent fuel pool special

ventilation system could be used if

necessary. The spent fuel pool special

ventilation system is required to be

operable during subsequent steps in the

procedure if spent fuel assemblies are

being moved and the system must be

tested and maintained in accordance

with the TS for the Prairie Island

Nuclear Generating Plant. In the

unlikely event that the licensee needs to

unload a cask containing degraded fuel

assemblies, confirming the operability

of those ventilation systems and

additional radiation monitoring

equipment being used to minimize the

release of radioactive materials is an

activity that the licensee would need to

address before revising the procedure

and proceeding with the unloading

process.

(d) Steam Build-up

The Petitioners expressed concerns

regarding the reaction of the cask and

stored fuel assemblies to the

introduction of spent fuel pool water

during the execution of the unloading

procedure. The unloading procedure

includes the partial immersion of the

TN–40 cask into the spent fuel pool,

connection of hoses to the vent and

drain connections, and the slow

introduction of spent fuel pool water to

the cask cavity and stored fuel

assemblies. The procedure instructs

personnel to continuously monitor the

temperature and pressure

instrumentation installed on the vent

connection and to stop pumping water

if the pressure exceeds 10 psig or the

temperature exceeds 240 °F. In the

staff’s judgment, the cooling process

imposed by these limitations on

temperatures and pressures at the vent

port of the cask will adequately ensure

that the cooling of the cask and spent

fuel is gradual and, thereby, prevent

safety problems that could

hypothetically result from damage to the
Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices 47231

1The Petitioners request that Materials License

No. SNM–2506 be suspended for cause in

accordance with 10 CFR 50.100. Provisions for the

modification, revocation, or suspension of the

licenses for ISFSI facilities are contained in 10 CFR

72.60. The possible reasons for suspending licenses

for ISFSIs in accordance with 10 CFR 72.60 are

similar to the corresponding reasons for suspending

licenses for production and utilization facilities in

accordance with 10 CFR 50.100.

2Recent NRC staff guidance pertaining to the

appropriate content of technical specifications is

provided in NUREG–1536, ‘‘Standard Review Plan

for Dry Cask Storage Systems,’’ published in

January 1997. Similar guidance is provided by NRC

Regulatory Guide 3.61, ‘‘Standard Format and

Content for a Topical Safety Analysis Report for a

Spent Fuel Dry Storage Cask,’’ issued in February

1989, and NRC Regulatory Guide 3.48, ‘‘Standard

Format and Content for the Safety Analysis Report

for an Independent Spent Fuel Storage Installation

(Dry Storage),’’ issued in October 1981.

cask or the fuel assemblies because of

stresses induced by a poorly controlled

addition of cooling water from the spent

fuel pool.

The Petitioners expressed concerns

pertaining to the range of the

instrumentation used during the venting

of a TN–40 cask and stated that higher

ranges for temperature and pressure are

necessary. The instrumentation ranges

specified in the unloading procedure’s

drawing of the cask vent port adapter

are 50–300 °F for temperature and 0–30

psig for pressure. While not judging if

these are the optimum ranges for the

instrumentation, the NRC staff finds that

the ranges are adequate to support the

administrative limits of 240 °F and 10

psig established in the procedure and

the related response action of stopping

the addition of water to the cask if these

administrative limits are exceeded.

Regarding the Petitioners’ concern

regarding the need to post hazard

warnings during the refilling of a cask,

the unloading procedure does include

several notes and precautions to remind

personnel that the fluid exiting the vent

port may present radiological and

thermal hazards.

In summary, many of the Petitioners’

concerns pertain to potential problems

with unloading spent fuel from a TN–

40 cask if the fuel cladding has

degraded during storage. While

acknowledging that such concerns

regarding the potential difficulties in

retrieving failed fuel from dry storage

casks are legitimate, the NRC staff has

concluded that licensees need not be

required to incorporate specific

guidance into the normal unloading

procedure to address this unlikely

situation. On the basis of its review of

the information provided by the

Petitioners and its reviews of the

licensee’s procedure for unloading TN–

40 casks at Prairie Island, the NRC staff

has not identified violations of 10 CFR

72.122(l) or other regulatory

requirements pertaining to the content

or quality of the licensee’s unloading

procedure.

Item 2: Suspend Materials License No.

SNM–2506

On the basis of the contention that the

licensee’s unloading procedure was

inadequate, the Petitioners requested

that Materials License No. SNM–2506

be suspended until such time as the

significant issues in the unloading

process have been resolved, the

unloading process has been

demonstrated, and an independent

third-party review of the TN–40

unloading procedure has been

conducted.1

As previously stated, the NRC staff

has performed a review of the procedure

for unloading a TN–40 cask at Prairie

Island. The review, including

verification that the licensee’s

unloading procedure was revised to

address deficiencies identified by the

NRC inspectors, is documented in NRC

Inspection Report 50–282/95002; 50–

306/95002; 72–10/95002(DRP). The

review performed during the NRC

inspection, subsequent evaluations

performed by the NRC staff as part of

the activities associated with the dry

cask storage action plan and the review

of this Petition, and the required control

of the procedure in accordance with

licensee programs developed in

accordance with NRC regulations,

facility licenses, and NRC-approved

quality assurance programs provide

reasonable confidence that the licensee

could, if necessary, safely unload a TN–

40 cask.

Regarding a third-party review, the

NRC staff’s concern about the quality of

licensees’ unloading procedures led it to

include the issue in the dry cask storage

action plan. The action plan provided a

framework for the identification and

resolution of various technical and

administrative issues related to the use

of dry storage casks. The previously

mentioned actions taken by the NRC

staff and licensees adequately resolved

the identified issues pertaining to cask

unloading procedures. In the specific

case of the unloading procedure at

Prairie Island, the licensee revised the

procedure to address the problems

identified by the staff during its

inspection. On the basis of the actions

it has already taken, the NRC staff does

not believe that the situation warrants

additional review of the licensee’s

unloading procedure by an independent

third party.

Item 3: Allow Petitioners to Review

Procedure, and for NRC to Hold

Hearings and Allow Petitioners to

Participate in the Proceedings

The licensee has provided the NRC

with the unloading procedure,

including Revision 2, dated November

8, 1996, for placement into the public

record, and the Petitioners have been

supplied with or have obtained copies

of the procedure from the NRC’s

document control system. Accordingly,

Petitioners have had the opportunity to

review a recent revision of the

unloading procedure. For the reasons

previously discussed in this decision,

the NRC staff sees no reason to

undertake additional reviews of the

procedure or to initiate a formal

proceeding in which the Petitioners

could participate. Although the NRC has

decided not to initiate a hearing in

response to this Petition, the Petitioners

are encouraged to continue their

interactions with the NRC staff

regarding concerns or questions about

the operation of the Prairie Island

Nuclear Generating Plant or the Prairie

Island ISFSI.

Item 4: Update the Technical

Specifications for the Prairie Island

ISFSI to Incorporate Mandatory

Unloading Procedure Requirements

The TS for ISFSIs are required, by 10

CFR 72.44, to include requirements in

the following categories:

(1) Functional and operating limits

and monitoring instruments and

limiting control settings;

(2) Limiting conditions;

(3) Surveillance requirements;

(4) Design features; and

(5) Administrative controls.

Although the TS for the Prairie Island

ISFSI requires that TN–40 casks be

unloaded if certain events or conditions

defined in the TS are satisfied, the TS

do not include specific requirements for

the unloading process. The content of

the TS for the Prairie Island ISFSI is

typical in this respect since neither 10

CFR 72.44 nor the associated regulatory

guidance documents specify that

technical specifications should include

special requirements for the unloading

procedure.2 Instead, the functional and

operating limits, limiting conditions,

administrative controls, and other

requirements included in the TS for the

Prairie Island ISFSI are intended to

maintain the cask and stored spent fuel

assemblies within the limits established

for safe operation during storage within

the ISFSI and activities such as loading

and unloading of the casks. For example

TS 2.3 limits the allowable lifting

heights during movement of the cask
47232 Federal Register / Vol. 62, No. 173 / Monday, September 8, 1997 / Notices

from the ISFSI and TS 3/4.2 requires a

measurement of the boron concentration

of the water in the spent fuel pool before

water is introduced to the cask during

the unloading process.

The absence of specific requirements

in the TS to control the unloading

process does not diminish the

importance that the NRC staff places on

this activity or the validity of the

Petitioners’ concerns. The NRC staff

believes that other regulatory

requirements provide an equivalent

level of protection to the Petitioners’

request to include specific requirements

in the TS to control the unloading of a

TN–40 cask. The administrative controls

in the TS for the Prairie Island ISFSI

require that the associated procedures,

including the unloading procedure, be

prepared, reviewed, and maintained in

accordance with the requirements of the

Prairie Island Nuclear Generating Plant

Facility Operating Licenses and

associated TS. In addition, under

existing NRC requirements, the licensee

must adequately implement procedures

to control loading, maintaining, and

unloading of dry storage casks (see 10

CFR 72.122, 10 CFR 72.150, and 10 CFR

72.152). For example, the NRC

inspection documented in Inspection

Report 50–282/95002; 50–306/95002;

72–10/95002(DRP) resulted in a Notice

of Violation issued to the licensee

because the licensee failed to satisfy the

NRC’s requirements in Criterion V of

Appendix B to 10 CFR Part 50 by not

having incorporated appropriate steps

and precautions into the original

procedure developed to control

unloading of a TN–40 cask. As

demonstrated by the example, no

changes to the TS or the Safety Analysis

Report (SAR) are needed to ensure that

enforceable operating controls and

limits are in place to address the

unloading of a cask.

In regard to another concern raised by

the Petitioners, the Prairie Island ISFSI

SAR and other docketed

correspondence do state that unloading

a TN–40 cask would be performed using

a procedure that is basically the reverse

of the procedure used to load the cask.

Although this statement, in a general

sense, is true, the NRC staff agrees with

the Petitioners that such statements may

be misleading in that they oversimplify

the description of the unloading

activity. For this reason, the NRC staff

included an item related to unloading

procedures in its dry cask storage action

plan to ensure that actual unloading

procedures did not reflect such an

oversimplified representation. The

unloading procedure for the dry storage

casks at Prairie Island was inspected by

the NRC staff and, as previously

discussed, was ultimately found to

provide adequate guidance to control

the unloading process.

IV. Conclusion

For the reasons described above, the

NRC has determined that no adequate

basis exists for granting the Petitioners’

request for suspension of Northern

States Power Company’s license for dry

cask storage of spent nuclear fuel at

Prairie Island or for taking the other

actions requested by the Petitioners.

While acknowledging that the

Petitioners’ concerns regarding the

potential difficulties in retrieving failed

fuel from dry storage casks are

legitimate, the NRC staff has concluded

that licensees need not be required to

incorporate specific guidance into the

normal unloading procedure to address

this unlikely situation.

A copy of this decision will be filed

with the Secretary of the Commission

for the Commission to review in

accordance with 10 CFR 2.206(c).

As provided by this regulation, this

decision will constitute the final action

of the Commission 25 days after

issuance, unless the Commission, on its

own motion, institutes a review of the

decision within that time.

Dated at Rockville, Maryland, this 29th day

of August 1997.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–23696 Filed 9–5–97; 8:45 am]

BILLING CODE 7690–01–P

NUCLEAR REGULATORY

COMMISSION

Regulatory Guide; Issuance,

Availability

The Nuclear Regulatory Commission

has issued a new guide in its Regulatory

Guide Series. This series has been

developed to describe and make

available to the public such information

as methods acceptable to the NRC staff

for implementing specific parts of the

Commission’s regulations, techniques

used by the staff in evaluating specific

problems or postulated accidents, and

data needed by the staff in its review of

applications for permits and licenses.

Regulatory Guide 3.70, ‘‘Use of Fixed

Neutron Absorbers at Fuels and

Materials Facilities,’’ provides guidance

that is acceptable to the NRC staff on

procedures for preventing criticality

accidents by using fixed neutron

absorbers in operations involving

handling, storing, and transporting

special nuclear fuels at fuels and

materials facilities.

Comments and suggestions in

connection with items for inclusion in

guides currently being developed or

improvements in all published guides

are encouraged at any time. Written

comments may be submitted to the

Rules and Directives Branch, Division of

Administrative Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555.

Regulatory guides are available for

inspection at the Commission’s Public

Document Room, 2120 L Street NW.,

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regulatory guides, both active and draft

guides, may be obtained free of charge

by writing the Office of Administration,

Attn: Printing, Graphics and

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standing order basis. Details on this

service may be obtained by writing

NTIS, 5285 Port Royal Road,

Springfield, VA 22161. Regulatory

guides are not copyrighted, and

Commission approval is not required to

reproduce them.

(5 U.S.C. 552(a))

Dated at Rockville, Maryland, this 9th day

of August 1997.

For the Nuclear Regulatory Commission.

Malcolm R. Knapp,

Acting Director, Office of Nuclear Regulatory

Research.

[FR Doc. 97–23698 Filed 9–5–97; 8:45 am]

BILLING CODE 7590–01–P

DEPARTMENT OF TRANSPORTATION

Aviation Proceedings, Agreements

Filed During the Week of August 29,

1997

The following Agreements were filed

with the Department of Transportation

under the provisions of 49 U.S.C. 412

and 414. Answers may be filed within

21 days of date of filing.

Docket Number: OST–97–2862.

Date Filed: August 27, 1997.

Parties: Members of the International

Air Transport Association.

Subject: PTC1 0049 dated August 26,

1997 r1–3, PTC1 0050 dated August 26,

1997 r4–7. Expedited TC1 Resolutions

(Summaries attached.) Intended

effective date: October 1, 1997.

Docket Number: OST–97–2861.

Date Filed: August 27, 1997.

Parties: Members of the International

Air Transport Association.

Subject: PTC 0048 dated August 26,

1997 r1–4. Expedited TC1 Longhaul