59912 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

Manager for Enrichment Facilities, Oak Ridge

Operations Office, DOE, and the Director,

Division of Fuel Cycle Safety and Safeguards,

Office of Nuclear Material Safety and

Safeguards, NRC.

VI. Effective Date and Modification

This MOU shall become effective upon

signing by the DOE Assistant Manager for

Enrichment Facilities, Oak Ridge Operations,

and the Director, Division of Fuel Cycle

Safety and Safeguards, Office of Nuclear

Material Safety and Safeguards, NRC, and

will be subject to periodic reviews and may

be amended or modified upon written

agreement by the parties. This MOU may be

terminated by mutual agreement or by

written notice of either party submitted six

months in advance of termination.

VII. Separability

If any provision(s) of this MOU, or the

application of any provision(s) to any person

or circumstances, is held invalid, the

remainder of this MOU and the application

of such provision(s) to other persons or

circumstances shall not be affected.

For the Nuclear Regulatory Commission.

Dated: October 27, 1997.

Elizabeth Q. Ten Eyck,

Director Division of Fuel Cycle Safety and

Safeguards, Office of Nuclear Material Safety

and Safeguards, Nuclear Regulatory

Commission.

For the Department of Energy.

Dated: October 28, 1997.

Joseph W. Parks,

Assistant Manager for Enrichment Facilities,

Oak Ridge Operations Office, Department of

Energy.

[FR Doc. 97–29244 Filed 11–4–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from October 10,

1997, through October 24, 1997. The last

biweekly notice was published on

October 22, 1997 (62 FR 54866).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

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. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 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 before

action is taken. 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 Chief, Rules 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, MD 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 a hearing and petitions

for leave to intervene is discussed

below.

By December 5, 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 for the particular

facility involved. 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 a 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
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59913

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

Commission will make a final

determination on the issue of no

significant hazards consideration. 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:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington DC, by the above date. 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 the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

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 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 for the particular

facility involved.

Baltimore Gas and Electric Company,

Docket No. 50–317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert

County, MD

Date of amendment request: October

2, 1997.

Description of amendment request:

The amendment request would change

the Technical Specifications to identify

a proposed upgrade of the electrical

capacity of the No. 1B emergency diesel

generator.

Basis for proposed no significant

hazards consideration determination:

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:

1. Would not involve a significant

increase in the probability or

consequences of an accident previously

evaluated.

The Engineered Safety Features (ESF)

electrical system provides a reliable

source of electrical power to the 4.16 kV

ESF busses to operate the necessary

accident mitigation equipment, should

offsite power be lost. The proposed

change to the Technical Specifications

was prompted by the upgrade of the

electrical and mechanical capacity of

the No. 1B Fairbanks Morse Emergency

Diesel Generator (EDG). The increased

electrical capacity of the No. 1B

Fairbanks Morse EDG will give the

operators greater flexibility in the choice

of discretionary loads for the mitigation

of accidents. This modification

necessitates changes to the Technical

Specifications.

The ESF electrical system, including

the four EDGs, is used to mitigate the

consequences of an accident. The

modification to upgrade the capacity of

No. 1B EDG will increase the electrical

output of the EDG, but will not change

the configuration of the ESF electrical

system or any support systems such that

the EDGs would become an accident

initiator. Therefore, the proposed

change would not increase the

probability of an accident previously

evaluated.

The proposed Technical

Specifications will continue to

demonstrate the reliability and

capability of the upgraded No. 1B EDG

to perform its accident mitigation

function. The proposed changes to the

surveillance requirements do not alter

the intent or performance of the

surveillance. Only the electrical

loadings changed, reflecting the change

in the EDG’s electrical capacity.

Implementation of the proposed

Technical Specifications will not reduce

the ability of No. 1B EDG to perform its

safety functions. Any auxiliary systems

that required modification or analysis to

support the upgraded ratings of the 1B

Fairbanks Morse EDG have been

determined not to adversely impact

operation of any other plant systems

necessary to mitigate the consequences

of an accident. Therefore, the proposed

change would not increase the

consequences of an accident previously

evaluated.

Therefore, the proposed change does

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of

a new or different type of accident from

any accident previously evaluated.

The proposed change increases the

electrical loading for surveillance

requirements to reflect the upgrade to

the electrical capacity of the No. 1B

Fairbanks Morse EDG. This change does

not add any new equipment, modify any

interfaces with any existing equipment,

change the equipment’s function, or the

method of operating the equipment to

be modified. The system will continue

to operate in the same manner as before

the capacity upgrades were

implemented. The modified No. 1B EDG

will continue to function as an accident
59914 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

mitigator, and will not become an

initiator of any accident.

Therefore, the proposed change does

not create the possibility of a new or

different type of accident from any

accident previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The safety function of the EDG is to

provide a reliable source of electrical

power to the ESF electrical system

sufficient to power the necessary

accident mitigation equipment, should

offsite power be lost. This safety

function is demonstrated by performing

the required surveillance tests. The

proposed changes do not alter the intent

or method of performance of any of the

surveillance tests.

The proposed change to the Technical

Specifications was prompted by the

upgrade of the electrical and mechanical

capacity of the No. 1B Fairbanks Morse

EDG. The higher electrical capacity will

result in an increase in the margin

between No. 1B EDG’s electrical

capacities and the electrical power

required to operate safety-related

equipment required for safe shutdown

or accident mitigation. The increased

electrical capacity results in the need to

increase the electrical loadings used in

the surveillance tests. The changes in

the surveillance tests will continue to

ensure that the EDG is tested

appropriately and will continue to

perform its safety function. In addition,

it should be noted that upgrades on

identical Fairbanks Morse EDGs have

already been performed on Unit 2 and

have resulted in identical changes to the

Unit 2 Technical Specifications.

Because of the increased electrical

margin afforded by the upgraded EDG,

these modifications may be considered

an increase in the margin of safety.

Therefore, the proposed change does

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, MD 20678.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: S. Singh Bajwa,

Director.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, IL; Docket Nos. STN 50–

456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

IL

Date of amendment request:

September 8, 1997.

Description of amendment request:

The proposed amendment would revise

Byron and Braidwood Technical

Specification (TS) 4.5.2.b and associated

bases as they relate to the requirement

to vent the Emergency Core Cooling

System (ECCS) pump casings and

discharge piping high points outside

containment. The change will revise the

Unit 1 requirement for ultrasonic

examinations every 31 days to also

include ultrasonic examination of the

piping at the 1CV206 valve for Byron

(1CV207 valve for Braidwood) if the 1B

Chemical and Volume Control (CV)

pump is idle. These changes are

required to align the surveillance

requirements for Unit 1 with those of

Unit 2. In addition, the condition that

the Unit 1 requirements will be

applicable only until the end of the

current cycle is deleted consistent with

the Unit 2 requirements. With these

changes there will no longer be the need

to maintain separate pages for Unit 1

and Unit 2 requirements.

Basis for proposed no significant

hazards consideration determination:

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:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes will align the

surveillance requirements for both Units

1 and 2 with the installed system design

and normal operating conditions. No

increase in the probability of an

accident will occur as a result of this

change. The conduct of surveillances

required by the Technical Specifications

is not postulated to initiate an accident.

The level of surveillance performed to

date has provided confidence that the

objective of the current surveillance

requirement has been met. As such, the

proposed change does not result in a

significant increase in the probability of

occurrence of a previously analyzed

accident.

The consequences of a previously

analyzed accident are not increased.

Operating experience has shown that

the level of surveillance performed to

date is sufficient to provide confidence

that no significant voiding has occurred

in the affected piping. Ultrasonic

examinations have confirmed the water

solid condition of the piping. Although

voiding is not expected, evaluation of

postulated voided conditions confirm

that unacceptable dynamic loading

would not occur, and, therefore, the

integrity of the ECCS piping is not

compromised. Thus, the ECCS will be

capable of performing its design

function of cooling the reactor core and

providing shutdown capability

following initiation of the certain

accidents. This will ensure that the

consequences of a previously analyzed

accident are not significantly increased.

Therefore, these proposed revisions

do not result in a significant increase in

the probability or consequences of an

accident previously analyzed.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes do not create

the possibility of a new or different kind

of accident. ComEd has evaluated the

piping configuration for the ECCS

discharge piping of the ECCS

subsystems. A specific engineering

evaluation of both a voided 2-inch and

8-inch RH [Residual Heat Removal] line

was performed. This evaluation

concluded that the piping can withstand

the dynamic loads caused by the

maximum credible air void. Due to the

higher-pressure rating and smaller size

of the SI [Safety Injection] and CV

discharge piping, this evaluation is

considered bounding for the ECCS

subsystems. The results of the

evaluation were submitted for staff

review in a letter dated March 12, 1990,

in support of Amendments 47 and 36 to

the Operating Licenses for Byron and

Braidwood, respectively. The proposed

changes will not result in new failure

modes because no new equipment is

installed, and installed equipment is not

operated in a new or different manner.

Manual venting operations have been

performed as permitted by system

operation and piping configuration.

This venting surveillance does not

apply to subsystems in communication

with operating systems because the

flows and/or pressures prevalent in

these systems are sufficient to provide

confidence that water hammer which

could occur from voiding would not

result in unacceptable dynamic loads

from water hammer will not occur.

Accordingly, this change will not create

the possibility of a new or different kind

of accident.

3. The proposed change does not

involve a significant reduction in a

margin of safety.
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59915

The margin of safety is not

significantly reduced because the

proposed change will provide sufficient

assurance that excessive voiding will

not occur. This will assure proper

system functioning. Venting of the idle

subsystems, in conjunction with the

operating conditions of the subsystems

in operation, provides confidence that

voiding is not present. This has been

confirmed by the performance of

ultrasonic examinations of the piping of

interest. This meets the objective of the

surveillance requirement and thus

preserves the margin of safety.

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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, IL 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, IL 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, IL 60603.

NRC Project Director: Robert A. Capra.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, MI

Date of amendment request: January

18, 1996, as revised October 1, 1997.

Description of amendment request:

The original proposed amendment

(January 18, 1996) would have deleted

the requirement in Section 6.5.6 of the

Technical Specifications (TS) to perform

inservice inspections of the primary

coolant pump (PCP) flywheels. The

October 1, 1997, submittal would revise

Section 6.5.6 of the TS to lengthen the

flywheel inspection period to 10 years

rather than delete it entirely. The note

added by Amendment 175 for the

deletion of the inspection at the end of

Cycle 12 would also be deleted. The

original submittal was previously

noticed in the Federal Register on

September 11, 1996 (61 FR 47976).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee provided its analysis of the

issue of no significant hazards

consideration in its original submittal.

In its revised submittal the licensee

stated that the conclusions reached in

the original no significant hazards

consideration determination were still

valid because the revised submittal just

reduces the frequency of the test as

opposed to deleting it. The original no

significant hazards consideration

discussion is presented below:

The following evaluation supports the

finding that operation of the facility in

accordance with the proposed change to

the Technical Specifications would not:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed change to the Technical

Specifications would delete the

requirement to perform non-destructive

examination of the upper flywheel on

the PCPs. The fracture mechanics

analyses conducted to support the

change show that a preexisting crack

sized just below detection level will not

grow to the flaw size necessary to result

in flywheel failure within the life of the

plant. This analysis conservatively

assumes minimum material properties,

maximum flywheel accident speed,

location of the flaw in the highest stress

area and a number of startup/shutdown

cycles eight times greater than expected.

Since an existing flaw in the flywheel

will not grow to the allowable flaw size

under normal operating conditions or to

the critical flaw size under LOCA [lossof-coolant accident] conditions over the

life of the plant, elimination of inservice

inspection for such cracks during the

plant’s life will not involve a significant

increase in the probability of an

accident previously considered.

The proposed changes do not increase

the amount of radioactive material

available for release or modify any

systems used for mitigation of such

releases during accident conditions.

Therefore, operation of the facility in

accordance with the proposed change to

the Technical Specifications would not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed change to the Technical

Specifications would not change the

design, configuration, or method of

operation of the plant and therefore,

operation of the facility in accordance

with the proposed change to the

Technical Specifications would not

create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change to the Technical

Specifications would not result in a

significant reduction in the margin of

safety. Significant conservatisms have

been used for calculating the allowable

flaw size, critical flaw size and crack

growth rate in the PCP flywheels. These

include minimum material properties,

maximum flywheel accident speed,

location of the postulated flaw in

highest stress area and a number of

startup/shutdown cycles eight times

greater than expected. Since an existing

flaw in the flywheel will not grow to the

maximum allowable flaw size under

normal operating conditions or to the

critical flaw size under LOCA

conditions over the life of the plant,

elimination of inservice inspections for

such cracks during the plant’s life will

not involve a significant reduction in

the margin of safety.

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. In addition, the staff agrees

that this analysis bounds the conditions

in the revised submittal. The editorial

change to delete an obsolete note has no

effect on plant operation or safety and

also satisfies the three standards of 10

CFR 50.92(c). Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, MI 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Energy Company,

212 West Michigan Avenue, Jackson, MI

49201.

NRC Project Director: John N.

Hannon.

GPU Nuclear Corporation, et al., Docket

No. 50–219, Oyster Creek Nuclear

Generating Station, Ocean County, NJ

Date of amendment request: October

10, 1997.

Description of amendment request:

The proposed change (TSCR 253) would

reflect the registered trade name of

‘‘GPU Nuclear’’ in the operating license

for the Oyster Creek Nuclear Generating

Station (OCNGS) and change the legal

name of the operator of OCNGS from

GPU Nuclear Corporation to GPU

Nuclear, Inc. In addition, two minor

editorial corrections are included.

Basis for proposed no significant

hazards consideration determination:

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:

Operation of the facility in accordance

with the proposed amendment would

not involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The proposed amendment

adds to the license and the technical

specifications the trade name of the
59916 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

Owner of Oyster Creek. The change in

the legal name of the operator of Oyster

Creek is an administrative change made

to reflect the name changes made

throughout the GPU family of

companies. The name change has no

impact on plant design or operation.

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 evaluated because

no new failure modes are created by the

proposed changes. The use of a trade

name for the Owner of Oyster Creek and

the change in the legal name of the

operator of Oyster Creek has no impact

on plant design or operation. Thus,

there is no creation of the possibility of

a new or different kind of accident from

those previously evaluated.

Operation of the facility in accordance

with the proposed amendment will not

involve a significant reduction in a

margin of safety. The proposed

amendment does not change any

operating limits for reactor operation.

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. In addition, the staff has

reviewed the licensee’s proposed

editorial changes and determined that

they do not effect the conclusions of the

analysis. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Ronald B.

Eaton, Acting Director.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 1, Oswego County,

NY

Date of amendment request: October

21, 1997. This notice supersedes a

previous notice, (62 FR 30625),

published June 4, 1997, which was

based upon the licensee’s application

for amendment dated May 16, 1997. The

licensee’s application dated October 21,

1997, supersedes the May 16, 1997,

submittal in its entirety.

Description of amendment request:

The proposed amendment would

change the administrative section of the

Technical Specifications (TS) regarding

the Operations organization.

Specifically, TS 6.2.2i currently states

that ‘‘The Manager Operations, Station

Shift Supervisor Nuclear and Assistant

Station Shift Supervisor Nuclear shall

hold senior reactor operator licenses.’’

This would be changed to state ‘‘As a

minimum, either the Manager

Operations or the General Supervisor

Operations shall hold a senior reactor

operator license. The Station Shift

Supervisor Nuclear and Assistant

Station Shift Supervisor Nuclear shall

hold senior reactor operator licenses.’’

In addition TS 6.3.1 would be revised to

indicate an additional exception to the

operating staff’s qualification

requirements set forth in American

National Standard Institute (ANSI)

N18.1–1971, ‘‘Selection and Training of

Nuclear Power Plant Personnel.’’

Specifically, this change would require

that the Manager Operation, in lieu of

meeting the senior reactor operator

(SRO) requirements of ANSI N18.1–

1971, shall (1) hold an SRO license at

the time of appointment, or (2) have

held an SRO license at Nine Mile Point

Nuclear Station Unit 1 or a similar unit,

or (3) have been certified for equivalent

SRO knowledge.

Basis for proposed no significant

hazards consideration determination:

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:

1. The operation of Nine Mile Point

Unit 1 [NMP1], in accordance with the

proposed amendment, will not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

The addition of the position of GSO

and the requirement for either the GSO

or the Manager Operations to have an

SRO license is a restructuring of the

Operations department. The proposed

changes are administrative changes that

provide additional Operations

management oversight capabilities.

Additional restrictions placed on the

Manager Operations minimum

qualification requirements for

experience and SRO level knowledge for

the resulting organization meet the

intent of ANSI N18.1–1971 and SRP

[Standard Review Plan, NUREG–0800]

13.1.1–13.1.3. No physical modification

of the plant is involved and no changes

to the methods in which plant systems

are operated are required.

None of the precursors of previously

evaluated accidents are affected, and no

new failure modes are introduced.

Therefore, this change will not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

2. The operation of Nine Mile Point

Unit 1, in accordance with the proposed

amendment, will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The addition of the position of GSO

and the requirement for either the GSO

or the Manager Operations to have an

SRO license is a restructuring of the

Operations department. The proposed

changes are administrative changes that

provide additional Operations

management oversight capabilities.

Additional restrictions placed on the

Manager Operations minimum

qualification requirements for

experience and SRO level knowledge

ensure the resulting organization meets

the intent of ANSI N18.1–1971 and SRP

13.1.1–13.1.3. No physical modification

of the plant is involved and no changes

to the methods in which plant systems

are operated are required. As such, the

change does not introduce any new

failure modes or conditions that may

create a new or different accident.

Therefore, this change does not itself

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. The operation of Nine Mile Point

Unit 1, in accordance with the proposed

amendment, will not involve a

significant reduction in a margin of

safety.

The addition of the position of GSO

and the requirement for either the GSO

or the Manager Operations to have an

SRO license is a restructuring of the

Operations department. The proposed

changes are administrative changes that

provide additional Operations

management oversight capabilities.

Additional restrictions placed on the

Manager Operations minimum

qualification requirements for

experience and SRO level knowledge

ensure the resulting organization meets

the intent of ANSI N18.1–1971 and SRP

13.1.1–13.1.3. No physical modification

of the plant is involved and no changes

to the methods in which plant systems

are operated are required. As such, this

change does not in itself adversely affect

any physical barrier to the release of

radiation to plant personnel or to the

public. Therefore, the change does not

involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59917

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, NY

13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, CT

Date of amendment request: October

7, 1997.

Description of amendment request:

Technical Specifications 4.6.1.1, 3/

4.6.1.2, and 3/4.6.1.3 require the testing

of the containment to verify leakage

limits at a specified test pressure. The

proposed amendment would (1) modify

the list of valves that can be opened in

Modes 1 through 4, (2) remove a

footnote on Type A testing, and (3)

make editorial changes to the Technical

Specifications and associated Bases

sections.

Basis for proposed no significant

hazards consideration determination:

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:

NNECO has reviewed the proposed

revision in accordance with 10 CFR

50.92 and has concluded that the

revision does not involve a significant

hazards consideration (SHC). The basis

for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does

not involve [an] SHC because the

revision would not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

The proposed change to Technical

Specification Surveillance 4.6.1.1

deletes valves from the list of

containment isolation valves that may

be opened under administrative control.

Deleting the valves, which means that

they are not allowed to be opened under

the Limiting Condition of Operation,

[cannot] cause an accident. The valves

being added in the steam lines to the

steam-driven auxiliary feedwater pump

can be used to heat the steam lines prior

to testing the steam-driven auxiliary

feed water pump. Heating the steam

lines prior to testing the steam-driven

auxiliary feedwater pump does not

increase the likelihood of a steam line

break.

The administrative change of

replacing the ‘‘-’’ with an ‘‘*’’ in the

valve designation can neither cause [an]

accident nor affect the consequences of

any accident.

The addition of the RHR [residual

heat removal] system containment

isolation valves reflects the fact that

these valves can be opened during Mode

4 to allow plant heatup and cooldown.

Plant heatup and cooldown, in

accordance with normal plant operation

and the Technical Specifications, does

not increase the likelihood of the above

accidents.

The administrative controls include

the appropriate considerations that

containment integrity will be

established, when required. By

establishing containment integrity, the

assumptions in the design basis

analyses are assured. This means that

for LOCA [loss-of-coolant accident],

steam line break and feed line break

accidents inside containment, there is

no effect on their consequences.

Valves in the steam lines to the steamdriven auxiliary feedwater pump are

being added to the list of valves allowed

to be opened under administrative

control. This means that these could be

open at the initiation of an accident.

The administrative controls under

which these valves are opened provides

assurance that containment integrity

will be established, when required.

Similarly, for an SGTR [steam generator

tube rupture], Locked Rotor or Control

Rod Ejection event, the administrative

controls provides assurance that these

valves will be closed and, therefore,

allowing them to be opened will not

adversely impact the consequences of

these events. If failure to close is

postulated as a single failure for these

events, the results would be bounded by

the analyses described in the FSAR

[final safety analysis report]. For

example, the Locked Rotor accident

assumes a stuck open steam generator

power-operated pressure relief valve

(SG PORV). The steam released by the

assumed single failure of the SG PORV,

for the twenty minutes until the valve

is isolated, would exceed the expected

releases as a result of failure to close

valve 3MSS*V885, 3MSS*V886, or

3MSS*V887, which are in 1⁄4 inch lines.

Therefore, allowing these valves to be

opened under administrative control

does not effect the consequences of the

previously evaluated accidents.

The FSAR, Section 15.1.5, provides

the assumptions on steam releases for

the consequences of the steam line

break accident. The steam generator

with the broken steam line is assumed

to be open to the atmosphere for the

duration of the event and, therefore,

these valves being open would not

impact that assumption. For the

unaffected steam generators, steam is

assumed released to the atmosphere to

remove decay heat. These valves are in

1⁄4 inch lines which means that any

steam released via this path would only

be a small fraction of decay heat and

will not adversely affect control of

decay heat removal. Therefore, whether

these valves are open or not will not

affect the consequences of a steam line

break outside containment.

Allowing the RHR system

containment isolation valves to be open,

under administrative control in Mode 4,

does not change the way the system is

operated. This proposed change to the

footnote does not change the operators

response to an accident in Mode 4.

Therefore, the addition of these valves

does not affect the consequences of the

previously evaluated accidents.

The proposed change to Technical

Specification Surveillance 4.6.1.2.a will

delete footnote ‘‘*’’ which referred to an

exemption granted by the NRC to permit

the Type A test to be delayed until

RFO6 [refueling outage 6]. However, the

current extended shutdown has

significantly delayed RFO6 and NNECO

intends to perform the Type A test

during this midcycle shutdown. The

deletion of the footnote does not alter

the operation of any system or the

containment or containment airlocks, as

assumed for accident analyses.

Additionally, Technical

Specifications 4.6.1.1, 3/4.6.1.2, and 3/

4.6.1.3, and Bases Sections 3/4.6.1.1, 3/

4.6.1.2, and 3/4.6.1.3 are reworded to

provide clarity and consistency. These

proposed changes do not alter the

operation of any system or the

containment or containment airlocks

during accident analyses.

Therefore, the proposed revision does

not involve a significant increase in the

probability or consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes to Technical

Specifications 4.6.1.1, 3/4.6.1.2, and 3/

4.6.1.3 and Bases Sections 3/4.6.1.1, 3/

4.6.1.2, and 3/4.6.1.3 do not alter the

operation of any system or the

containment or containment airlocks,

during normal operation or as assumed

in accident analyses.

Deleting containment isolation valves

from the list of those that are allowed to

be opened under administrative control

can not modify plant response to an

accident. Adding administrative control

when the RHR system containment

isolation valves are opened in Mode 4

for normal plant cooldown and heatup

can not create a new or different

accident. Allowing valves to be opened
59918 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

to heat the steam lines to the steamdriven auxiliary feedwater pump prior

to testing does not create the possibility

of a new or different accident. The

administrative change to the valve

designation can not modify plant

response.

Therefore, the proposed revision does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes to Technical

Specifications 4.6.1.1, 3/4.6.1.2, and 3/

4.6.1.3, and Bases Sections 3/4.6.1.1, 3/

4.6.1.2, and 3/4.6.1.3 do not alter the

design, maintenance or function of any

system or the containment or the

containment airlocks. Additionally, the

proposed changes do not alter the

testing of any system or the containment

or containment airlocks, or alter any

assumption used in the accident

analyses.

The considerations associated with

administrative control are being added

to the bases of the technical

specification. These considerations are

identical to those provided in GL 91–08

[Generic Letter 91–08]. This means that

the changes will maintain the margin of

safety. The valves that are allowed to be

open in the steam lines to the steamdriven auxiliary feedwater [pump] do

not impact the accident analyses and

therefore do not reduce the margin of

safety. The addition of the RHR system

containment isolation valves reflects the

fact that these valves are opened for

heatup and cooldown in Mode 4. The

change adds the requirements of

administrative controls to these RHR

system valves in Mode 4, but does not

modify the use of these valves. The

administrative change to the valve

designation can not affect the margin of

safety.

Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

In conclusion, based on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT, and the Waterford Library,

ATTN: Vince Juliano, 49 Rope Ferry

Road, Waterford, CT.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT.

NRC Deputy Director: Phillip F.

McKee.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, CT

Date of amendment request: October

15, 1997.

Description of amendment request:

Technical Specification Surveillances

4.1.2.3.1, 4.1.2.4.1, 4.5.2, 4.6.2.1, and

4.6.2.2 require the recirculation spray,

quench spray, residual heat removal,

centrifugal charging, and safety

injection pumps to be tested on a

periodic basis and after modifications

that alter subsystem flow characteristics.

The proposed changes to these

surveillances would include replacing

the specific surveillance pump pressure

with a statement that the test be

conducted in accordance with

Specification 4.0.5, Inservice Testing

Program. The proposed changes would

also include a decrease in the required

individual safety injection and

centrifugal charging pump injection line

flow rates, an increase in the allowed

individual safety injection pump runout

flow rate, and editorial changes to the

surveillances.

Basis for proposed no significant

hazards consideration determination:

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:

NNECO has reviewed the proposed

revision in accordance with 10 CFR

50.92 and has concluded that the

revision does not involve a significant

hazards consideration (SHC). The basis

for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does

not involve an SHC because the revision

would not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

The Technical Specification changes

transfer control of the pump developed

head requirements for the Centrifugal

Charging, Safety Injection, Quench

Spray, Residual Heat Removal, and

Recirculation Spray pumps from the

Technical Specifications to the Inservice

Test program. The acceptance criteria

will still assure that the safety analysis

assumptions are valid. The Technical

Specification changes reduce the

minimum flow requirements for the

Charging and Safety Injection pumps

and increase the maximum allowed

flow for the Safety Injection pumps.

Modifying the surveillance

requirements [cannot] cause an accident

and, therefore, [cannot] increase the

probability of an accident. The revised

minimum required flows are consistent

with the flows used in the accident

analyses and, therefore, the change

[cannot] increase the consequences of

any accident. The safety injection

pumps are disabled such that they

[cannot] be a source of mass addition to

the RCS [reactor coolant system]

whenever the cold overpressure system

is required to be operable. Therefore, the

increase in the allowed maximum safety

injection pump flow has no effect on the

cold overpressure accident analysis.

Therefore, the proposed revision does

not involve a significant increase in the

probability or consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed changes transfer control

of the pump developed head

requirements from the Technical

Specifications to the Inservice Test

program and modify the required flow

surveillance values. The surveillance

values that are used in the Inservice

Test program and the Technical

Specification are consistent with the

accident analysis. The increase in the

allowed maximum safety injection

pump flow does not impact the cold

overpressure accident analysis. The

changes do not involve any changes to

the way that the pumps are operated.

The pumps will be used post-accident

the same way as they are used prior to

the change.

Therefore, the proposed revision does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The control of the pump developed

head acceptance criteria is being

transferred from the Technical

Specification to the Inservice Test

program. The acceptance criteria, at a

minimum, will assure that the design

basis analyses are valid. The minimum

pump flow surveillance requirements in

Specification 4.5.2.h are consistent with

the assumptions of the accident

analysis. The maximum allowed Safety

Injection flow does not exceed the

vendor recommendation for maximum

continuous runout flow. The NPSH [net

positive suction head] available to the

pumps during both the injection and

recirculation phases post-accident
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59919

exceeds the NPSH required at the higher

allowed flow. Also, the safety injection

pumps are disabled so that they [cannot]

be an injection source when the cold

overpressure system is required to be

operable which means that the increase

in maximum flow does not affect the

cold overpressure accident analysis.

Restricting orifices are being installed in

the injection lines from the safety

injection and charging pumps to the

Reactor Coolant System as required. The

restricting orifices and the changes to

the required flows will allow for

resetting the throttle position of the

existing throttle valves. The sizing of the

restricting orifices and the associated rethrottling of the throttle valves will be

in accordance with Regulatory Guide

1.82. The proposed changes allow for

the setting of the throttle valve positions

so that the openings will be larger than

the sump screen mesh opening size

while assuring that the design basis flow

values are valid.

Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

In conclusion, based on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT, and the Waterford Library,

ATTN: Vince Juliano, 49 Rope Ferry

Road, Waterford, CT.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270.

NRC Deputy Director: Phillip F.

McKee.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, NE

Date of amendment request: July 25,

1997.

Description of amendment request:

The proposed amendment request

would revise the Technical

Specifications (TS) to implement 10

CFR Part 50 Appendix J, Option B by

referring to Regulatory Guide 1.163,

‘‘Performance-Based Containment

Leakage-Test Program,’’ with certain

exceptions detailed in the licensee’s

application. This revision supersedes

the staff’s description of amendment

request that was published on October

8, 1997 (62 FR 52586).

Basis for proposed no significant

hazards consideration determination:

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:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed change implements

Option B of 10 CFR Part 50 Appendix

J on performance-based containment

leakage testing. The proposed change

does not involve a change to the plant

design or operation. As a result, the

proposed change does not affect any

parameters or conditions that contribute

to the initiation of any accidents

previously evaluated. The proposed

change potentially affects the leak-tight

integrity of the containment structure

designed to mitigate the consequences

of a Loss-of-Coolant Accident (LOCA).

The function of the containment is to

maintain functional integrity during and

following the peak transient pressures

and temperatures and limit fission

product leakage following the design

basis LOCA. Because the proposed

change does not alter the plant design,

only the frequency of measuring Type

A, B, and C leakage, the proposed

change does not directly result in an

increase in containment leakage.

Test intervals will be established

based on the performance history of

components being tested. The frequency

of monitoring the relatively few

containment isolation valves and/or

containment penetrations subject to

above normal leakage will not decrease

by implementing Option B of Appendix

J. A performance based program will

identify those valves and penetrations

which must continue to be tested each

refueling outage.

The risk resulting from the proposed

changes is characterized as follows,

based primarily on the results contained

in NUREG–1493 ‘‘Performance-Based

Containment Leakage Test Program,’’

the principal Technical Support

Document used by the NRC as the basis

for the Appendix J Final Rule:

Type A Testing

NUREG–1493 found that the effect of

containment leakage on overall accident

risk is minimal since risk is dominated

by accident sequences that result in

failure or bypass of the containment.

Industry wide, Integrated Leak Rate

Tests (ILRTs) have only found a small

fraction of the leaks that exceed current

acceptance criteria. Only three percent

of all leaks are detectable only by ILRTs,

and therefore, by extending the Type A

testing intervals, only three percent of

all leaks have a potential for remaining

undetected for longer periods of time. In

addition, when leakage has been

detected by ILRTs, the leakage rate has

been only marginally above existing

requirements. The Fort Calhoun Station

Unit No. 1 Type A testing confirms the

industry-wide experience that a

majority of the leakage experienced

during Type A testing is through

components tested by Type B and C

tests.

NUREG–1493 found that these

observations, together with the

insensitivity of reactor accident risk to

the containment leakage rate, show that

increasing the Type A leakage test

intervals would have a minimal impact

on public risk.

Type B and C Testing

NUREG–1493 found that while Type

B and C tests can identify the vast

majority (greater than 95 percent) of all

potential leakage paths, performancebased alternatives to current local

leakage-testing requirements are feasible

without significant risk impacts. The

risk model used in NUREG–1493

suggests that the number of components

tested would be reduced by about 60

percent with less than a three-fold

increase in the incremental risk due to

containment leakage. Since, under

existing requirements, leakage

contributes less than 0.1 percent of

overall accident risk, the overall impact

is very small. In addition, the NRC’s

Final Regulatory Impact Analysis

concluded that while the extended

testing intervals for Type B and C tests

led to minor increases in potential

offsite dose consequences, the beneficial

expected decrease in onsite worker dose

received during ILRT and local leak rate

testing exceeds (by at least an order of

magnitude) the potential off-site dose

consequences.

Therefore, the proposed change will

not result in a significant increase in the

probability or consequences of any

accident previously evaluated.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

There will be no physical alterations

to the plant configuration, changes to

setpoint values, or changes to the

implementation of setpoints or limits as

a result of this proposed change. As a

result, the proposed change does not

affect any of the parameters or

conditions that could contribute to

initiation of any accidents.
59920 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

This change involves the reduction of

Type A, B, and C test frequency. Except

for the method of defining the test

frequency, the methods for performing

the actual tests are not changed. No new

accident modes are created by extending

the testing intervals. No safety-related

equipment or safety functions are

altered as a result of this change.

Extending the test frequency has no

influence on, nor does it contribute to,

the possibility of a new or different kind

of accident or malfunction from those

previously analyzed. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

The proposed change only affects the

frequency of Type A, B, and C testing.

Except for the method of defining the

test frequency, the methods for

performing the actual tests are not

changed.

The frequency of monitoring the

relatively few containment isolation

valves and/or containment penetrations

subject to above normal leakage will not

decrease by implementing Option B of

Appendix J. A performance based

program will identify those valves and

penetrations which must continue to be

tested each refueling outage. NUREG–

1493 has determined that, under several

different accident scenarios, the

increased risk of radioactivity release

from containment is negligible with the

implementation of these proposed

changes.

The margin of safety that has the

potential of being impacted by the

proposed change involves the offsite

dose consequences of postulated

accidents which are directly related to

containment leakage rate. The

containment isolation system is

designed to limit leakage to La, which

is stated in the Fort Calhoun Station

Unit No. 1 Technical Specifications to

be 0.1 percent by weight of the

containment air per 24 hours at 60 psig.

The limitation on containment

leakage rate is designed to ensure that

total leakage volume will not exceed the

value assumed in the accident analyses

at the peak accident pressure. The

margin to safety for the offsite dose

consequences of postulated accidents

directly related to the containment

leakage rate is maintained by meeting

the 1.0 La acceptance criteria. The La

value is not being modified by this

proposed change.

Except for the method of defining the

test frequency, no change in the method

of testing is being proposed. The Type

B and C tests will continue to be done

at 60 psig or greater. Other programs are

in place to ensure that proper

maintenance and repairs are performed

during the service life of the primary

containment and systems and

components penetrating the primary

containment.

Therefore, the proposed change will

not result in a significant reduction in

a margin of safety.

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.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, NE 68102.

Attorney for licensee: Perry D.

Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005–

3502.

NRC Project Director: William H.

Bateman.

Power Authority of the State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, NY

Date of amendment request:

September 3, 1997.

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TSs) to revise the number of hours

operating personnel can work in a

normal shift. The proposed amendment

also contains some administrative

changes to the TSs.

Basis for proposed no significant

hazards consideration determination:

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:

1. Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

A. Establishing operating personnel

work hours at, ‘‘an 8 to 12 hour day,

nominal 40 hour week,’’ allows normal

plant operations to be managed more

effectively and does not adversely effect

performance of operating personnel.

Overtime remains controlled by site

administrative procedures in

accordance with NRC Policy Statement

on working hours (Generic Letter 82–

12). If 8 hour shifts are maintained in

part or whole, then acceptable levels of

performance from operating personnel

is assured through effective control of

shift turnovers and plant activities. No

physical plant modifications are

involved and none of the precursors of

previously evaluated accidents are

affected. Therefore, this change will not

involve a significant increase in the

probability or consequence of an

accident previously evaluated.

B. Editorial changes clarify section

6.2.2.g without changing the intent or

meaning. The proposed change meets

the intent of the NRC Policy Statement

on working hours (Generic Letter 82–

12).

C. Changes to sections 3.10.6.1.a and

3.10.9 do not change the intent or

meaning of the technical specification

sections. Clarification to the table

notation in section 4.1 related to the

definition of shift checks to monitor

plant conditions will continue as

intended but are allowed to increase up

to at least once per 12 hours. This

increase is consistent with standard

industry practice as represented by the

Standard Technical Specifications

(STS), Reference 1.

2. Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

A. Establishing operating personnel

work hours at, ‘‘an 8 to 12 hour day,

nominal 40 hour week,’’ allows normal

plant operations to be managed more

effectively and does not adversely effect

performance of operating personnel. If 8

hour shifts are maintained in part or

whole, then acceptable levels of

performance from operating personnel

is assured through effective control of

shift turnovers and plant activities.

Overtime remains controlled by site

administrative procedures in

accordance with the NRC Policy

Statement on working hours (Generic

Letter 82–12). No physical modification

of the plant is involved. As such, the

change does not introduce any new

failure modes or conditions that may

create a new or different accident.

Therefore, operation in accordance with

the proposed amendment will not create

the possibility of a new or different kind

of accident from any previously

evaluated.

B. Editorial changes clarify section

6.2.2.g without changing the intent or

meaning. The proposed change meets

the intent of the NRC Policy Statement

on working hours (Generic Letter 82–

12).

C. Changes to sections 3.10.6.1.a and

3.10.9 do not change the intent or

meaning of the technical specification

sections. Clarification to the table

notation in section 4.1 related to the

definition of shift checks to monitor

plant conditions will continue as

intended but are allowed to increase up
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59921

to at least once per 12 hours. This

increase is consistent with standard

industry practice as represented by the

Standard Technical Specifications

(STS), Reference 1.

3. Does the proposed amendment

involve a significant reduction in a

margin of safety?

A. Establishing operating personnel

work hours at, ‘‘an 8 to 12 hour day,

nominal 40 hour week,’’ allows normal

plant operations to be managed more

effectively and does not adversely effect

performance of operating personnel. If 8

hour shifts are maintained in part or

whole, then acceptable levels of

performance from operating personnel

is assured through effective control of

shift turnovers and plant activities.

Overtime remains controlled by site

administrative procedures in

accordance with the NRC Policy

Statement on working hours (Generic

Letter 82–12) and is consistent with the

Standard Technical Specifications. The

proposed change involves no physical

modification of the plant, or alterations

to any accident or transient analysis.

There is no Basis to section 6 of the

Technical Specifications, and the

changes are administrative in nature.

Therefore, the change does not involve

any significant reduction in a margin of

safety.

B. Editorial changes clarify section

6.2.2.g without changing the intent or

meaning. The proposed change meets

the intent of the NRC Policy Statement

on working hours (Generic Letter 82–

12).

C. Changes to sections 3.10.6.1.a and

3.10.9 do not change the intent or

meaning of the technical specification

sections. Clarification to the table

notation in section 4.1 related to the

definition of shift checks to monitor

plant conditions will continue as

intended but are allowed to increase up

to at least once per 12 hours. This

increase is consistent with standard

industry practice as represented by the

Standard Technical Specifications

(STS), Reference 1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, NY

10601.

Attorney for licensee: Mr. David

Blabey, 10 Columbus Circle, New York,

NY 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Rochester Gas and Electric Corporation,

Docket No. 50–244, R. E. Ginna Nuclear

Power Plant, Wayne County, NY

Date of amendment request:

September 29, 1997, as supplemented

October 8, 1997. The September 29

application and October 8, 1997,

supplement supersede the September

13, 1996, application and its April 24,

1997, supplement. This notice

supersedes the notice published on

October 9, 1996 (61 FR 197) in its

entirety.

Description of amendment request:

The proposed amendment would

change the Ginna Station Technical

Specifications (TSs) which would allow

referencing of revision of the Ginna

Station pressure and temperature limits

report (PTLR) for the reactor coolant

system (RCS) pressure and temperature

(P/T) limits and low temperature

overpressure protection (LTOP) limits.

The proposed amendment would

correct some typographical errors in the

TSs.

Basis for proposed no significant

hazards consideration determination:

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:

1. Operation of Ginna Station in

accordance with the proposed changes

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed changes revise

Administrative Controls Section 5.6.6.c

to update the reference to the NRC’s

approval of the first use of the PTLR

methodology, update the RCS P/T

methodology to the final NRC approved

version, allow use of ASME Code Case

N–514 for LTOP enable temperature

methodology, and to correct a

typographical error. These changes

complete implementation of Generic

Letter 96–03 by referencing NRC

approved methodology within the

Administrative Controls. The updated

RCS P/T methodology has been

generically approved by the NRC while

the use of ASME Code Case N–514 for

LTOP enable temperature methodology

was previously approved for use at

Ginna Station by the NRC. As such,

these changes are administrative in

nature and do not impact initiators or

analyzed events or assumed mitigation

of accident or transient events.

Therefore, these changes do not involve

a significant increase in the probability

or consequences of an accident

previously analyzed.

2. Operation of Ginna Station in

accordance with the proposed changes

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated. The

proposed changes do not involve a

physical alteration of the plant (i.e., no

new or different type of equipment will

be installed) or changes in the methods

governing normal plant operation. The

proposed changes will not impose any

new or different requirements. Thus,

this change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Operation of Ginna Station in

accordance with the proposed changes

does not involve a significant reduction

in a margin of safety. The proposed

changes will not reduce a margin of

plant safety because the methodology

have been shown to ensure that the P/

T and LTOP limits in the PTLR continue

to meet all necessary requirements for

reactor vessel integrity. These changes

are administrative in nature since the

limits were previously relocated to the

PTLR under a separate LAR [License

Amendment Request]. As such, no

question of safety is involved, and the

change does not involve a significant

reduction in a margin of safety.

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.

Local Public Document Room

Location: Rochester Public Library, 115

South Avenue, Rochester, NY 14610.

Attorney for licensee: Nicholas S.

Reynolds, Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005.

NRC Project Director: S. Singh Bajwa,

Director.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

CA

Date of amendment requests:

December 22, 1995.

Description of amendment requests:

The licensee proposes to delete the

physical protection program reporting

requirement from License Condition

2.G, and to clarify in License Condition

2.E that all the documents composing

the physical protection program plans

may not contain safeguards information.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the
59922 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

This proposed change is considered

an administrative change. It has no

impact on the probability or

consequences of any of the accidents

previously evaluated. This change

revises license conditions for

clarification and removes the burden of

duplicate reporting requirements. This

change does not affect the physical

protection program as previously

approved by the Nuclear Regulatory

Commission (NRC). License Condition

2.E is being revised to clarify that the

physical security, security force training

and qualification, and safeguards

contingency plans may or may not

contain safeguards information. The

security force training and qualification

plan does not currently contain

safeguards information.

A reporting requirement in License

Condition 2.G is being revised to

remove the reference to License

Condition 2.E for the physical

protection program. The reporting

requirements for the physical protection

program are located in the regulations,

10 CFR 73.71 and 10 CFR 73 part,

Appendix G.

Therefore, the probability and

consequences of an accident previously

evaluated are not affected by these

proposed changes.

2. The proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

This proposed change is considered

an administrative change. It has no

impact on equipment, systems, or

structures such that a new or different

kind of accident is created. This change

revises license conditions to clarify that

safeguards information may be located

in the physical protection program

plans and to remove duplicate and

unnecessary reporting requirements for

the physical protection program. There

is no change associated with the

implementation and maintenance of the

physical protection program as

previously approved by the NRC.

Therefore, the possibility of a new or

different kind of accident from an

accident previously evaluated is not

created.

3. The proposed change does not

involve a significant reduction in a

margin of safety.

This proposed change is considered

an administrative change only. It has no

impact on the margin of safety

associated with the physical protection

program. This change revises license

conditions to clarify the location of

safeguards information in the physical

protection program plans and remove

duplicative and unnecessary reporting

requirements for the physical protection

program. The maintenance and

implementation of the physical

protection program is not affected by

this change.

Therefore, there will not be a

significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine, CA

92713.

Attorney for licensee: T.E. Oubre,

Esquire, Southern California Edison

Company, P.O. Box 800, Rosemead, CA

91770.

NRC Project Director: William H.

Bateman.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50–440, Perry Nuclear Power

Plant, Unit 1, Lake County, OH

Date of amendment request: October

22, 1997.

Description of amendment request:

The amendment would change the Perry

Nuclear Power Plant design basis as

described in the Updated Safety

Analysis Report. The change will add a

description of the temperature control

valves and associated bypass lines

around the Emergency Closed Cooling

System heat exchangers. These features

are designed to ensure operability of the

Control Complex Chilled Water System

under post-accident load conditions,

without the need for compensatory

actions.

Basis for proposed no significant

hazards consideration determination:

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:

1. The proposed change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed amendment is

requesting Nuclear Regulatory

Commission (NRC) review and approval

of changes to the Perry Nuclear Power

Plant (PNPP) Updated Safety Analysis

Report (USAR) to incorporate

descriptions (in the form of text, tables

and drawings) of a modification to the

plant involving two temperature control

valves and associated temperature

elements, and piping segments that have

been installed in the Emergency Closed

Cooling Water (ECC) System. These

valves, temperature elements, and

piping segments were installed to

increase the overall reliability of the

ECC System and the other safety related

plant systems that it serves, to help

ensure that they perform their specified

safety functions without reliance on

manual throttling actions.

The probability of occurrence and the

consequences of an accident previously

evaluated in the USAR are not

considered to be increased as a result of

the temperature control valve

modification.

Based on conformance with the

original system design criteria, the fact

that the ECC System is an accident

mitigation system, and that this

modification does not introduce any

new initiators to a previously postulated

accident, the addition of this

temperature control function can not

increase the probability of occurrence of

an accident previously evaluated in the

USAR. Accidents reviewed involve the

Loss of Coolant Accident applications

described in USAR Chapter 6 with their

corresponding consequence

postulations shown in USAR Chapter

15, accident and transient scenarios as

described in USAR Chapter 15, flooding

and rupture postulations as described in

USAR Chapter 3, and fire protection

analyses as described in USAR Chapter

9.

The modification has been designed,

procured, and installed to the original

design codes and standards. The

modification also satisfies single failure

criteria and does not adversely affect the

mitigation function of the ECC System.

Therefore, the ability to mitigate

accidents previously evaluated in the

USAR is maintained and the

radiological consequences of such

accidents remain unaffected.

Therefore, the proposed changes do

not involve a significant increase in the

probability or consequences of

previously evaluated accidents.

2. The proposed change would not

create the possibility of a new or

different kind of accident from any

previously evaluated.

The modification has been designed

to satisfy the requirements of the

original ECC System. A single failure of

the new configuration will not result in

more than the loss of one respective
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59923

ECC System loop as already analyzed.

Analysis of flooding shows no scenario

greater than the currently bounding

event. Missile generation is not a

concern since no mechanisms

conducive to that potential have been

introduced. From the electrical analysis

perspective, analysis has shown no

adverse effects on the Emergency Diesel

Generator loadings or other system

applications.

Based on the above discussions, the

proposed change would not create the

possibility of a new or different kind of

accident than those previously

evaluated.

3. The proposed change will not

involve a significant reduction in the

margin of safety.

This request does not involve a

significant reduction in a margin of

safety. The modification, including

design, procurement, and installation,

has been performed in accordance with

the applicable codes, standards, and

installation specifications. The

modification does not change the heat

removal capabilities or any previously

designed parameters of the ECC System.

Hence, the ECC System margin of safety

with respect to safety classification,

protection, redundancy, heat removal

capability, and seismic classification

remains unaffected.

The margins of safety contained in the

Technical Specifications and the

associated Bases also remain unaffected

by this modification due to conformance

with the applicable codes, standards,

and installation specifications.

Specifically, Technical Specification

3.7.10, ‘‘Emergency Closed Cooling

Water (ECCW) System’’ and the

description in the Bases remain

unchanged and fully applicable. The

following Technical Specifications also

remain unaffected and applicable:

3.3.3.2, ‘‘Remote Shutdown System’’;

3.7.1, ‘‘Emergency Service Water (ESW)

System—Divisions 1 and 2’’; 3.7.4,

‘‘Control Room Heating, Ventilation,

and Air Conditioning (HVAC) System’’;

and the Technical Specifications related

to Sections 3.8 (Electrical Power

Systems), 3.5 (Emergency Core Cooling

Systems (ECCS) and Reactor Core

Isolation Cooling (RCIC) System) and

3.6 (Containment Systems). On this

basis, the margins of safety defined in

the Technical Specifications remain

unchanged.

Therefore, the changes associated

with this license amendment request do

not involve a significant reduction in

the margin of safety.

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.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Gail H. Marcus.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, SC

Date of application for amendment:

August 27, 1996, as supplemented

December 18, 1996, January 17,

February 18, March 27, April 4, April

25, April 29, May 30, June 2, June 13,

June 18, August 4, August 8, September

10, October 2 (RNP RA/97–0216),

October 2, (RNP RA/97–0207), October

13, and October 21, 1997.

Brief description of amendment: This

amendment addresses a more restrictive

change proposed by the licensee in

minimum allowable containment

pressure.

Date of publication of individual

notice in Federal Register: October 7,

1997 (62 FR 52362).

Expiration date of individual notice:

October 21, 1997.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, SC

29550.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, NJ

Date of amendment request:

September 24, 1997.

Brief description of amendment

request: The proposed amendment

would add a surveillance requirement

in Section 3/4.5.1 to perform a monthly

valve position verification for each of

the four residual heat removal crosstie

valves.

Date of publication of individual

notice in Federal Register: October 6,

1997 (62 FR 52162).

Expiration date of individual notice:

November 5, 1997.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, NJ

Date of amendment request:

September 29, 1997.

Brief description of amendment

request: The proposed amendment

would change Technical Specification

3/4.11.1, ‘‘Liquid Effluents—

Concentration.’’ The proposed change

adds a requirement to perform weekly

sampling and monthly and quarterly

composite analyses of the Station

Service Water System when the Reactor

Auxiliaries Cooling System is

contaminated.

Date of publication of individual

notice in Federal Register: October 6,

1997 (62 FR 52161).

Expiration date of individual notice:

November 5, 1997.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in
59924 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are 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 rooms for the

particular facilities involved.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 & 50–324,

Brunswick Steam Electric Plant, Units 1

& 2, Brunswick County, NC

Date of amendment request: January

7, 1997, as supplemented on July 25,

1997, August 27, 1997, and September

15, 1997.

Brief description of amendment: The

amendments correct an error involving

the transposition of two of the reactor

pressure vessel (RPV) pressuretemperature (P–T) limits curves between

the Technical Specifications for the

Brunswick Steam Electric Plant, Units 1

and 2 and update the hydrostatic

pressure test limits curves for both

units.

Date of issuance: October 7, 1997.

Effective date: October 7, 1997.

Amendment No.: 189 and 220.

Facility Operating License Nos. DPR–

71 and DPR–62: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: March 12, 1997 (62 FR 11485).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 7, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, NC 28403–3297.

Carolina Power & Light Company,

Docket No. 50–261, H.B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, SC

Date of application for amendment:

August 27, 1996, as supplemented

December 18, 1996, January 17,

February 18, March 27, April 4, April

25, April 29, May 30, June 2, June 13,

June 18, August 4, August 8, September

10, October 2 (RNP RA/97–0216),

October 2, (RNP RA/97–0207), October

13, and October 21, 1997.

Brief description of amendment: This

amendment addresses a more restrictive

change proposed by the licensee in

minimum allowable containment

pressure.

Date of issuance: October 24, 1997.

Effective date: October 24, 1997.

Amendment No.: 176.

Facility Operating License No. DPR–

23: Amendment revises the License and

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (62 FR 52362

dated October 7, 1997). The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by November 6,

1997, but indicated that if the

Commission makes a final NSHC

determination, any such hearing would

take place after issuance of the

amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of NSHC are contained in a Safety

Evaluation dated October 24, 1997.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, SC

29550.

Carolina Power & Light Company, et al.,

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, NC

Date of application for amendment:

February 21, 1997.

Brief description of amendment: This

amendment adds a specific time limit to

Technical Specification Table 3.3–3 to

place an inoperable refueling water

storage tank level channel in a bypassed

condition.

Date of issuance: September 30, 1997.

Effective date: September 30, 1997.

Amendment No: 74.

Facility Operating License No. NPF–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17225).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated September 30, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, NC

27605.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, IL

Date of application for amendments:

March 5, 1997 as supplemented October

3, 1997.

Brief description of amendments: The

amendments would revise the Technical

Specifications by removing the main

steamline radiation monitor reactor

scram function and the main steamline

tunnel radiation isolation function.

Date of issuance: October 24, 1997.

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 163, 158.

Facility Operating License Nos. DPR–

19 and DPR–25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 18, 1997 (62 FR 19141).

The October 3, 1997, submittal provided

additional clarifying information that

did not change the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

October 24, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris, IL

60450.

Entergy Gulf States, Inc., Cajun Electric

Power Cooperative, and Entergy

Operations, Inc., Docket No. 50–458,

River Bend Station, Unit 1, West

Feliciana Parish, LA

Date of amendment request: August

29, 1996, supplemented August 29,

1996 (proprietary), September 5, and

October 8, 1997.

Brief description of amendment: The

amendment eliminates the Average

Power Range Monitor (APRM) setpoint

T-Factor setdown requirements and

provides for reactivity anomaly

calculation improvements. The request

to decrease the local power range
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59925

monitor (LPRM) calibration frequency

will be handled by separate review and

action.

Date of issuance: October 10, 1997.

Effective date: October 10, 1997.

Amendment No.: 100.

Facility Operating License No. NPF–

47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 23, 1997 (61 FR

55032). The Licensee’s letters dated

August 29, 1996 (proprietary),

September 5, and October 8, 1997,

provided additional clarification and

corrections to other TSs that would have

erroneously referenced the TSs being

eliminated and did not change the

staff’s initial no significant hazards

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County,

PA

Date of application for amendment:

July 30, 1997, as supplemented

September 19, and September 24, 1997.

Brief description of amendment: The

amendment reduces current technical

specification leakage limit from the

decay heat removal system from 6.0

gallons per hour (gph) to 0.6 gph.

Date of issuance: October 15, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 205.

Facility Operating License No. DPR–

50: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 27, 1997 (62 FR

45458). The September 19, and

September 24, 1997, submittals did not

affect the initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 15,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

GPU Nuclear Corporation, et al., Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County,

PA

Date of application for amendment:

August 12, 1997, as supplemented

August 28, September 15, October 3, 9,

and 10, 1997.

Brief description of amendment: The

amendment changes the technical

specifications surveillance requirements

for once-through steam generator

inservice inspection for Cycle 12

operation.

Date of issuance: October 16, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 206.

Facility Operating License No. DPR–

50: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 27, 1997 (62 FR

45458). The supplemental letters did

not affect the initial no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 16,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Law/Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box

1601, Harrisburg, PA 17105.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, TX, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, TX

Date of amendment request: August

14, 1997, as supplemented September

23, 1997. The supplement provided

clarifying information within the scope

of the amendment request and did not

change the initial no significant hazards

consideration determination.

Brief description of amendments: The

amendments revise the allowed

tolerance of the reactor coolant system

volume provided in Technical

Specification 5.4.2 to account for steam

generator tube plugging.

Date of issuance: October 20, 1997.

Effective date: October 20, 1997.

Amendment Nos.: Unit 1—

Amendment No. 92; Unit 2—

Amendment No. 79.

Facility Operating License Nos. NPF–

76 and NPF–80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 26, 1997 (62 FR

45278). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 20, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J.M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488.

Northeast Nuclear Energy Company,

Docket No. 50–245, Millstone Nuclear

Power Station, Unit 1, New London

County, CT

Date of application for amendment:

February 7, 1997, as supplemented

April 3 and September 19, 1997.

Brief description of amendment: The

amendment clarifies the requirement for

calibration of instrument channels that

use resistance temperature detectors or

thermocouples.

Date of issuance: October 22, 1997.

Effective date: As of the date of

issuance, to be implemented within 90

days.

Amendment No.: 102.

Facility Operating License No. DPR–

21: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17236).

The April 3 and September 19, 1997,

letters provided additional and

clarifying information that did not

change the scope of the February 7,

1997, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 22,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT, and at the Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, CT

Date of application for amendment:

June 19, 1997.

Brief description of amendment:

Technical Specification Table 2.2–1

NOTES 1 and 3 define the values for the

constants used in the Overtemperature

Delta-T and Overpower Delta-T reactor

trip system instrumentation setpoint

calculators. The amendment makes

changes to the NOTES as well as the

associated Bases section.
59926 Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices

Date of issuance: October 22, 1997.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 152.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40852).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 22, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT, and the Waterford Library,

ATTN: Vince Juliano, 49 Rope Ferry

Road, Waterford, CT.

Northern States Power Company,

Docket Nos. 50–282 and 50–306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County, MI

Date of application for amendments:

November 6, 1996, as supplemented

April 10 and October 1, 1997.

Brief description of amendments: The

amendments revise Technical

Specifications governing the cooling

water system and are a partial response

to the licensee’s application. The

changes improve plant operation based

on operational experience with the

vertical motor-driven cooling water

pump. The changes also incorporate

information gathered by the licensee

during its self-assessment Service Water

System Operational Performance

Inspection (SWSOPI) completed in late

1995. The remainder of the licensee’s

application will be addressed in a

separate licensing action.

Date of issuance: October 21, 1997.

Effective date: October 21, 1997, with

full implementation within 90 days.

Amendment Nos.: 131 and 123.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4338)

The April 10 and October 1, 1997,

letters provided clarifying information

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards considerations determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 21,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis, MI

55401.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Docket No. 50–

278, Peach Bottom Atomic Power

Station, Unit No. 3, York County, PA

Date of application for amendment:

June 30, 1997, as supplemented by letter

dated September 26, 1997.

Brief description of amendment:

Revises the minimum critical power

ratio (MCPR) safety limit in Section 2.1

of the Technical Specifications from

1.07 to 1.11 for two recirculation loops

in operation. For a single loop in

operation, the MCPR will change from

1.08 to 1.12. The new MCPR safety

limits reflect the effect of the new

General Electric—13 part length fuel

design and other Peach Bottom corespecific parameters.

Date of issuance: October 9, 1997.

Effective date: As of the date of

issuance, to be implemented prior to

startup from Unit 3 refueling outage

3R11.

Amendment No.: 225.

Facility Operating License No. DPR–

56: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR

43373).

The supplemental letter provided

clarifying information that did not

change the original no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 9, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, PA

Date of application for amendments:

April 9, 1997.

Brief description of amendments:

These amendments revise the TSs to

clarify existing battery-specific gravity

requirements, delete the requirement to

correct specific gravity values based on

electrolyte level, and allow the use of

charging current measurements to verify

the battery’s state of charge.

Date of issuance: October 8, 1997.

Effective date: Both units, as of date

of issuance and shall be implemented

within 30 days.

Amendment Nos.: 123 and 88.

Facility Operating License Nos. NPF–

39 and NPF–85: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30643).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 8, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, AL

Date of amendments request: March

7, 1997.

Brief Description of amendments: The

amendments change the Technical

Specifications for both Farley units to

allow operability testing for certain

containment isolation valves during

defueled status.

Date of issuance: October 17, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1—130; Unit

2—123.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: April 23, 1997 (62 FR 19834).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 17,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, AL 36302.

Southern Nuclear Operating Company,

Inc., Docket No. 50–348, Joseph M.

Farley Nuclear Plant, Unit 1, Houston

County, AL

Date of amendment request:

September 3, 1997.

Brief Description of amendment: The

changes reduce the number of required

incore detectors necessary for continued

operation for the remainder of Cycle 15

only.

Date of issuance: October 23, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 131.
Federal Register / Vol. 62, No. 214 / Wednesday, November 5, 1997 / Notices 59927

Facility Operating License No. NPF–2:

Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: September 10, 1997 (62 FR

47695).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 23,

1997.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, AL.

Tennessee Valley Authority, Docket No.

50–390 Watts Bar Nuclear Plant, Unit 1,

Rhea County, TN

Date of application for amendment:

June 20, 1997.

Brief description of amendment:

Modify the Watts Bar Technical

Specifications (TS) to incorporate the

use of Code Case N–514 into the

methodology for the PressureTemperature Limits Report.

Date of issuance: October 21, 1997.

Effective date: October 21, 1997.

Amendment No.: 9.

Facility Operating License No. NPF–

90: Amendment revises the TS.

Date of initial notice in Federal

Register: September 10, 1997 (62 FR

47700).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 21,

1997.

No significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

TN 37402.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, KS

Date of amendment request: July 3,

1997, as supplemented by letter dated

August 20, 1997.

Brief description of amendment: The

amendment revises Surveillance

Requirements 4.3.1.2 and 4.3.2.2, and

Technical Specifications 3/4.3.1 and 3/

4.3.2, and associated Bases Sections B 3/

4.3.1 and B 3/4.3.2 to eliminate periodic

response time testing requirements for

selected pressure and differential

pressure sensors in the reactor trip

system and engineered safety features

actuation system instrumentation

channels.

Date of issuance: October 20, 1997.

Effective date: October 20, 1997, to be

implemented prior to restart from the

ninth refueling outage currently

scheduled to start on October 4, 1997.

Amendment No.: 113.

Facility Operating License No. NPF–

42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40862).

The August 20, 1997, supplemental

letter provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated October 20, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, KS 66801

and Washburn University School of Law

Library, Topeka, KS 66621.

Dated at Rockville, Maryland, this 29th day

of October 1997.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 97–29138 Filed 11–4–97; 8:45 am]

BILLING CODE 7590–01–P

THE PRESIDENT’S COUNCIL ON

SUSTAINABLE DEVELOPMENT

The Eighteenth Meeting of the

President’s Council on Sustainable

Development (PCSD) in Atlanta,

Georgia

Summary: The President’s Council on

Sustainable Development (PCSD), a

Presidential Commission with

representation from industry,

government, environmental, and Native

American organizations, will convene

its eighteenth meeting in Atlanta,

Georgia on Thursday, November 20,

1997.

Under its current charter, the

Administration asked the Council to

continue its work by continuing to forge

consensus on policy, demonstrating

implementation, getting the word out

about sustainable development, and

evaluating progress. The Council will

advise the President in four specific

areas: domestic implementation of

policy options to reduce greenhouse gas

emissions, next steps in building the

new environmental management system

of the 21st century, promoting multijurisdictional and community

cooperation in metropolitan and rural

areas, and policies that fosters the

United States’ leadership role in

sustainable development

internationally.

At the Council’s last meeting in Tulsa,

Oklahoma on September 22, 1997,

members were briefed on the science

impacts, technology impacts, and

economics related to climate change.

The Council also heard from Tulsa’s

community about ways in which the

climate change issue affects their lives.

At this next meeting, the Council will

receive input from a community forum

on climate change, focus on technology

options to reduce greenhouse gas

emissions and hear from a series of

experts in the field. Specifically, the

discussion will address the following

agenda items:

• Current sources of greenhouse gas

emissions; and

• Technology opportunities in a

variety of sectors within the United

States economy to reduce greenhouse

gas emissions.

Public comment period: The Council

will seek public comment on potential

Council activities to implement the

Administration’s directive.

Specifically, the Council is interested

in hearing from the public on the

following questions:

• How might climate change affect

the quality of life in the Atlanta region?

• Are there local opportunities in

Atlanta, Georgia and surrounding

regions to reduce greenhouse gas

emissions?

• What policy recommendations

should the Council give to President

Clinton to more quickly develop and

deploy energy efficient technologies?

The Council’s previous

recommendations to the President may

be found in two reports: Sustainable

America: A New Consensus for

Prosperity, Opportunity, and a Healthy

Environment for the Future (March

1996) and Building on Consensus: A

Progress Report on Sustainable America

(January 1997). Copies of both reports

can be ordered by calling 1–800–363–

3732 or downloaded off the Internet at

‘‘http://www.whitehouse.gov/PCSD’’.

Dates/Times: Thursday, November 20,

1997 from 9:00 a.m. to 1:00 p.m.

Place: Georgia Public Broadcasting

Building, 206 14th Street in the main

floor television studio, Atlanta, Georgia,

30318. PH: 404–685–2253; FAX: 404–

756–2417.

Status: Open to the public. Public

comments are welcome and may be

submitted orally on November 20 or in

writing any time prior to or during the

meeting. Please submit written

comments prior to meeting to: PCSD,

Public Comments, 730 Jackson Place,

NW, Washington, D.C. 20503, or fax to: