Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. 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 8, 1999, through October 22, 1999.

The last biweekly notice was published on October 20, 1999 (64 FR

56526).

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

[[Page 59797]]

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 Administration 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 6D59, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The

filing of requests for a hearing and petitions for leave to intervene

is discussed below.

By December 10, 1999, 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 electronically from

the ADAMS Public Library component on the NRC Web site,

http://www.nrc.gov (the Electronic Reading Room). 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 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 electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Carolina Power & Light Company, et al., Docket No. 50-325,

Brunswick Steam Electric Plant, Unit 1, Brunswick County, North

Carolina

Date of amendment request: September 28, 1999.

Description of amendment request: The licensee has proposed to

revise Technical Specification (TS) 2.1.1, ``Reactor Core Safety

Limits,'' and TS 5.6.5, ``Core Operating Limits Report.'' These

revisions would remove cycle-specific safety limit restrictions which

are no longer necessary.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the

[[Page 59798]]

issue of no significant hazards consideration, which is presented

below:

1. The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The procedures for determining the MCPR [Minimum Critical Power

Ratio] Safety Limit are described in General Electric Standard

Application for Reactor Fuel (i.e., topical report NEDE-24011-P-A,

otherwise referred to as GESTAR II). The basis for the MCPR Safety

Limit calculation is to ensure that greater than 99.9 percent of all

fuel rods in the core avoid transition boiling in the event of a

postulated accident. The existing MCPR Safety Limit preserves this

margin to transition boiling and fuel damage. The MCPR Safety Limits

for the BSEP [Brunswick Steam Electric Plant], Unit 1 TSs, and their

use in determining cycle-specific operating limits documented in the

Core Operating Limits Report, are determined using NRC-approved

methods (i.e., GESTAR II). The use of these methods ensures that the

MCPR Safety Limit values are within the existing design and

licensing bases, and cannot increase the probability or consequences

of an accident previously evaluated.

2. The proposed license amendment will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

The MCPR Safety Limit is a TS numerical value that has been

established to ensure that fuel damage from transition boiling does

not occur in at least 99.9 percent of the fuel rods in the core as a

result of a limiting postulated accident. The MCPR Safety Limit is

not an accident initiator; therefore, it cannot create the

possibility of any new type of accident. The MCPR Safety Limits are

calculated using NRC-approved methods. The function, location,

operation, and handling of the fuel will remain unchanged. In

addition, the initiating sequence of events for previously evaluated

accidents has not been changed. Therefore, no new or different kind

of accident has been created.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

The MCPR Safety Limit preserves the existing margin to

transition boiling and fuel damage in the event of a postulated

accident. The margin of safety, as defined in the TS Bases, will

remain the same. The MCPR Safety Limit remains unchanged, and will

ensure that greater than 99.9 percent of all fuel rods in the core

will avoid transition boiling if the limit is not violated, thereby

preserving the fuel cladding integrity. The MCPR Safety Limits will

continue to be calculated using NRC-approved generic and cycle-

specific methodologies that are described in GESTAR II. 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: William D. Johnson, Vice President and

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602

NRC Section Chief: Ron Hernan, Acting.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: September 28, 1999.

Description of amendment request: The amendment revises Technical

Specifications (TS) surveillance requirement (SR) 3.7.6.2 ``Component

Cooling Water (CCW) System,'' to change the CCW pump automatic start

actuation signal basis from Engineered Safety Feature Actuation Signal

(ESFAS) to Loss-of-Power Diesel Generator (LOP DG). This change is

required to reflect the original plant design which was not properly

incorporated during conversion of the TS to Improved TS.

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:

Carolina Power & Light (CP&L) Company has evaluated the proposed

Technical Specification change and has concluded that it does not

involve a significant hazards consideration. The CP&L conclusion is

in accordance with the criteria set forth in 10 CFR 50.92. The bases

for the conclusion that the proposed change does not involve a

significant hazards consideration are discussed below.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change to Surveillance Requirement (SR) 3.7.6.2

does not involve any physical alteration of plant systems,

structures or components, changes in parameters governing normal

plant operation, or methods of operation. The safety function of the

Loss of Power (LOP) Diesel Generator (DG) start signal for the

Component Cooling Water (CCW) pumps is to start the CCW pumps in

order to provide the minimum heat removal capability assumed in the

safety analysis for the systems to which it supplies cooling water.

The CCW System provides a heat sink for the removal of process and

operating heat from safety related components during a Design Basis

Accident (DBA) or transient. During normal operation, the CCW System

also provides this function for various nonessential components, as

well as the spent fuel storage pool. The CCW System serves as a

barrier to the release of radioactive byproducts between potentially

radioactive systems and the Service Water System, and thus to the

environment. The CCW pumps start upon receipt of a LOP DG start

signal from undervoltage on the emergency bus. The LOP DG start

signal to the CCW pumps is not an Engineered Safety Features

Actuation System (ESFAS) signal. Since this proposed change only

corrects the description of the start signal, the proposed change

does not involve an increase in the probability or consequences of

an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. The

proposed change does not introduce a new mode of operation or

changes in the method of normal plant operation. Therefore, the

possibility of a new or different kind of accident from any accident

previously evaluated is not created.

3. Does this change involve a significant reduction in a margin

of safety?

The proposed change corrects the word description of the start

signal for the CCW pumps and does not alter any plant design margin

or analysis assumption as described in the Updated Safety Analysis

Report. The proposed change does not affect any limiting safety

system setpoint, calibration method, or setpoint calculation.

Therefore, the proposed change does not involve a 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.

Attorney for licensee: William D. Johnson, Vice President and

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602 .

NRC Section Chief: Sheri R. Peterson.

CBS Corporation (licensee), Westinghouse Test Reactor, Waltz Mill

Site, Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-

2

Date of amendment request: September 15, 1999, as supplemented on

October 4, 1999.

Description of amendment request: CBS Corporation is the licensee

for the Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania.

The licensee is authorized to only possess the reactor and a

decommissioning plan has been approved.

The licensee is planning to revise four Technical Specifications

(TS) in their approved Decommissioning Plan. The

[[Page 59799]]

first TS change deals with what doors need to be closed when restricted

activities are taking place within containment. Access to containment

is through three locations, i.e., the truck lock door and the east and

west airlock doors. Each entry point has two doors, an outer door and

an inner door. In the existing TS either door could be closed except

during personnel ingress or egress or while equipment is being passed

through the doorways. In the proposed TS the licensee has specified the

following. For the truck lock door the inner door to containment needs

to be closed. The reason given for the change is that the containment

boundary is more accurately defined as the interior access door between

the truck lock area and containment. The truck lock area was

transferred to the SNM-770 license in April 1970 and the outer doors

are controlled by this license.

For the east and west airlock doors, fire doors with an interior

crash bar have been installed at the outer door as a safety feature to

minimize the risk of personnel being trapped in containment during an

emergency. The airlock doors (inner doors) do not allow quick and

efficient egress during a postulated fire in containment; therefore,

the original air lock doors have been removed and confinement is

maintained by the newly installed fire doors.

Therefore the proposed TS require that the inner truck lock door be

closed and the outer east and west lock doors be closed except during

personnel ingress or egress or while equipment is being passed through

the doorways, and this meets the original goal of the existing TS.

The second TS change deals with the condition of the containment

when the containment is open for removal of materials and equipment. In

the existing TS Restricted Activities in containment are suspended. In

the proposed TS, containment extension is permitted if an enclosure is

provided around the opening to effectively isolate the containment from

the outside environment. If these extensions are not in place, all

Restricted Activities in containment are suspended. Negative pressure

(airflow into containment) is maintained in containment in the existing

as well as the proposed TS. Containment isolation is effectively

maintained under the proposed TS as it was in the existing TS.

The third TS change deals with the control of access into

containment. In the existing TS the outer doors in the air lock and the

truck lock outer doors shall be locked or blocked closed to prevent

unauthorized entry except when authorized personnel are inside the

containment building or outside with the door in view. In the proposed

TS access into containment is through a Health Physics (HP) control

point, which is on the first floor of the G-Building. To prevent

unauthorized entry the accesses into and out of containment shall be

locked or blocked closed except when this access control point is

supervised and the provisions of the first TS change are implemented.

Normal access to the containment is through a door in the G-

Building basement (east and west airlock doors). The G-Building

basement is a ``Radiation Area''. Routine activities during the day may

require workers to exit containment (rest, lunch, equipment change out,

etc). Locking or blocking the doors after workers temporarily exit

during the working day does not minimize radiation dose and reduces

worker efficiency. Access control will be established on the first

floor of the G-Building outside the radiation area. Therefore, the

access control point would provide positive control into and out of

containment and meets the original intent of the TS.

The fourth TS is being changed to include the HP control point in

the monthly visual surveillance, which assures that accesses into

containment are locked or blocked when no on is inside containment and

the HP control point is not occupied.

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

considerations. The proposed amendment to a license of a facility

involves no significant hazards consideration if operation of the

facility in accordance with the proposed amendment would not: (1)

Involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in the margin of safety.

The staff agrees with the licensee's no significant hazards

consideration determination submitted on September 15, 1999, for the

following reason:

The changes are consistent with the original intent of the TS,

i.e., to maintain confinement during Restricted Activities and to

prevent uncontrolled spread of contamination. Access control is still

being maintained.

Based on a review of the licensee's analysis, and on the staff's

analysis detailed above, 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.

Attorney for licensee: William David Wall, Assistant General

Counsel, CBS Corporation, 11 Stanwix Street, Pittsburgh, Pennsylvania

15222.

NRC Branch Chief: Ledyard B. Marsh.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: June 2, 1999, as supplemented August 25,

1999.

Description of amendment request: The proposed amendment would

relocate the quality assurance (QA) related requirements to the

licensee's Quality Assurance Program Description (QAPD) in accordance

with NRC Administrative Letter (AL) 95-06, ``Relocation of Technical

Specifications Administrative Controls Related to Quality Assurance,''

dated December 12, 1995. Specifically, Technical Specification (TS)

Section 6.5, ``Review and Audit,'' TS Section 6.8, ``Procedures and

Programs,'' and TS Section 6.10, ``Record Retention'' would be

relocated from the current TS to the QAPD in accordance with 10 CFR

50.36 (60 FR 30957).

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 analyzed?

Response: This amendment application does not involve a

significant increase in the probability or consequences of an

accident previously analyzed. The relocation of the administrative

controls from the Technical Specification to the Quality Assurance

Program Description (QAPD) does not alter the performance or

frequency of these activities. Any future changes to the QA Program

Description, which might constitute a reduction in commitments, are

governed by 10 CFR 50.54(a). Therefore, sufficient controls for

these requirements exist and these changes do not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

2. Does the proposed license amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

Response: This amendment application does not create the

possibility of a new or different kind of accident from any accident

previously evaluated. The proposed changes involve the relocation of

requirements from the Technical Specifications to the QAPD.

[[Page 59800]]

Relocation of these requirements does not affect plant equipment or

the way the plant operates. The functions continue to be performed

in the identical manner as they are currently being performed.

Therefore, the proposed revisions can not create a new or different

kind of accident.

3. Does the proposed license amendment involve a significant

reduction in a margin of safety?

Response: This amendment application does not involve a

significant reduction in a margin of safety. The requested Technical

Specification revisions relocate the administrative control

requirements from the Technical Specifications to the QAPD. These

requirements are not being altered by this relocation. The functions

continue to be performed in the identical manned as they are

currently being performed. Any future changes to the QA Program

Description, which might constitute a reduction in commitments, are

governed by 10 CFR 50.54(a). Therefore, sufficient controls for

these requirements exist and these changes do 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.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Section Chief: Sheri Peterson.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: July 30, 1999 (NRC-99-0048).

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to include provisions related

to enabling the oscillation power range monitor (OPRM) upscale trip

function in the average power range monitor. This change is associated

with the power range neutron monitoring (PRNM) system installed during

the last refueling outage. The associated Bases would also be revised.

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 is to enable the OPRM Upscale Function that

is contained in the previously installed PRNM equipment. Enabling

the OPRM hardware provides the long-term stability solution required

by Generic Letter 94-02. This hardware incorporates the Option III

detect and suppress solution reviewed and approved by the NRC in the

Reference 6, 7, and 8 [of the licensee's application dated July 30,

1999] Licensing Topical Reports and their Supplements. The OPRM is

designed to meet all requirements of GDC [General Design Criteria]

10 and 12 by automatically detecting and suppressing design basis

thermal-hydraulic power oscillations prior to violating the fuel

MCPR [minimum critical power ratio] Safety Limit. The OPRM system

provides this protection in the region where Interim Corrective

Actions (ICAs) restricted operation because of stability concerns.

Thus, the ICA restrictions on plant operation are deleted from the

TS, including region avoidance and the requirement for the operator

to manually scram the reactor with no recirculation loops operating.

Operation at high core powers with low core flows may cause a

slight, but not significant, increase in the probability that an

instability may occur. This slight increase is acceptable because

subsequent to the automatic detection of an instability, the OPRM

Upscale function provides an automatic scram signal to the RPS that

is faster than the operator-initiated manual scram required by the

current ICAs. Because of this rapid automatic action, the

consequences of an instability event are not increased as a result

of the installation of the OPRM system because it eliminates

dependence on operator actions.

Based on the above discussion, the proposed change does not

involve a significant increase in the probability or consequences of

an 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.

The proposed change permits Fermi 2 to enable the OPRM power

oscillation detect and suppress function provided in previously

installed PRNM hardware, and it simultaneously deletes certain

restrictions which preclude operation in regions of the power-flow

map where oscillations potentially may occur. Enabling the OPRM

Upscale function does not create any new system hardware interfaces

nor create any new system interactions. Potential failures of the

OPRM Upscale function result either in failure to perform a

mitigation action or in spurious initiation of a reactor scram.

These failures would not create the possibility of a new or

different kind of accident.

Based on the above discussion, the proposed change does not

create the possibility of a new or different kind of accident from

any accident previously evaluated.

3. The change does not involve a significant reduction in the

margin of safety.

The OPRM Upscale function implements BWROG [Boiling Water

Reactor Owners Group] Stability Option III, which was developed to

meet the requirements of GDC 10 and GDC 12 by providing a hardware

system that detects the presence of thermal-hydraulic instabilities

and automatically initiates the necessary actions to suppress the

oscillations prior to violating the MCPR Safety Limit. The NRC has

reviewed and accepted the Option III methodology described in the

Reference 6, 7, and 8 [of the licensee's application dated July 30,

1999] Licensing Topical Reports and their supplements, and concluded

that this solution will provide the intended protection. Therefore,

it is concluded that there will be no reduction in the margin of

safety as defined in the TS as a result of enabling the OPRM Upscale

function and simultaneously removing the operating restrictions

previously imposed by the ICAs.

Based on the above discussion, the proposed change does 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.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Section Chief: Claudia M. Craig.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: September 10, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirements (SRs)

3.8.4.1, 3.8.4.6, and 3.8.6.2 to accommodate changes in battery

parameters associated with the replacement of the Division I battery.

The licensee also plans to revise the Bases section for SR 3.8.6.2.

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 changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a change in the manner in

which the plant is operated. TS Sections [SRs] 3.8.4.1, 3.8.4.6,

3.8.6.2 and Bases Surveillance Requirement Section 3.8.6.2 are being

revised to reflect the new Division I battery cell/system

characteristics and associated requirements. The new battery will

have an increased capacity over the present battery, while

maintaining the existing battery system voltage requirements. This

is possible because the present and new battery specific gravity

(1.215) and type (lead calcium) are the same. Also, the end of

battery system discharge voltage remains the same as 210 VDC. The

Division I batteries will continue

[[Page 59801]]

to furnish power to redundant essential loads as required and as

designed. The new surveillance requirement voltages are based on the

same volts/cell criteria used for the existing batteries.

Furthermore, failure or malfunction of the station batteries does

not initiate any of the analyzed accidents previously evaluated in

the UFSAR [updated final safety analysis report]. The changes

described will therefore not involve an increase in the probability

or consequences of an accident previously evaluated.

2. The changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The new battery is Class 1E qualified equipment and is being

maintained within the same overall design parameters as the existing

battery. That is, the battery terminal voltage on float voltage

conditions (2.167 volt[s]/cell), overvoltage conditions (2.5 volts/

cell) and charger capability (2.15 volts/cell) are the same as the

original design. Furthermore, the end of system discharge voltage of

the battery system is maintained the same; therefore, there is no

negative impact to plant loads supplied by the batteries. Failures

of the batteries and chargers have been considered in both the

existing and modified configurations. The proposed changes will not

change performance or reliability nor introduce any new or different

failure modes or common mode failure and will therefore not create

the possibility of a new or different kind of accident from any

accident previously evaluated.

3. The changes do not involve a significant reduction in the margin

of safety.

The changes act to increase overall battery capacity from 560

ampere-hours to 1200 ampere-hours with the minimum battery discharge

voltage remaining at 210 VDC (or 105 VDC per battery). The battery

terminal voltage on float voltage conditions (2.167 volt[s]/cell),

overvoltage conditions (2.5 volts/cell) and charger capability (2.15

volts/cell) are the same as the original design. The new surveillance

requirement voltages are based on the same volts/cell criteria used for

the existing batteries. The batteries' ability to satisfy the design

requirements (battery duty cycle) of the dc system will not be reduced

from original plant design and will therefore not have any negative

impact to plant loads [that] the battery supplies. The proposed changes

therefore do not involve a 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.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Section Chief: Claudia M. Craig.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of amendment request: April 5, 1999; supplemented October 7,

1999.

Description of amendment request: The proposed amendments would

revise the Improved Technical Specifications (TS), Updated Final Safety

Analysis Report (UFSAR), and Core Operating Limits Report to

incorporate Topical Report (TR) DPC-NE-3005-P, ``Thermal-Hydraulic

Transient Analysis Methodology.'' The proposed changes are: (1)

Modification of a note for TS Surveillance Requirement (SR) 3.4.1.2,

``RCS [Reactor Coolant System] Pressure, Temperature, and Flow DNB

[Departure from Nucleate Boiling] Limits,'' to add that the SR would

apply for the condition where there is a 0 deg.F delta-Tcold setpoint;

(2) modification of TS 3.4.10, ``Pressurizer Safety Valves,'' to

increase the setpoint range of the lift settings for the pressurizer

safety valves; (3) modification of SR 3.4.10.1 to specify that the

pressurizer safety valve lift settings shall be within plus or minus 1

percent; (4) addition of TS 3.7.4, ``Atmospheric Dump Valve (ADV) Flow

Paths,'' to address the applicability and required actions related to

the ADS valves; (5) addition of TS 3.9.7, ``Unborated Water Source

Isolation Valves,'' to require valves that are used to isolate

unborated water sources to be secured in the closed position while in

Mode 6, provide required actions if one or more of the valves is not

secured in the closed position, and related SRs; (6) TS 5.6.5b would be

changed to update the Core Operating Limits Report references; and (7)

modification of the appropriate Bases to reflect the above changes and

consistentcy with the revision to the TR analysis. In addition,

proposed changes to the UFSAR revisions were provided.

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. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. The proposed changes to the Technical Specifications, Bases,

Updated Final Safety Analysis Report (UFSAR), and Core Operating

Limits Report (COLR) incorporate the accident analyses established

in Topical Report DPC-NE-3005-P, ``UFSAR Chapter 15 Transient

Analysis Methodology, Revision 1.'' On February 1, 1999, Duke

submitted Topical Report DPC-NE-3005-P to the NRC for approval. The

NRC found DPC-NE-3005-P acceptable as noted in SER [Safety

Evaluation Report] dated May 25, 1999.

The analyzed events are initiated by the failure of specific

plant structures, systems or components. These proposed changes do

not impact the condition or performance of those structures, systems

or components.

The revised accident analyses in DPC-NE-3005-P demonstrate that

the applicable acceptance criteria are met. In addition, the

calculations show that the applicable radiological and environmental

acceptance criteria will continue to be met.

Based on the above, the proposed changes do 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 accident previously evaluated?

No. The proposed changes do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

Where setpoints and operating limits have been revised, the revised

accident analyses demonstrate that the applicable acceptance

criteria are met. As a result, no new failure modes are being

introduced.

Based on the above, the proposed changes do not create the

possibility of any new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a margin of safety?

No. The margin of safety is established through the design of

the plant structures, systems and components, the parameters within

which the plant is operated, and the establishment of the setpoints

for the actuation of equipment relied upon to respond to an event.

The proposed changes do not involve a physical alteration of the

plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

Where setpoints and operating limits have been revised, the revised

accident analyses in DPC-NE-3005-P demonstrate that the applicable

acceptance criteria are met.

Based on the above, the proposed changes do 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.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

[[Page 59802]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 9, 1999.

Description of amendment request: The proposed amendment would

increase the authorized rated thermal power level of 3579 megawatts

thermal by 5 percent to 3758 megawatts thermal. The proposal follows

the NRC-approved generic format and content for Boiling Water Reactor

power uprate licensing topical reports documented in NEDC-31897P-A,

``Generic Guidelines for General Electric Boiling Water Reactor Power

Uprate,'' and NEDC-31984P, ``Generic Evaluations of General Electric

Boiling Water Reactor Power Uprate.''

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) Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The increase in power level discussed herein will not

significantly increase the probability or consequences of an

accident previously evaluated.

The probability (frequency of occurrence) of Design Basis

Accidents occurring is not affected by the increased power level, as

the regulatory criteria established for plant equipment (ASME code,

IEEE standards, NEMA standards, Regulatory Guide criteria, etc.) are

still complied with at the uprated power level. An evaluation of the

boiling water reactor (BWR) probabilistic risk assessments concludes

that the calculated core damage frequencies do not significantly

change due to power uprate. Scram setpoints (equipment settings that

initiate automatic plant shutdowns) are established such that there

is no significant increase in scram frequency due to uprate. No new

challenge to safety-related equipment results from power uprate.

The changes in consequences of hypothetical accidents which

would occur from 102% of the uprated power, compared to those

previously evaluated from greater than or equal to 102% of the

original power, are in all cases insignificant, because the accident

evaluations from power uprate compared with 105% of original power

do not result in exceeding the NRC-approved acceptance limits. The

spectrum of hypothetical accidents and transients has been

investigated, and shown to meet the plant's currently licensed

regulatory criteria. In the area of core design, for example, the

fuel operating limits such as Maximum Average Planar Linear Heat

Generation Rate (MAPLHGR) and Safety Limit Minimum Critical Power

Ratio (SLMCPR) are still met at the uprated power level, and fuel

reload analyses will show plant transients meet the criteria

accepted by the NRC as specified in NEDO-24011, ``GESTAR II.''

Challenges to fuel (ECCS performance) are evaluated, and shown to

still meet the criteria of 10 CFR 50.46 and Appendix K (Section 4.3

above, and Regulatory Guide 1.70 Safety Analysis Report Section

6.3).

Challenges to the containment have been evaluated, and the

containment and its associated cooling systems will continue to meet

10 CFR Appendix A Criterion 38, Long Term Cooling, and Criterion 50,

Containment.

Radiological release events (accidents) have been evaluated, and

shown to meet the guidelines of 10 CFR 100 (Regulatory Guide 1.70

Safety Analysis Report Chapter 15).

(2) Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

As summarized below, this change will not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

Equipment that could be affected by power uprate has been

evaluated. No new operating mode, safety-related equipment lineup,

accident scenario or equipment failure mode was identified. The full

spectrum of accident considerations defined in Regulatory Guide 1.70

has been evaluated and no new or different kind of accident has been

identified. Power uprate uses existing technology, and applies it

within the capabilities of already existing plant equipment in

accordance with existing regulatory criteria and includes NRC

approved codes, standards, and methods. General Electric has

designed BWRs of higher power and no new power dependent accidents

have been identified.

The technical specifications needed to implement power uprate

require some small adjustments, with no change to the plant's

physical configuration. All technical specification changes have

been evaluated and are acceptable.

(3) Will the change involve a significant reduction in a margin

of safety?

As summarized below, this change will not involve a significant

reduction in a margin of safety.

The calculated loads on all affected structures, systems and

components remain within their design allowables for all design

basis event categories. No NRC acceptance criteria are exceeded.

Some design and operational margins are affected by power uprate,

however, the margins of safety originally designed into the plant

are not affected by power uprate. Because the plant configuration

and reactions to transients and hypothetical accidents do not exceed

the presently approved NRC acceptance limits, power uprate 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.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 9, 1999.

Description of amendment request: The proposed amendment would

revise Perry Operating License Appendix B, the Perry Environmental

Protection Plan. The proposed change will eliminate the requirement in

the Environmental Protection Plan to sample Lake Erie sediment in the

Perry and Eastlake Plant area for Corbicula, since Corbicula and zebra

mussels have already been identified, and control and treatment plans

have been implemented which are effective on both species.

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 Perry Plant water source (Lake Erie) is now known to have

mussels and clams present. Therefore, it is no longer necessary to

use lake sampling techniques designed to provide advance notice of

their arrival. Treatment programs and monitoring for system fouling

are in place. The treatment programs and system monitoring for

fouling makes it highly likely that equipment degradation due to

Corbicula would be avoided or readily identified, allowing time for

corrective actions. Therefore, the programs will ensure that plant

systems remain capable of performing their intended functions. Since

the lake sampling was designed to allow time to implement a control

program, and the control program is now in place, elimination of the

lake sampling program will not involve a significant increase in the

probability or radiological consequences of an accident previously

evaluated.

(2) The proposed change would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed change will eliminate the lake sampling program

designed to detect the arrival of Corbicula, a particular species of

clam, at the Perry Plant. Since the clam is now known to exist in

the vicinity, and control methods are developed and implemented,

advanced detection is no longer required. Since the proposed change

involves only a monitoring program and does not change or modify the

design, maintenance or operation of any plant equipment, the

proposed change would not create the possibility of a new or

different

[[Page 59803]]

kind of accident from any accident previously evaluated.

(3) The proposed change will not involve a significant reduction

in the margin of safety.

The current requirements for aquatic monitoring are designed to

detect Corbicula prior to plant cooling water systems and heat

exchangers becoming infested with clams and flow becoming degraded,

and thus reducing the cooling available to safety systems.

Since an effective control method has already been implemented,

the deletion of a lake sampling method to provide advance warning of

clams in the area provides no significant benefit. The proposed

change will continue to provide the same level of protection against

system or component fouling that currently exists, thus the proposed

change 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. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street, Akron, OH 44308

NRC Section Chief: Anthony J. Mendiola.

First Energy Nuclear Operating Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 9, 1999.

Description of amendment request: The proposed amendment includes

nine separate changes to the Perry technical specifications. The

proposed changes include increasing the minimum water volume of the

condensate storage tank, clarification of minimum ECCS pump

differential pressures, clarifications to Required Action and Condition

statements, as well as minor nomenclature and editorial changes.

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 changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

A summary of the proposed changes is:

1. (Condensate Storage Tank (CST) Level-Low.) The Allowable

Values for the CST low water level limits (Technical Specification

(TS) Table 3.3.5.1-1 Function 3.d and Table 3.3.5.2-1 Function 3)

are being revised from greater than or equal to 59,700 gallons to

greater than or equal to 90,300 gallons based on recent revisions to

calculations taking into account potential vortex issues. This

change also results in raising the TS Surveillance Requirement (SR)

3.5.2.2.b value for the normal CST level limit to greater than or

equal to 249,700 gallons.

2. (Emergency Core Cooling System Pump Differential Pressure) TS

SRs 3.5.1.4 and SR 3.5.2.5 are being revised to better describe what

the differential pressures listed in the SRs represent at Perry

Nuclear Power Plant, in lieu of the phrase ``pump differential

pressure'.

3. (RCIC/RHR Steam Line Flow-High) The proposed change revises

the nomenclature on a table to match the plant-specific instrument

nomenclature.

4. (Containment Average Temperature-To-Relative-Humidity) This

revision is a clarification to prevent misinterpretation of the

Required Actions.

5. (Containment Vacuum Breakers) T 3.6.1.11 Required Action A.2

is being revised to clarify the proper actions to take if the

required number of vacuum breakers is not operable. Required Action

A.2 is being revised to add the word ``required'.

6. (Reporting Requirements) TS Administrative Controls Reporting

Requirement 5.6.1 is being revised to clarify the definition of the

time period of the report. ``Calendar'' is being removed from the

term ``calendar year'' to clarify the time period that the

Occupational Radiation Exposure Report is required to cover, to be

consistent with the revised wording in 10 CFR 20.1003.

7. (High Radiation Area) TS Administrative Control 5.7 is being

revised to update the titles of individuals responsible for

radiation protection. The term ``health physics'' is being revised

to ``radiation protection'' to be consistent with plant terminology.

8. (ECCS Instrumentation) Required Action E.1 Note 1 is being

revised for consistency with other specifications. The word ``in''

is being added.

9. (Electrical Power Systems) In TS 3.8.3, the word

``continued'' is being added to the bottom of the page for

consistency with other specifications.

The CST level change is adjusted in a conservative direction, as

recommended by NRC inspectors during a Safety System Functional

Inspection (SSFI) that was conducted in the spring of 1997. The

current setpoints were reviewed and determined to be adequate,

however it was suggested that some additional margin should be

added. The ``low level'' limits are being raised to move the

setpoint further away from the level at which vortexing would begin,

and the normal water level limit is also being raised to ensure that

at least 150,000 gallons of water would be available for HPCS and

RCIC. Since the existing limits are already considered adequate, and

the proposed changes are in the conservative direction, the proposed

change does not involve a significant increase in the probability or

radiological consequences of an accident previously evaluated.

The other eight proposed changes are administrative only, and

can have no effect on any previously evaluated accident scenario.

These eight changes have no effect on plant hardware, plant design,

safety limit settings, or system operation and therefore do not

modify or add any initiating parameters that would significantly

increase the probability of an accident previously evaluated, or the

radiological consequences of an event.

(2) The proposed changes would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes will raise the Condensate Storage Tank

level, which is conservative, and also includes some administrative

changes to improve clarity, update titles or terminology. None of

these changes can create the possibility of a new of different kind

of accident from any accident previously evaluated.

(3) The proposed changes will not involve a significant

reduction in the margin of safety.

The Condensate Storage Tank level change increases the margin of

safety by providing more margin between the setpoint that causes the

HPCS and RCIC suctions to shift from the CST to the Suppression Pool

and the beginning of the formation of a vortex at their pump

suctions. The other administrative changes have no effect on the

margin of safety. Therefore the proposed change 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. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 14, 1999.

Description of amendment request: The proposed amendment would

delete one Operating License Condition, and revise another. License

Condition 2.C.10 regarding controls over the containment air locks

during plant outages would be deleted due to the effective

implementation of Shutdown Safety administrative controls at Perry.

License Condition 2.F would be revised to clarify the intent of

reporting requirements for violations of the technical specifications

and the Environmental Protection Plan.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the

[[Page 59804]]

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 delete or revise two Operating License

Conditions, one that addresses administrative controls on air locks

during refueling outages, and one regarding reporting of violations

of the technical specifications and the Environmental Protection

Plan.

These proposed changes to the Operating License are

administrative only, and have no effect on any previously evaluated

accident scenario. The proposed changes have no effect on plant

hardware, plant design, safety limit setting, or plant system

operation and therefore do not modify or add any initiating

parameters that would significantly increase the probability of an

accident previously evaluated.

The changes will not alter the operation of equipment assumed to

be available for the mitigation of accidents or transients, nor will

they alter the operation of equipment important to safety previously

evaluated in the accident analyses.

The proposed activity does not affect accident mitigation

capabilities or the radiation release amounts for postulated

accidents. Since there are no changes to previous accident analyses,

the radiological consequences associated with these analyses remain

unchanged.

Therefore, the proposed change does not significantly increase

the probability or consequences of an accident previously evaluated.

(2) The proposed change would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, and do not

involve any physical alteration of the plant (no new or different

type of equipment will be installed). They do not alter the design

assumptions, conditions, configuration of the facility or the manner

in which the plant is operated. The proposed changes have no impact

on component and system interactions.

The safety functions of plant structures, systems, and

components are also not changed in any manner, nor is the

reliability of any structure, system, or component reduced.

The proposed changes are not providing for operation in a mode

that is not already evaluated. These changes do not affect the

operation of any systems or components, nor do they involve any

potential initiating events that would create any new or different

kind of event.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(3) The proposed change will not involve a significant reduction

in the margin of safety.

The proposed changes are administrative in nature (they delete

or revise two license conditions). Administrative controls will

continue to be applied to the opening of the air locks during plant

shutdown periods, and to the reporting of violations of the

technical specifications and the Environmental Protection Plan.

There is no impact on safety limits or limiting safety system

settings. The changes do not affect any plant safety parameters or

setpoints. No physical or operational changes to the facility will

result from the proposed changes.

The proposed changes have no impact on any safety analysis

assumptions. Consequently, no margin of safety as described in the

Final Safety Analysis Report or defined in the basis of any

technical specification is reduced as a result of these changes.

These proposed changes do not detrimentally affect the ability of

structures, systems, and components important to safety to fulfill

their intended safety functions.

Therefore, the proposed changes do not cause 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.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County,

Michigan

Date of amendment requests: October 12, 1999.

Description of amendment requests: The proposed amendments would

revise Technical Specification (T/S) Surveillance Requirement (SR)

4.6.2.2.d for the spray additive system to relocate the details

associated with the acceptance criteria and test parameters to the

associated T/S Bases. Additionally, certain administrative text format

changes are being proposed.

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 change involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated?

The proposed changes relocate the details associated with the

acceptance criteria and test parameters from the T/S SR to the

associated Bases and do not affect system operability or

performance. The format changes in the text on each page are

administrative in nature and do not result in any change in plant

operation. Relocation of this information to the Bases is

administrative in nature and does not affect the probability or

consequences of any accident previously evaluated. No actual change

to the requirement is made. Actual plant operation is not affected

by the administrative changes. No methods of operation of plant

systems, structures or components are changed. Operation of accident

mitigation features is not changed. Consequently, there is no affect

upon the probability of any previously analyzed accident, transient,

accident initiators, or precursor events. Additionally, because

there is no actual change in plant design or operation, there is no

affect upon radioactive material inventories, plant shielding, or

effluent release points. Therefore, these changes do not

significantly increase the probability of occurrence or consequences

of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes relocate the details associated with the

acceptance criteria and test parameters from the

T/S SR to the associated Bases and do not affect system operability

or performance. The format changes in the text on each page are

administrative in nature and do not result in any change in plant

operation. Facility operation and procedures are not changed.

Relocation of this information to the Bases is administrative in

nature and does not affect [sic] create any new accident scenarios,

accident initiators, or precursor events. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes relocate the details associated with

acceptance criteria and test parameters from the T/S SR to the

associated Bases and do not modify T/S safety settings, setpoints,

or other values. The format changes in the text on each page are

administrative in nature and do not result in any change in plant

operation. There is no effect upon operating margins and accident

margins because the administrative changes do not change the manner

of operation of plant systems, structures, or components. Plant

emergency and abnormal operating procedures are not affected. There

is no change of actual testing methodology, test parameters, or

acceptance criteria. The response of the plant to an event is the

same. Potential offsite doses are unaffected because operation of

the facility is unchanged. Relocation of the testing details to the

Bases is acceptable because controls are in place for T/S Bases

changes which require evaluation of changes under the provisions of

10 CFR 50.59. Therefore, the proposed changes do 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

[[Page 59805]]

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: September 30, 1999.

Description of amendment request: The proposed amendment would

change the Technical Specification surveillance periodicity

requirements for the control room emergency filtration system.

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 amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

During an accident, the Control Room Emergency Filtration [EFT]

System provides filtered air to pressurize the Control Room to

minimize the activity, and therefore the radiological dose, inside

the Control Room. Technical Specification surveillance requirements

are established in order to ensure that the EFT System will perform

its safety function during an accident. The proposed amendment

eliminates unnecessary testing which is not required to show that

the filters are operable and which causes unnecessary wear and tear

on the system. The remaining surveillances adequately show that the

system is operable and capable of performing its safety function.

Dose to the public and the Control Room operators are not affected

by the proposed change.

The proposed Technical Specification change does not introduce

new equipment operating modes, nor does the proposed change alter

existing system relationships. The proposed amendment does not

introduce new failure modes.

Therefore, the proposed amendment will not significantly

increase the probability or the consequences of an accident

previously evaluated.

2. The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

The proposed Technical Specification change does not introduce

new equipment operating modes, nor does the proposed change alter

existing system relationships. The proposed amendment does not

introduce new failure modes. The proposed surveillance requirements

are consistent with industry and regulatory guidance and show that

the system is capable of performing its safety function. System

reliability is enhanced by the proposed change by eliminating

unnecessary wear on the system.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed amendment will not involve a significant reduction

in the margin of safety.

The proposed amendment is within current industry and regulatory

standards for testing filters. The proposed amendment maintains

margins of safety. Off-site and Control Room dose assessments are

not affected by the proposed amendment, since the ability of the EFT

System to perform its safety function is shown by the proposed

surveillance requirements. The proposed change to the surveillance

provides assurance that the system will perform at the filter

efficiency used in the evaluation of the radiological consequences

of the postulated events. Therefore, the proposed amendment 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. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Section Chief: Claudia M. Craig.

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: September 30, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications associated with the Safety Limit

Minimum Critical Power Ratios (SLMCPRs) in order to support the

operation of Hope Creek in the upcoming Cycle 10 with a mixed core of

General Electric (GE) and Asea Brown Bovieri/Combustion Engineering

(ABB/CE) fuel. In addition, administrative changes would be made to the

Technical Specifications to reflect the change in fuel vendor from GE

to ABB/CE.

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 changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The derivation of the revised SLMCPRs for Hope Creek for

incorporation into the Technical Specifications, and its use to

determine cycle-specific thermal limits, have been performed using

NRC [U.S. Nuclear Regulatory Commission] approved methods. These

calculations do not change the method of operating the plant and

have no effect on the probability of an accident initiating event or

transient.

There are no significant increases in the consequences of an

accident previously evaluated. The basis of the MCPR Safety Limit is

to ensure that no mechanistic fuel damage due to clad overheating is

calculated to occur if the limit is not violated. The new SLMCPRs

preserve the existing margin to transition boiling and the

probability of fuel damage is not increased.

Removal of the cycle specific footnote for the Safety Limit

applicability will not involve a significant increase in the

probability or consequences of an accident previously evaluated

since the change is administrative and does not affect the plant or

fuel design or operation.

Likewise, the proposed changes to the Average Planar Heat

Generation Rate (APLHGR), Minimum Critical Power Ratio (MCPR),

Recirculation Loop Limiting Condition for Operation (LCO) Action

Statements, and references to fuel vendor analyses and reports do

not involve a significant increase in the probability or

consequences of an accident previously evaluated. The changes to the

APLHGR, MCPR and Recirculation Loop LCOs are considered to be

administrative in nature since the Core Operating Limits Report

(COLR) will continue to be used to appropriately control and limit

the bounds of plant operation with slow control rods or during

single recirculation loop operation, and the COLR will still be

developed in accordance with NRC approved methods. Similarly, the

revised references to the fuel vendor throughout the Technical

Specifications are also considered to be administrative in nature

since they reflect the current status of NRC approval of

methodologies utilized by PSE&G [Public Service Electric and Gas

Company] and the fuel vendor to develop operating and safety limits

for the fuel and core designs. These proposed changes do not alter

the method of operating the plant and have no effect on the

probability of an accident initiating event or transient.

There are no significant increases in the consequences of an

accident previously evaluated. The basis of the COLR and the PSE&G

and fuel vendor methodologies is to ensure that no mechanistic fuel

damage is calculated to occur if the limits on plant operation are

not violated. The COLR will continue to preserve the existing margin

to fuel damage and the probability of fuel damage is not increased.

Therefore, the proposed change does not involve an increase in

the probability or consequences of an 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.

[[Page 59806]]

The proposed changes contained in this submittal result from an

analysis of the reload core using the same fuel types as previous

cycles and an ABB/CE fuel design with extensive operating

experience. These changes do not involve any new method for

operating the facility and do not involve any facility modifications

for the reload core operation. No new initiating events or

transients result from these changes. Therefore, the proposed

Technical Specification changes do not create the possibility of a

new or different kind of accident, from any accident previously

evaluated.

Removal of the cycle specific footnote for the Safety Limit

applicability does not create the possibility of a new or different

kind of accident from any accident previously evaluated since the

change is administrative and does not affect the plant or fuel

design or operation.

The changes to the APLHGR, MCPR and Recirculation Loop LCOs are

considered to be administrative in nature since the Core Operating

Limits Report (COLR) will continue to be used to appropriately

control and limit the bounds of plant operation with slow control

rods or during single recirculation loop operation, and the COLR

will still be developed in accordance with NRC approved methods.

These changes do not involve any new method for operating the

facility and do not involve any facility modifications in addition

to the new fuel design. No new initiating events or transients

result from these changes. Therefore, the proposed Technical

Specification changes do not create the possibility of a new or

different kind of accident.

The revised references to the fuel vendor throughout the

Technical Specifications are also considered to be administrative in

nature since they reflect the current status of NRC approval of

methodologies utilized by PSE&G and the fuel vendor to develop

operating and safety limits for the fuel and core designs. These

changes do not involve any new method for operating the facility and

do not involve any facility modifications in addition to the new

fuel design. No new initiating events or transients result from

these changes. Therefore, the proposed Technical Specification

changes do 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.

The margin of safety as defined in the Technical Specification

bases will remain the same. The new SLMCPRs are calculated using NRC

approved methods, which are in accordance with the current fuel

designs, and licensing criteria. The MCPR Safety Limit remains high

enough to ensure that greater than 99.9% of all fuel rods in the

core will avoid transition boiling if the limit is not violated,

thereby preserving the fuel cladding integrity. Therefore, the

proposed Technical Specification changes do not involve a

significant reduction in a margin of safety.

Removal of the cycle specific footnote for the Safety Limit

applicability does not create the possibility of a new or different

kind of accident from any accident previously evaluated since the

SLMCPR will continue to be evaluated on a cycle-specific basis.

The margin of safety as defined in the Technical Specification

bases will likewise remain unaffected by the proposed changes to

APLHGR, MCPR and Recirculation Loop LCOs, and the revised references

to the fuel vendor throughout the Technical Specifications. These

changes establish controls for plant operation and establish bases

for fuel analyses that reflect NRC approved methods, and are in

accordance with the current fuel design and licensing criteria.

These changes will continue to ensure that the plant is operated

within specified acceptable fuel design limits. Therefore, the

proposed Technical Specification changes do 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.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: September 8, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.8.1, ``A.C. Sources,

Operating,'' and associated Bases, by eliminating the requirement for

accelerated testing of the standby diesel generators and the associated

reporting requirements. The TS Index would also be revised to reflect

these changes.

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 change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes do not involve hardware changes nor do they

affect the operational limits or design of the standby diesel

generators or power systems. These changes do not alter assumptions

made in the safety analysis. In conjunction with the maintenance

rule program, these changes continue to assure the operability and

reliability of the standby diesel generators while minimizing the

number of required engine starts and associated wear. These changes

are also consistent with the guidance provided in Generic Letter 94-

01, ``Removal of Accelerated Testing and Special Reporting

Requirements for Emergency Diesel Generators.''

Therefore, the proposed changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes minimize the number of required standby

diesel generator starts; they do not affect the operational limits

or design. The performance capability of the standby diesel

generators is not affected. These changes do not alter the plant

configuration (no new or different type of equipment will be

installed) or make changes in methods governing normal plant

operation. These changes do not alter assumptions made in the safety

analysis. These changes are also consistent with the guidance

provided in Generic Letter 94-01.

Therefore, the changes will not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes do not involve a change in the operational

limits or design of the emergency power system. The design and

capabilities of the standby diesel generators are not affected by

these changes. These changes are also consistent with the guidance

provided in Generic Letter 94-01.

The proposed changes do not involve a significant reduction in

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

request for amendments involves no significant hazards consideration.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Robert A. Gramm.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: September 8, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.8.1, ``A.C. Sources, Operating,''

and associated Bases, by relocating the 18-month surveillance to

subject the standby diesel generator to inspections in accordance with

procedures prepared in conjunction with its manufacturer's

recommendations, to the Updated Final Safety Analysis Report.

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

[[Page 59807]]

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the Updated Final

Safety Analysis Report (UFSAR). The change does not result in any

hardware or operating procedure changes. The requirement being

removed from the Technical Specifications is not the initiator of

any analyzed event. The UFSAR is maintained using the provisions of

10 CFR 50.59. Since any changes will be evaluated per 10 CFR 50.59,

no significant increase in the probability or consequences of an

accident previously evaluated will be allowed without prior NRC

approval. Therefore, the changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the Updated Final

Safety Analysis Report (UFSAR). The change does not alter the plant

configuration (no new or different type of equipment will be

installed) or make changes in methods governing normal plant

operation. The change does not impose different requirements. The

change does not alter assumptions made in the safety analysis and

licensing basis. Therefore, the change will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. Does this change involve a significant reduction in a margin

of safety?

The proposed change moves the requirement to perform

manufacturer's recommended inspections of the Standby Diesel

Generators from the Technical Specifications to the Updated Final

Safety Analysis Report (UFSAR). The change does not reduce the

margin of safety since the location of details has no impact on any

safety analysis assumptions. In addition, the requirement being

transposed from the Technical Specification to the UFSAR [is the]

same as the existing Technical Specification. Also, the UFSAR is

maintained using the provisions of 10 CFR 50.59. Since any changes

will be evaluated per 10 CFR 50.59, no significant reduction in a

margin of safety will be allowed without prior NRC approval.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

request for amendments involves no significant hazards consideration.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority (TVA), Docket Nos. 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 2 and 3, Limestone County,

Alabama

Date of amendment request: September 28, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to increase the maximum allowable

leakage rates for main steam isolation valves.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

TVA proposes to utilize the main steam drain lines to

preferentially direct MSIV leakage to the main condenser. This drain

path takes advantage of the large volume of the steam lines and

condenser to provide holdup and plate-out of fission products that

may leak through the closed MSIVs. In this approach, the main steam

lines, steam drain piping, and the main condenser are used to

mitigate the consequences of an accident to limit potential off-site

exposures below those specified in 10 CFR 100 and 10 CFR 50 Appendix

A, GDC 19 for control room dose limits.

Seismic verification walkdowns and evaluations of representative

piping/supports were performed to demonstrate the main steam line

piping and components that comprise the ALT path were rugged, and

able to perform the safety function of MSIV leakage control

following an Design Basis Earthquake (DBE). Thus, it has been

concluded the primary components in the MSIV alternate treatment

flow path can be relied upon to maintain structural integrity.

Therefore, the proposed amendment does not involve changes to

structures, components, or systems which would affect the

probability of an accident previously evaluated in the Browns Ferry

Final Safety Analysis Report (FSAR).

A plant-specific radiological analysis has been performed to

assess the effects of the proposed increase in MSIV leakage criteria

in terms of off-site doses and main control room dose. This analysis

uses the holdup and plate-out factors described in NEDC-31858P,

Revision 2. The analysis shows the dose contribution from the

proposed increase in leakage criteria is acceptable compared to

doses limits prescribed in 10 CFR 100 and 10 CFR 50, Appendix A, GDC

19. Therefore, the proposed changes do not significantly increase

the consequences of an accident previously evaluated.

B. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes require the use of the main steam piping

and the condenser to process MSIV leakage. This additional function

does not compromise the reliability of these systems. They will

continue to function as intended and not be subject to a failure of

a different kind than previously considered. In addition, MSIV

functionality will not be adversely impacted by the increased

leakage limit. Therefore, the proposed change does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change to TS Surveillance Requirement 3.6.1.3.10 to

increase the allowable MSIV leakage does not involve a significant

reduction in the margin of safety. The allowable leak rate specified

for the MSIVs is used to quantify a maximum amount of leakage

assumed to bypass containment. The results of the re-analysis

supporting these changes were evaluated against the dose limits

contained in 10 CFR 100 for off-site doses and 10 CFR 50, Appendix

A, GDC 19 for control room doses. Sufficient margin relative to the

regulatory limits is maintained even when conservative assumptions

and methods are utilized. 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Section Chief: Sheri R. Peterson.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone

County, Alabama

Date of amendment request: September 30, 1999.

Description of amendment request: The proposed amendments consist

of administrative revisions to the Operating Licenses for BFN Units 1,

2 and 3 that delete license conditions that have become outdated, are

no longer applicable, or are redundant, and consolidate license

conditions which currently exist in two locations in each units'

Technical Specifications.

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

[[Page 59808]]

consideration, which is presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The changes requested by this submittal are administrative in

nature and do not change the way BFN operates. The proposed changes

are intended to: delete redundant paragraphs, delete requirements

and authorizations for modifications that have been completed,

delete an authorization to temporarily store radioactive material on

site, delete an exemption from a General Design Criterion which has

expired, and consolidate license conditions which currently exist in

two locations in each units Technical Specifications.

The change does not affect any design bases accident or the

ability of any safe shutdown equipment to perform its design

function. There are no physical modifications that are required to

implement this license condition update. There is no impact on plant

equipment or changes to operating procedures. Therefore, the

proposed amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

B. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The changes described above are administrative in nature and do

not change the way BFN operates. There are no physical modifications

authorized by the proposed changes and there are no procedure or

process changes that are requested. Changes requested are intended

to ensure the license conditions reflect the current status of the

plant. There is no impact on any accident analysis created by this

change. Therefore, the proposed amendment does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The changes described above are administrative in nature and do

not change the way BFN operates. There are no procedural or physical

changes required by this amendment. The license conditions are being

updated partially as a result of NRC Information Notice 97-43 which

highlighted the importance of periodically verifying compliance with

the Operating License. These changes are intended to delete license

conditions which are no longer needed or are redundant in order to

ensure the license conditions accurately reflect the current status

of the licensed facility. The change does not affect any design

bases accident or the ability of any safe shutdown equipment to

perform its design function, therefore no margins of safety have

been affected by any of these changes. Accordingly, the proposed

amendment 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.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Acting Section Chief: Ronald W. Hernan.

Previously Published Notices of Consideration of Issuance of

Amendment 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 of the same as above. They were

published as individual notices either because the 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 considerations.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: June 8, 1999.

Brief description of amendments request: The proposed amendments

would revise Technical Specification (TS) 3.7.15, ``Fuel Storage Pool

Boron Concentration,'' TS 3.7.17, ``Spent Fuel Assembly Storage,'' and

TS 4.3.1, ``Criticality,'' to increase spent fuel pool storage capacity

by crediting soluble boron and decay time in the safety analysis for

the spent fuel pool storage racks. The proposed amendments would also

increase the maximum radially averaged fuel enrichment from 4.3 weight

percent to 4.8 weight percent.

Date of publication of individual notice in Federal Register:

September 20, 1999 (64 FR 50835)

Expiration date of individual notice: October 20, 1999.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: October 8, 1999.

Brief description of amendments request: The proposed amendment

would revise Technical Specification (TS) Section 3.8.4, ``DC Sources--

Operating,'' to waive, on a one-time basis, the requirement to perform

Surveillance Requirement (SR) 3.8.4.8 for Unit 1 channels A, B, and C.

Date of publication of individual notice in Federal Register:

October 19, 1999 (64 FR 56369).

Expiration date of individual notice: For comments on proposed no

significant hazards consideration determination: November 2, 1999; for

opportunity for hearing: November 18, 1999.

Southern California Edison Company, et al., Docket Nos. 50-361 and

50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San

Diego County, California

Date of amendment request: October 20, 1998 (PCN 485), as

supplemented August 13, 1999.

Brief description of amendment request: The proposed amendments

would revise the San Onofre Nuclear Generating Station Units 2 and 3

technical specifications Surveillance Requirement 3.3.9 to include a

response time testing requirement for the control room isolation

signal.

Date of publication of individual notice in Federal Register:

October 12, 1999 (64 FR 55311.

Expiration date of individual notice: November 12, 1999.

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,

[[Page 59809]]

and Opportunity for A Hearing in 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 electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: October 27, 1998.

Brief description of amendment: The amendments update the Operating

Licenses for the Brunswick Steam Electric Plant, Units 1 and 2.

Date of issuance: October 5, 1999.

Effective date: October 5, 1999.

Amendment No.: 206 and 236.

Facility Operating License Nos. DPR-71 and DPR-62: Amendment

revises the Operating Licenses.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71964).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 5, 1999.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of application for amendment: June 2, 1999, as supplemented on

September 1, 1999.

Brief description of amendment: This amendment relocates Technical

Specification (TS) Section 6.5, ``REVIEW AND AUDIT,'' TS 6.8.2, TS

6.8.3, and TS Section 6.10, ``RECORD RETENTION,'' intact from the

Harris Nuclear Plant (HNP) TS to the Quality Assurance Program

Description (QAPD) currently located in HNP Final Safety Analysis

Report Section 17.3. Future changes to the associated relocated TS will

be processed in accordance with 10 CFR 50.54(a). The change is

consistent with NUREG-1431, Revision 1, ``Standard Technical

Specifications, Westinghouse Plants,'' dated April 1995, and with the

guidance provided in NRC Administrative Letter 95-06, ``Relocation of

Technical Specification Administrative Controls related To Quality

Assurance,'' dated December 12, 1995.

Date of issuance: October 19, 1999.

Effective date: October 19, 1999.

Amendment No.: 92.

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 30, 1999 (64 FR

35201).

The September 1, 1999, submittal contained clarifying information

only, 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 19, 1999.

No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket

Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of application for amendments: June 30, 1999.

Brief description of amendments: The amendments revised the

requirements related to the cross-tie of DC power buses between units,

remove references to the AT&T batteries which have been replaced at

Braidwood Station, and remove references to the 10-day allowed outage

time (AOT) required for replacement of the AT&T batteries at Braidwood,

Unit 2, which was granted in Amendment Nos. 99 and 99 issued to

Braidwood Station, Unit Nos. 1 and 2, on March 26, 1999.

Date of issuance: October 13, 1999.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 111 and 104.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43767).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 13, 1999.

No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: November 9, 1998, and July 7,

1999.

Brief description of amendments: The amendments revised Technical

Specification Table 3.3.3-2, ``Emergency Core Cooling System Actuation

Instrumentation Setpoints,'' to modify the degraded voltage second

level undervoltage relay setpoint and allowable value.

Date of issuance: October 15, 1999.

Effective date: Immediately, to be implemented prior to startup

from L1R08 for Unit 1 and prior to startup from L2R08 for Unit 2.

Amendment Nos.: 135 and 120.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 13, 1999 (64 FR

2245) and August 11, 1999 (64 FR 43769).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 15, 1999.

No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: August 13, 1999, as

supplemented on August 27, 1999.

Brief description of amendments: The amendments revise Technical

Specification (TS) Section 1.0, ``Definitions,'' Item 1.7, ``Core

Alteration,'' to specify that instrumentation and control rod movements

are not considered core alterations if there are no fuel assemblies in

the associated cell. The amendments also revise TS Sections 3/4.1, 3/

4.3, and 3/4.9 to reflect the change in definition. In addition, a

license condition is added as follows: ``The licensee is prohibited

from moving any fuel assemblies within the reactor pressure vessel

unless all control rods except one are fully inserted during refueling

in Mode 5''.

Date of issuance: October 18, 1999.

[[Page 59810]]

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 136 and 121.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Operating Licenses and Technical Specifications.

Date of initial notice in Federal Register: September 8, 1999 (64

FR 48860).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 18, 1999.

No significant hazards consideration comments received: No.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of application of amendment: June 3, 1999, and as supplemented

by letter dated August 24, 1999 .

Brief description of amendment: The amendment revises the Operating

License to clarify that the license is not terminated until the

Commission notifies the licensee in writing, and relocates certain

Technical Specification (TS) requirements to licensee-controlled

documents. The administrative controls section of the TSs have been

revised to more closely conform to the standardized TSs. Administrative

controls have been added for the control of radioactive effluents. A TS

Bases Control Program has been added. The weight limit for loads

carried over the spent fuel pool (SFP) has been increased. The

amendment deletes certain TSs that are either (1) no longer applicable

to the permanently shutdown and defueled state of the reactor, or (2)

which duplicate regulatory requirements, or (3) which duplicate

information located in the Updated Final Safety Analysis Report. A

number of editorial changes were made to clarify the language used, to

correct typographical errors, to renumber the listings, to remove

section numbers that no longer contain requirements, and to renumber

the pages in the TSs.

Date of issuance: October 19, 1999.

Effective date: As of the date of issuance and shall be implemented

within 60 days of issuance.

Amendment No.: 195.

Facility Operating License No. DPR-61: The amendment revised the

Operating License and the Technical Specifications.

Date of original notice in Federal Register: July 14, 1999 (64 FR

38024).

The August 24, 1999, supplement contained clarifications of the

June 3, 1999 amendment request. The supplemental information did not

change the staff's initial proposed no significant hazards

consideration determination nor expand the scope of the original

notice. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 19, 1999.

No significant hazards consideration received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: September 24, 1999.

Description of amendment request: The amendment revises current

Technical Specification (TS) 3.6.1.8 by adding footnote ``**'' to

Action b. The footnote allows continued operation of Fermi 2 with the

leakage of penetration X-26 exceeding the limit in TS 4.6.1.8.2,

provided certain compensatory measures are taken. Operation is allowed

to continue until the next plant shutdown.

Because the NRC staff issued the Fermi 2 improved standard TSs

(ITS) on September 30, 1999, with implementation within 90 days, this

amendment also provides pages that are compatible with the ITS. The

amendment adds a new special operations TS, ITS 3.10.8, to address the

compensatory actions and other requirements associated with penetration

X-26.

Date of issuance: October 19, 1999.

Effective date: October 19, 1999, and shall be implemented within 5

days.

Amendment No.: 135.

Facility Operating License No. NPF-43: Amendment revises the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration (NSHC): Yes (64 FR 53421, dated October 1, 1999). 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 1,

1999, 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 NSHC determination are contained in a

Safety Evaluation dated October 19, 1999.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Section Chief: Claudia M. Craig.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi,

Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,

Claiborne County, Mississippi

Date of application for amendment: June 23, 1999, as supplemented

by letters dated August 6, September 8, and October 4, 1999.

Brief description of amendment: The amendment revises Technical

Specification requirements for handling irradiated fuel in the

Containment Building and in the Auxiliary Building, and selected

specifications associated with performing core alterations.

Date of issuance: October 20, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment No: 139.

Facility Operating License No. NPF-29: The amendment revises the

Technical Specifications and Operating License.

Date of initial notice in Federal Register: August 25, 1999 (64 FR

46435).

The August 6, September 8, and October 4, 1999, submittals provided

additional clarifying information and did not change the initial

proposed no significant hazards consideration determination and did not

expand the scope of the original application.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 20, 1999.

No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant, Units 3 and 4, Dade County, Florida

Date of application for amendments: March 8, 1999.

Brief description of amendments: The amendments revised the

Technical Specifications (TS), Section 6.0, Administrative Controls, by

removing requirements that are adequately controlled by existing

regulations other than 10 CFR 50.36 and the TS. The amendments also

relocate selected requirements from TS 6.0 to licensee-controlled

documents or programs (e.g., the final safety analysis report or the

quality assurance plan). Guidance on the changes was developed by the

NRC and provided in the Standard Technical Specifications for

Pressurized Water Reactor Plants, NUREG-1431, and Administrative Letter

95-06, ``Relocation of Technical Specification

[[Page 59811]]

Administrative Controls Related to Quality Assurance,'' issued on

December 12, 1995.

Date of issuance: October 6, 1999.

Effective date: As of date of issue, to be implemented within 90

days of issuance.

Amendment Nos.: 201 and 195.

Facility Operating License Nos. DPR-31 and DPR-41: Amendments

revised the TS.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17025).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 6, 1999.

No significant hazards consideration comments received: No.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of application for amendment: May 10, 1999, as supplemented

July 16 and October 4, 1999.

Brief description of amendment: The amendment revised Duane Arnold

Energy Center (DAEC) Technical Specification (TS) 2.1.1.2 to revise the

Safety Limit Minimum Critical Power Ratio (SLMCPR) to support operation

with GE-12 fuel with a 10x10 pin array.

Date of issuance: October 20, 1999.

Effective date: Immediately, to be implemented within 30 days

Amendment No.: 229.

Facility Operating License No. DPR-49: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38029).

The July 16 and October 4, 1999, letters provided additional

clarifying information within the scope of the original Federal

Register notice and did not affect the NRC staff's initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 20, 1999.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County,

Michigan

Date of application for amendments: September 14, 1998.

Brief description of amendments: The amendments revise Technical

Specification page 3/4 5-6, ``Limiting Conditions for Operation and

Surveillance Requirements--Emergency Core Cooling Systems (ECCS),'' and

its associated Bases to change pump runout limits for a safety

injection pump to 675 gallons per minute (gpm) unless the pump is

specifically tested to a higher flow rate not to exceed 700 gpm for

Units 1 and 2.

Date of issuance: October 21, 1999.

Effective date: October 21, 1999, with full implementation within

45 days.

Amendment Nos.: 229 and 212.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 1999 (64 FR

47533).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 21, 1999.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County,

Michigan

Date of application for amendments: May 21, 1999.

Brief description of amendments: The amendments change the

Technical Specifications (TS) to allow reactor coolant system

temperature changes in certain Mode 5 and 6 action statements if the

shutdown margin is sufficient to accommodate the expected temperature

change. In addition, footnotes regarding additions of water from the

refueling water storage tank to the reactor coolant system are

clarified and relocated to action statements. Additional actions are

added in Table 3.3-1, ``Reactor Trip System Instrumentation,'' when the

required source range neutron flux channel is inoperable. Corresponding

changes are proposed for the Bases for TS 3/4.1.1, ``Boration

Control,'' and TS 3/4.1.2, ``Boration Systems.'' Administrative changes

are proposed to improve clarity. Finally, additions are made to

shutdown margin TS surveillance requirements to address use of a boron

penalty (requirement for additional boron) during residual heat removal

system operation in Modes 4 and 5.

Date of issuance: October 21, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment Nos.: 230 and 213.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 12, 1999 (64 FR

37574).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated October 21, 1999.

No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: December 31, 1998, as

supplemented May 17, 1999.

Brief description of amendment: The amendment revises the technical

specification reactor pressure vessel (RPV) pressure-temperature limit

curves, deletes completed RPV sample surveillance requirements, deletes

the requirement to withdraw a specimen at the next refueling outage,

removes the standby liquid control system relief valve setpoint, and

makes associated administrative changes.

Date of issuance: October 12, 1999.

Effective date: October 12, 1999, with full implementation within

45 days.

Amendment No.: 106.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6706). The May 17, 1999, submittal added clarifying information that

was within the scope of the original Federal Register notice and did

not change the staff's initial proposed no significant hazards

considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 12, 1999.

No significant hazards consideration comments received: No.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: April 9, 1999.

Brief description of amendment: The amendment changes the Technical

Specifications by increasing the allowable outage time for any one

safety injection pump.

Date of issuance: October 12, 1999.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 196.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 2, 1999 (64 FR

297147).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 12, 1999.

No significant hazards consideration comments received: No.

[[Page 59812]]

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: January 29, 1999, as

supplemented August 2, 1999.

Brief description of amendment: The amendment changes the Technical

Specifications by increasing the allowable control rod misalignment

when operating at or below 85% power.

Date of issuance: October 14, 1999.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 197.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: April 21, 1999 (64 FR

19564).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 14, 1999.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: August 19, 1999, as supplemented

by letter dated October 8, 1999.

Brief description of amendment: The amendment revises the TS to

incorporate the new Pressure/Temperature Limits Curves consistent with

the analysis results of reactor specimen W.

Date of issuance: October 21, 1999.

Effective date: October 21, 1999.

Amendment No.: 143.

Facility Operating License No. NPF-12: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 8, 1999 (64

FR 48865). The October 8, 1999, submittal contained clarifying

information only, 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 21, 1999.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 2, 1999 (TS 98-05).

Brief description of amendments: The amendments delete the Sequoyah

Nuclear Plant. License Conditions that require an Independent Safety

Engineering Group.

Date of issuance: October 12, 1999.

Effective date: As of the date of issuance to be implemented no

later than 45 days after issuance.

Amendment Nos.: 248 and 239.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the License.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24201).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 12, 1999.

No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 18, 1999.

Brief description of amendment: The amendment revises the

definition of ``Surveillance Frequency'' to incorporate provisions that

apply upon the discovery of a missed Technical Specification

surveillance. This change allows a delay in performing the actions of

the associated limiting conditions for operation for up to 24 hours or

up to the limit of the specified frequency, whichever is less, when it

is discovered that a surveillance was not performed within its

specified frequency.

Date of Issuance: October 13, 1999.

Effective date: October 13, 1999, and shall be implemented within

30 days.

Amendment No.: 179.

Facility Operating License No. DPR-28. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 9, 1999 (64

FR 48867).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated October 13, 1999.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 8, 1999, as supplemented by letter

dated September 2, 1999.

Brief description of amendment: The amendment increased the

allowable values for engineered safety features actuation system

(ESFAS) loss-of-power 4 kV undervoltage trips in the current Technical

Specifications (TSs) Table 3.3-4 (functional units 8.a and 8.b) and in

surveillance requirement (SR) 3.3.5.3 of the improved TSs. The word

``nominal'' is also added to describe the trip setpoint in SR 3.3.5.3

and in the Bases of the improved TSs. The improved TSs were issued in

Amendment 123 dated March 31, 1999, but have not yet been implemented.

Date of issuance: October 12, 1999.

Effective date: October 12, 1999, to be implemented within 60 days

from the date of issuance.

Amendment No.: 128.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43782).

The September 2, 1999, supplemental letter provided additional

clarifying information, did not expand the scope of the application as

originally noticed, and did not change the staff's initial no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 12, 1999.

No significant hazards consideration comments received: No.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power

Station (YNPS) Franklin County, Massachusetts

Date of application for amendment: March 17, 1999

Brief description of amendment: Revises the Possession Only License

by deleting technical specifications related to hours of work and

putting these requirements in appropriate Administrative Procedures.

Date of issuance: October 8, 1999.

Effective date: October 8, 1999, Implementation of this amendment

includes incorporation of hours of work restrictions into the

Administrative Procedures as described in the licensee's application

dated March 17, 1999, and evaluated in the staff's safety evaluation

attached to the amendment, and written notification to NRC that the

amendment has been fully implemented.

Amendment No.: 153.

Facility Operating License No. DPR-3. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17032).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 8, 1999.

No significant hazards consideration comments received: No.

[[Page 59813]]

Dated at Rockville, Maryland, this 27th day of October 1999.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Deputy Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-28598 Filed 11-2-99; 8:45 am]

BILLING CODE 7590-01-P