Federal Register / Vol. 62, No. 63 / Wednesday, April 2, 1997 / Notices 15733

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If the amendment is issued before the

expiration of the 30-day hearing period,

the Commission will make a final

determination on the issue of no

significant hazards consideration. If a

hearing is requested, the final

determination will serve to decide when

the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to John

Hannon: petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to John Flynn,

Esq., Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226, attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

CFR 2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment dated March 27, 1997,

which is available for public inspection

at 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 Monroe County Library System,

3700 South Custer Road, Monroe,

Michigan 48161.

Dated at Rockville, Maryland, this 28th day

of March 1997.

For the Nuclear Regulatory Commission.

Linh N. Tran,

Project Manager, Project Directorate III–1,

Division of Reactor Projects—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–8548 Filed 4–1–97; 8:45 am]

BILLING CODE 7590–01–P

[Docket No. 50–390]

Watts Bar Nuclear Plant, Unit 1; Notice

of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity For a Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. NPF–

90, issued to the Tennessee Valley

Authority (TVA or the licensee), for

operation of the Watts Bar Nuclear Plant

(WBN), Unit 1 located in Rhea County,

Tennessee. This Notice supersedes a

Notice placed in the Federal Register on

March 26, 1997 (62 FR 14469) on this

matter.

The proposed amendment would

revise the Watts Bar Nuclear Plant

(WBN) Unit 1 Technical Specifications

to increase the enrichment and storage

capacity of the spent fuel pool racks.

The proposed modification increases

the WBN spent fuel storage capacity

from 484 fuel assemblies to 1835 fuel

assemblies. The initial enrichment of

the fuel to be stored in the spent fuel

storage racks will be increased from 3.5

weight percent (wt%) to 5.0 wt%. This

modification would also change the

spacing of stored fuel assembly centerto-center spacing from a nominal 10.72

inches to 10.375 inches in 24 PaR flux

trap rack modules and 8.972 inches in

ten smaller burnup credit rack modules

to be installed peripherally along the

south and west pool walls and in a

single 15 x 15 burnup credit rack to be

installed in the cask pit.

In addition to the above proposed

revisions, two limiting conditions for

operation will be added to require that

the combination of initial enrichment

and burnup of each spent fuel assembly

to be stored is in the acceptable region

and to require boron concentration of

the cask pit to be greater than or equal

to 2000 parts per million (ppm) during

fuel movement in the flooded cask pit.

As an added protection to the fuel

stored in the cask pit area, the Technical

Requirements Manual (TRM) is being

revised to require that an impact shield

be in place over the fuel when heavy

loads are moved near or across the cask

pit area.

The WBN Unit 1 Technical

Specification Bases and the TRM would

be revised to support these changes.

Before issuance of the proposed

license amendment, the Commission

will have made findings required by the

Atomic Energy Act of 1954, as amended

(the Act) and the Commission’s

regulations.

The Commission has made a

proposed determination that the

amendment request involves no

significant hazards consideration. Under

the Commission’s regulations in 10 CFR

50.92, this means that operation of the

facility in accordance with the proposed

amendment would not (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of

a new or different kind of accident from

any accident previously evaluated; or

(3) involve a significant reduction in a

margin of safety. As required by 10 CFR

50.91(a), the licensee has provided its

analysis of the issue of no significant

hazards consideration, which is

presented below:

The Nuclear Regulatory Commission has

provided standards for determining whether

a significant hazards consideration exists (10

CFR 50.92(c)). A proposed amendment to an

operating license for a facility involves no

significant hazards consideration if operation

of the facility in accordance with the

proposed amendment would not (1) involve

a significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

Each standard is discussed below for the

proposed amendment.

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The following potential scenarios were

considered:

1. A spent fuel assembly drop.

2. Drop of the transfer canal gate or the

cask pit divider gate.

3. A seismic event.

4. Loss-of-cooling flow in the spent fuel

pool.

5. Installation activities.

The effect of additional spent fuel pool

storage cells fully loaded with fuel on the
15734 Federal Register / Vol. 62, No. 63 / Wednesday, April 2, 1997 / Notices

first four potential accident scenarios listed

above has been considered. It was concluded

that after installation activities have been

completed, the presence of additional fuel in

the pool does not increase the probability of

occurrence of these four events. Also, based

on evaluations of bulk pool temperature, rack

seismic responses, and refueling accidents, it

is reasonable to conclude that there is no

significant increase in the consequences of

these events after installation is complete

(See Reference 1). During the installation

activities, the following considerations

support a conclusion that neither the

probability or consequences of these four

scenarios would be significantly increased.

A spent fuel assembly cannot be dropped

during installation of the 24 Programmed and

Remote System Corporation (PaR) flux trap

rack modules because this activity will take

place before the end of operating cycle one

and there will be no spent fuel in the WBN

pool to be moved or shuffled. Before

installing the ten smaller burnup credit racks

in the pool, some fuel will be moved to create

a three foot lateral free zone clearance from

stored fuel. This would involve a one-time

movement of an estimated maximum of 225

fuel assemblies, which is less tha[n] half the

fuel movements during one refueling outage.

This does not significantly increase the

probability of dropping a fuel assembly,

particularly when the many administrative

controls and physical limitations imposed on

fuel handling operations are considered. The

fuel handling system consists of equipment

and structures utilized for safely

implementing refueling operations in

accordance with requirements of General

Design Criteria 61 and 62 of 10 CFR 50,

Appendix A. The radiological dose

consequences of dropping a 5.0 wt% fuel

assembly are different from the previous

FSAR [Final Safety Analysis Report]

evaluation for the 3.5 wt% fuel assembly.

The Beta and Gamma doses decrease and the

maximum thyroid dose increase is less than

9%. Therefore, the change in calculated dose

values is insignificant and remains well

within regulatory guidelines.

It may be necessary to move the transfer

canal gate and the cask pit divider gate

between their gated and stored positions

during installation of the burnup credit

‘‘baby’’ rack modules along the south and

west walls. During rack installation, the

previously mentioned three foot lateral free

zone clearance to stored fuel would exist.

Therefore, no heavy load would be carried

directly over irradiated fuel during

installation of the racks. There are numerous

design features which comply with NUREG–

0612 to preclude these gates from dropping

on spent fuel. These features include design

of the lifting devices, design of the crane, and

use of written procedures. Also, the

evaluation results for a gate drop on the racks

indicates that permanent damage to a fuel

storage cell is limited to a maximum depth

of less than six inches below the top of the

rack with no effect on the subcriticality of

fuel stored in adjacent cells. Based on the

foregoing, it is reasonable to conclude that

gate handling during the installation of the

‘‘baby’’ racks would not involve a significant

increase in the probability or consequences

of an accident.

The probability of a seismic event is not

related to installation activities. The worst

consequence resulting from a seismic event

during installation activities would occur

during handling of a rack. The consequences

would be insignificant because the Auxiliary

Building crane is seismically qualified and

both handling equipment and operations

meet the criteria of NUREG–0612.

Nevertheless, if the seismic event resulted in

a rack drop, the consequences are

insignificant, i.e., localized damage to the

pool liner and a minor leak rate which would

be small in comparison to available installed

makeup capacity. The cooling and shielding

of the spent fuel would remain unaffected.

Also the racks being moved are empty during

installation and therefore, the criticality

consequences of seismic events are bounded

by evaluations for loaded racks.

Rack installation activities cannot

cause an accidental loss-of-cooling flow

in the spent fuel pool. The vital

components of the spent fuel pool

cooling and cleanup system (SFPCCS)

are not located proximate to the pool

installation activities. Coolant flow may

be deliberately curtailed to facilitate

installation of the ‘‘baby’’ racks directly

beneath the discharge piping in the

southwest corner of the pool. The effects

of such an action would be readily

minimized and made inconsequential

during the detailed installation planning

phase by selecting a time when decay

heat input from stored fuel is relatively

constant. Also careful preplanning of

the work would minimize out-of-service

time and provide for intermittent

coolant flow restart, if necessary, to

maintain acceptable bulk coolant

temperatures. Similarly, the effect of an

independently initiated loss-of-coolant

flow incident on reracking activities can

be easily accommodated by stopping

work, as necessary, to mitigate any

adverse effects on the installation

process. The consequences of loss-ofcooling flow in the spent fuel pool

during installation are bounded by the

analysis in Chapter 5 of the report

which includes the situation in which

‘‘baby’’ racks and the 15 x 15 cask pit

rack are installed, and the pool is filled

to capacity with spent fuel.

With regard to the actual installation

activities, the existing WBN TRM prohibits

loads in excess of 2059 pounds from travel

over fuel assemblies in the storage pool and

requires the associated crane interlocks and

physical stops be periodically demonstrated

operable. During installation, racks and

associated handling tools will be moved over

the spent fuel pool, however there will be no

fuel in the pool when the 24 flux trap rack

modules are installed. A three foot lateral

free zone clearance from stored spent fuel

will be maintained during installation of the

ten smaller burnup credit rack modules.

Installation work in the spent fuel pit area

will be controlled and performed in strict

accordance with specific written

instructions.

NUREG–0612 states that in lieu of

providing a single failure-proof crane system,

the control-of-heavy-loads guidelines can be

satisfied by establishing that the potential for

a heavy load drop is extremely small. Storage

rack movements to be accomplished with the

WBN Auxiliary Building crane will conform

with NUREG–0612 guidelines in that the

probability of a drop of a storage rack is

extremely small. The crane has a tested

capacity of 125 tons. The maximum weight

of any existing, replacement, or new storage

rack and its associated handling tool is less

than 20 tons. Therefore, there is ample safety

factor margin for movements of the storage

racks by the Auxiliary Building crane.

Special lifting devices, which have

redundancy or a rated capacity sufficient to

maintain adequate safety factors, will also be

utilized in the movements of the storage

racks. In accordance with NUREG–0612,

Appendix B, the safety margin ensures that

the probability of a load drop is extremely

low.

Future load travel over fuel stored in a rack

specifically designed for the cask loading

area of the cask pit will be prohibited unless

an impact shield, which has been specifically

designed for this purpose, is covering the

area. Loads that are permitted when the

shield is in place must meet analytically

determined weight, travel height, and crosssectional area criteria that preclude

penetration of the shield. A Technical

Requirement (TR) has been proposed that

incorporates the previously mentioned load

criteria.

Also a rack change-out sequence is being

developed that addresses removal of the

existing racks, movement of the new racks

into the Auxiliary Building, initial staging on

the refueling floor, and final installation in

the pool. The change-out sequence objectives

include establishing lift heights, travel

distances, and number of lifts to be as low

as reasonably achievable. Accordingly, it is

concluded that the proposed installation

activities will not significantly increase the

probability of a load-handling accident. The

consequences of a load-handling accident are

unaffected by the proposed installation

activities.

The consequences of a spent fuel assembly

drop were evaluated, and it was determined

that the racks will not be distorted such that

the racks would not perform their safety

function. The criticality acceptance criterion,

Keff less than or equal to 0.95, is not violated,

and the calculated doses are well within 10

CFR Part 100 guidelines. The radiological

consequences of the fuel assembly drop

accident evaluated for WBN, have changed,

however, the changes do not involve a

significant increase in consequences and are

well within the 10 CFR 100 requirements.

A TRM change has been proposed that

would permit the transfer-canal gate and the

divider gate for the cask pit to travel over fuel

assemblies in the spent fuel pool during

movement between their gated and stored

position. Rack damage is restricted to an area

above the active fuel region, therefore,

neither criticality nor radiological concerns

exist.

The consequences of a seismic event have

been evaluated. The replacement racks are
Federal Register / Vol. 62, No. 63 / Wednesday, April 2, 1997 / Notices 15735

designed and fabricated and the new racks

will be fabricated to meet the requirements

of applicable portions of the NRC regulatory

guides and published standards. Design

margins have been provided for rack tilting,

deflection, and movement such that the racks

do not impact each other or the spent fuel

pool walls in the active fuel region during the

postulated seismic events. The free-standing

racks will maintain their integrity during and

after a seismic event. The fuel assemblies

also remain intact and therefore no criticality

concerns exist.

The spent fuel pool system is a passive

system with the exception of the fuel pool

cooling train and heating, ventilating, and

air-conditioning (HVAC) equipment.

Redundancies in the cooling train and HVAC

hardware are not reduced by the planned fuel

storage modification. The potential increased

heat load resulting from any additional

storage of spent fuel is well within the

existing system cooling capacity. Therefore,

the probability of occurrence or malfunction

of safety equipment leading to the loss-ofcooling flow in the spent fuel pool is not

significantly affected. Furthermore, the

consequences of this type incident are not

significantly increased from previously

evaluated cooling system loss of flow

malfunctions. Thermal-hydraulic scenarios

assume the reracked pool is approximately

90% full with spent fuel assemblies. From

this starting point, the remaining storage

capacity is utilized by analyzing both normal

and unplanned full core off loads using

conservative assumptions and previously

established methods. Calculated values

include maximum pool water bulk

temperature, coincident maximum pool

water local temperature, the maximum fuel

cladding temperature, time-to-boil after lossof-cooling paths, and the effect of flow

blockage in a storage cell.

Although the proposed modification

increases the pool heat load, results from the

above analyses yield a maximum bulk

temperature less than 160 degrees Fahrenheit

which is below the bulk boiling temperature.

Also the maximum local water temperature

is below nucleate boiling condition values.

Associated results from corresponding lossof-cooling evaluations give minimums of 5.3

hours before boiling begins and 45 hours

before the pool water level drops to the

minimum required for shielding spent fuel.

This is sufficient time to begin utilization

of available alternate sources of makeup

cooling water. Also, the effect of the

increased thermal loading on the pool

structure, associated cooling system, and

components was evaluated and determined

to establish an acceptable design basis with

the new storage configuration. No

modifications were necessary because of the

increased temperature.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

The proposed modification has been

evaluated in accordance with the guidance of

the NRC position paper entitled, ‘‘OT

Position for Review and Acceptance of

Spent-Fuel Storage and Handling

Applications’’, appropriate NRC regulatory

guidelines; appropriate NRC standard review

plans; and appropriate industry codes and

standards. Proven analytical technology was

used in designing the planned fuel storage

expansion and will be utilized in the

installation process. Basic reracking

technology has been developed and

demonstrated in applications for fuel pool

capacity increases that have already received

NRC staff approval.

Proposed TSs for the spent fuel storage

racks use burnup credit and fuel assembly

administrative placement restrictions for

criticality control. These restrictions are

described in the proposed change to the

design features section of the TSs by

reference to the Spent Fuel Pool

Modifications report. Additional evaluations

were required to ensure that the criticality

criterion, keff less than or equal to 0.95, is

maintained. These include evaluation for the

abnormal placement of unirradiated (fresh)

fuel assemblies of 5.0 wt% enrichment into

a storage cell location designed for lower

enrichment or irradiated fuel. Soluble boron,

for which credit is permitted under these

abnormal conditions, ensures that reactivity

is maintained substantially less than the

design requirement. For example, if the PaR

flux trap racks are inadvertently all loaded

with fresh assemblies of the maximum 5.0

wt% fuel instead of observing the 3.8 wt%

and 6.75 MWD/KgU controls, the worth of

the 2000 ppm borated water is sufficient to

lower the keff of the storage racks to 0.83. The

existing and proposed TSs require boron

concentration in the pool and cask pit to be

greater than or equal to 2000 ppm during fuel

movement. An analytical determination of

the reactivity worth of 2000 ppm borated

water in the spent fuel storage pool predicted

the change in keff to be approximately 17

percent keff. Although no credit for soluble

boron was proposed in the TSs, it was also

determined by an independent calculation

that a minimum concentration of 520 ppm

soluble boron allows the unrestricted storage

of 5.0 wt% enriched fuel in the PaR flux trap

racks.

The Holtec-designed peripheral ‘‘baby’’

racks and the 15 x 15 racks in the cask

loading area can safely and conservatively

store fuel of 5 wt% initial enrichment burned

to 41 MWD/kgU or lower enriched fuel with

lower burnup, i.e., fuel of equivalent

reactivity. Evaluations have confirmed that,

for the abnormal placement of a fresh fuel

assembly of 5.0 wt% in these racks, the

criticality criterion is maintained with the

existing and proposed TS requirements of

2000 ppm soluble boron.

Although these changes required

addressing additional aspects of a previously

analyzed accident, the possibility of a

previously unanalyzed accident is not

created.

The impact shield design together with its

attendant administrative controls and

NUREG–0612 heavy load lift compliance,

renders the possibility of a heavy load drop

on fuel as not credible in accordance with the

NUREG–0612 single-failure-proof criteria.

Accordingly, since this particular part of the

proposed reracking modification is not a

change that could malfunction by a new

single failure, the movement of heavy loads

over the cask pit does not create the

possibility of a new or different kind of

accident.

It is therefore concluded that the proposed

reracking does not create the possibility of a

new or different kind of accident from any

previously analyzed.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The design and technical review process

applied to the reracking modification

included addressing the following areas:

1. Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material, and structural

considerations.

The established acceptance criterion for

criticality is that the neutron multiplication

factor shall be less than or equal to 0.95,

including all uncertainties. The results of the

criticality analyses for the rack designs

demonstrate that this criterion is satisfied.

The methods used in the criticality analysis

conform to the applicable portions of NRC

guidance and industry codes, standards, and

specifications. In meeting the acceptance

criteria for criticality in the spent fuel pool

and the cask loading area, such that keff is

always less than 0.95 at a 95/95 percent

probability tolerance level, the proposed

amendment does not involve a significant

reduction in the margin of safety for nuclear

criticality.

Conservative methods and assumptions

were used to calculate the maximum fuel

temperature and the increase in temperature

of the water in the spent fuel pit area. The

thermal-hydraulic evaluation used methods

previously employed. The proposed storage

modification will increase the heat load in

the spent fuel pool, but the evaluation shows

that the existing spent fuel cooling system

will maintain the bulk pool water

temperature at or below 160 degrees

Fahrenheit. Thus it is demonstrated that the

worst-case peak value of the pool bulk

temperature is considerably lower than the

bulk boiling temperature. Evaluation also

shows that maximum local water

temperatures along the hottest fuel assembly

are below the nucleate boiling condition

value. Thus, there is no significant reduction

in the margin of safety for thermal hydraulic

or spent fuel cooling considerations.

The mechanical, material, and structural

design of the spent fuel racks is in

accordance with applicable portions of NRC’s

position in ‘‘OT Position for Review and

Acceptance of Spent-Fuel Storage and

Handling Applications,’’ dated April 14,

1978 (as modified January 18, 1979), as well

as other applicable NRC guidance and

industry codes. The primary safety function

of the spent fuel racks is to maintain the fuel

assemblies in a safe configuration through

normal and abnormal loading conditions.

Abnormal loadings that have been evaluated

with acceptable results and discussed

previously include the effect of an

earthquake and the impact because of the

drop of a fuel assembly. The rack materials

used are compatible with the fuel assemblies

and the environment in the spent fuel pool.
15736 Federal Register / Vol. 62, No. 63 / Wednesday, April 2, 1997 / Notices

The structural design for the new racks

provides tilting, deflection, and movement

margins such that the racks do not impact

each other or the spent fuel pit walls in the

active fuel region during the postulated

seismic events. Also the spent fuel

assemblies themselves remain intact and no

criticality concerns exist. In addition, finite

element analysis methods were used to

evaluate the continued structural

acceptability of the spent fuel pit. The

analysis was performed in accordance with

‘‘Building Code Requirements for Reinforced

Concrete,’’ (ACI 318–63,77). Therefore, with

respect to mechanical, material, and

structural considerations, there is no

significant reduction in a margin of safety.

Summary

Based on the above analysis, TVA has

determined that operation of WBN, in

accordance with the proposed amendment,

would not: (1) involve a significant increase

in the probability of consequences of an

accident previously evaluated, (2) create the

possibility of a new or different kind of

accident from any accident previously

evaluated, or (3) involve a significant

reduction in a margin of safety. Therefore,

operations of WBN in accordance with the

proposed amendments as described do not

involve significant hazard considerations as

defined in 10 CFR 50.92 and that the criteria

of 10 CFR 50.91 have accordingly been met.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within thirty (30) days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period, such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received. Should

the Commission take this action, it will

publish in the Federal Register a notice

of issuance and provide for opportunity

for a hearing after issuance. The

Commission expects that the need to

take this action will occur very

infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville Maryland, from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC.

The filing of requests for hearing and

petitions for leave to intervene is

discussed below.

By May 2, 1997, the licensee may file

a request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at 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

Chattanooga-Hamilton County Library,

1001 Broad Street, Chattanooga,

Tennessee 37402. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing.

The petitioner must also provide

references to those specific sources and

documents of which the petitioner is

aware and on which the petitioner

intends to rely to establish those facts or

expert opinion. Petitioner must provide

sufficient information to show that a

genuine dispute exists with the

applicant on a material issue of law or

fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the final

determination will serve to decide when

the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,
Federal Register / Vol. 62, No. 63 / Wednesday, April 2, 1997 / Notices 15737

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to Mr.

Frederick J. Hebdon: petitioner’s name

and telephone number, date petition

was mailed, plant name, and

publication date and page number of

this Federal Register notice. A copy of

the petition should also be sent to the

Office of the General Counsel, U.S.

Nuclear Regulatory Commission,

Washington, DC, and to General

Counsel, Tennessee Valley Authority,

ET 10H, 400 West Summit Hill Drive,

Knoxville, Tennessee 37902, attorney

for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

CFR 2.714(a)(1)(i)–(v) and 2.714(d).

The Commission hereby provides

notice that this is a proceeding on an

application for a license amendment

falling within the scope of section 134

of the Nuclear Waste Policy Act of 1982

(NWPA), 42 U.S.C. 10154. Under

section 134 of the NWPA, the

Commission, at the request of any party

to the proceeding, must use hybrid

hearing procedures with respect to ‘‘any

matter which the Commission

determines to be in controversy among

the parties.’’ The hybrid procedures in

section 134 provide for oral argument

on matters in controversy, preceded by

discovery under the Commission’s

rules, and the designation, following

argument, of only those factual issues

that involve a genuine and substantial

dispute, together with any remaining

questions of law, to be resolved in an

adjudicatory hearing. Actual

adjudicatory hearings are to be held on

only those issues found to meet the

criteria of section 134 and set for

hearing after oral argument.

The Commission’s rules

implementing section 134 of the NWPA

are found in 10 CFR Part 2, Subpart K,

‘‘Hybrid Hearing Procedures for

Expansion of Spent Nuclear Fuel

Storage Capacity at Civilian Nuclear

Power Reactors’’ (published at 50 FR

41670, October 15, 1985) to 10 CFR

2.1101 et seq. Under those rules, any

party to the proceeding may invoke the

hybrid hearing procedures by filing with

the presiding officer a written request

for oral argument under 10 CFR 2.1109.

To be timely, the request must be filed

within 10 days of an order granting a

request for hearing or petition to

intervene. (As outlined above, the

Commission’s rules in 10 CFR Part 2,

Subpart G, and 2.714 in particular,

continue to govern the filing of requests

for a hearing or petitions to intervene,

as well as the admission of contentions.)

The presiding officer shall grant a

timely request for oral argument. The

presiding officer may grant an untimely

request for oral argument only upon

showing of good cause by the requesting

party for the failure to file on time and

after providing the other parties an

opportunity to respond to the untimely

request. If the presiding officer grants a

request for oral argument, any hearing

held on the application shall be

conducted in accordance with the

hybrid hearing procedures. In essence,

those procedures limit the time

available for discovery and require that

an oral argument be held to determine

whether any contentions must be

resolved in adjudicatory hearing. If no

party to the proceedings requests oral

argument, or if all untimely requests for

oral argument are denied, then the usual

procedures in 10 CFR Part 2, Subpart G,

apply.

For further details with respect to this

action, see the application for

amendment dated, October 23, 1996, as

supplemented on December 11, 1996,

January 31, February 10 and 24 and

March 11, 1997 which is available for

public inspection at 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

Chattanooga-Hamilton County Library,

1001 Broad Street, Chattanooga,

Tennessee.

Dated at Rockville, Maryland, this 27th day

of March 1997.

For the Nuclear Regulatory Commission.

Robert E. Martin,

Sr. Project Manager, Project Directorate II–

3, Division of Reactor Projects—I/II, Office

of Nuclear Reactor Regulation.

[FR Doc. 97–8401 Filed 4–1–97; 8:45 am]

BILLING CODE 7590–01–P

[Docket No. 72–9]

Department of Energy Issuance of

Environmental Assessment and

Finding of No Significant Impact

Regarding The Transfer of The

Materials License SNM–2504 and

Subsequent License Amendment For

The Fort St. Vrain Independent Spent

Fuel Storage Installation From The

Public Service Company of Colorado

to The U.S. Department of Energy

The Nuclear Regulatory Commission

is considering the issuance of an order

approving an application from the U.S.

Department of Energy, Idaho Operations

Office (the applicant or DOE–ID) dated

December 17, 1996, as supplemented

February 4, 5, and 18, and March 12 and

13, 1997, for the transfer of Materials

License SNM–2504 and subsequent

license amendment, under the

provisions of 10 CFR Part 72. In its

application, DOE–ID included the

‘‘Environmental Report for the Fort St.

Vrain Independent Spent Fuel Storage

Installation’’ (December 1996).

Environmental Assessment

Identification of the Proposed Action

The applicant is seeking NRC

approval to take possession of spent

nuclear fuel and other radioactive

materials associated with spent nuclear

fuel storage presently in the possession

of the Public Service Company of

Colorado (PSCo) at its Fort St. Vrain

(FSV) independent spent fuel storage

installation (ISFSI) located in Weld

County, Colorado, and to own and

operate the FSV ISFSI. The transfer of

an ISFSI license is subject to NRC

approval under 10 CFR 72.50, ‘‘Transfer

of License.’’ Pursuant to the provisions

of 10 CFR Part 72, the term of the

license for the ISFSI would remain as is

currently licensed, and the license

would expire on November 30, 2011. If

the application for transfer is approved,

the Commission will issue an order

consenting to the transfer. The NRC is

also considering an amendment to the

materials license to reflect DOE–ID as

the new licensee for the FSV ISFSI and

the addition of revised Appendices A,

B, and C to the license.