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 Pub. L. 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

[[Page 47697]]

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 August 18, 1997, through August 28, 1997.

The last biweekly notice was published on August 27, 1997 (62 FR

45452).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and should cite the publication date and

page number of this Federal Register notice. Written comments may also

be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, 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 October 10, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for 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

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hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: August 15, 1997.

Description of amendment request: The proposed amendment would

revise portions of the facility Technical Specifications regarding

facility staffing and training requirements to power operations. By

letter dated August 7, 1997, the licensee certified permanent cessation

of power operations and permanent removal of fuel from the reactor

vessel. By two letters both dated August 15, 1997, the licensee has

also submitted a related ``Request for Exemption from Certain

Requirements of 10 CFR 50.54, Conditions of License,'' and a ``Request

for Approval of the Certified Fuel Handler Training and Retraining

Program.''

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.

The proposed change does not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The purpose of the proposed change is to eliminate the

requirements for licensed operators and a licensed operator training

program and to replace those with certified fuel handlers and a

certified fuel handler training and retraining program. Since the

plant has permanently ceased operation and will be maintained in a

defueled condition, the range of accidents for which an operator

needs to be trained has significantly diminished such that a

training program of the depth and breadth of that required by 10 CFR

[Part] 55 is no longer needed. In lieu of a 10 CFR [Part] 55

licensed operator training program, a[n] NRC-approved certified fuel

handler training and retraining program will be implemented. Since

this training program will adequately equip appropriate operations

personnel for fuel handling operations, including responses to

abnormal events/accidents, there will be no increase in the

probability of these events occurring or in the consequences of

these events. The proposed changes do not affect plant equipment or

the procedures for equipment operation or response to abnormal

events/accidents.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The purpose of this proposed change is to eliminate the

requirements for licensed operators and a licensed operator training

program and to replace those with certified fuel handlers and a

certified fuel handler training and retraining program. This change

ensures the qualifications of operations personnel are commensurate

with the tasks to be performed and the conditions to be responded

to. This change does not affect plant equipment or the procedures

for operating plant equipment and, therefore, does not create the

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

previously evaluated.

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

The proposed change is to eliminate the requirements for

licensed operators and a licensed operator training program to

replace those with certified fuel handlers and a certified fuel

handler training and retraining program. This change ensures the

qualifications of the operations personnel are commensurate with the

tasks to be performed and the conditions to be responded to. The

assumptions for a fuel handling accident in the Fuel Building are

not affected by the proposed changes. Therefore, the proposed

amendment 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.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578.

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, ME 04011.

NRC Acting Project Director: Ronald B. Eaton.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: August 1, 1997.

Description of amendment request: The amendments would change

Technical Specification Section 4.2.1 of Appendix B to the licenses.

The changes include rewording of the section to generically state that

Public Service Gas & Electric (PSE&G) will adhere to the Section 7,

Incidental Take Statement, approved by the National Marine Fisheries

Service (NMFS). Removing the specific requirements of this section

enables PSE&G to utilize relief granted by the NMFS on a case-by-case

basis.

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 changes are administrative in nature and would in no way

affect the initial conditions, assumptions, or conclusions of the

Salem [Nuclear] Generating Station, Units 1 and 2, accident

analyses. In addition, the proposed changes would not affect the

operation or performance of any equipment assumed in the accident

analyses. Based on the above information, we conclude that the

proposed changes would not significantly increase 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 changes are administrative in nature and would in no way

impact or alter the configuration or operation of the facilities and

would create no new modes of operation. We therefore conclude that

the proposed changes would 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.

As indicated in the discussion of Criterion 1, the changes are

administrative in nature and would in no way affect plant or

equipment operation or the accident analysis. We therefore conclude

that the proposed changes would not result in a significant

reduction in a margin of safety.

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

this review, it appears that the three

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standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: John F. Stolz.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: July 23, 1997.

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TSs) by relocating the reactor

coolant system pressure and temperature limits from the TSs to the

proposed Pressure Temperature Limits Report in accordance with the

guidance provided by Generic Letter 96-03, ``Relocation of the Pressure

Temperature Limit Curves and Low Temperature Overpressure Protection

System Limits.''

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 proposed removal of the Reactor Coolant System (RCS)

pressure temperature (P/T) limits from the Technical Specifications

(TSs) and relocation to the proposed Pressure Temperature Limits

Report (PTLR) in accordance with the guidance provided by Generic

Letter (GL) 96-03 is administrative in that the requirements for the

P/T limits are unchanged. The P/T limits proposed for inclusion in

the PTLR are based on the fluence associated with 2775 MW

[megawatts] thermal power and operation through 36 effective full

power years (EFPY). GL 96-03 requires that the P/T limits be

generated in accordance with the requirements of 10 CFR [Part] 50,

Appendices G and H, documented in an NRC-approved topical report

incorporated by reference in the TSs. Accordingly, the proposed

curves have been generated using the NRC-approved methods described

in WCAP-14040-NP-A, Revision 2, and meet the requirements of 10 CFR

[Part] 50, Appendices G and H. TS 3.4.10.1 will continue to require

that the RCS pressure and temperature be limited in accordance with

the limits specified in the PTLR. The NRC-approved methodology for

generating the P/T limit, WCAP-14040-NP-A, Revision 2, will be

specified in TS 6.9.1.15 and NRC approval will be required in the

form of a TS Amendment prior to changing the methodology. Use of P/T

limit curves generated using the NRC-approved methods described in

WCAP-14040-NP-A, Revision 2, as specified in TS 6.9.1.15, will

provide additional protection for the integrity of the reactor

vessel, thereby assuring that the reactor vessel is capable of

providing its function as a radiological barrier.

TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit 1 and Unit 2

provides the operability requirements for RCS low temperature

overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires

that two residual heat removal (RHR) system suction relief valves

(RHRRVs) be operable or that the RCS be vented at RCS cold leg

temperatures less than or equal to 310 deg.F. GL 96-03 recognizes

that RHRRVs do not have variable pressure lift setpoints and states

that those plants that rely on the RHRRVs for LTOP should continue

to address the LTOP requirements in the TS. Consistent with GL 96-

03, the Farley Unit 1 and Unit 2 requirements for LTOP will be

retained in TS 3.4.10.3.

Based on the above evaluation, the proposed changes are

administrative in nature and do 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.

As stated above, the proposed changes to remove the RCS P/T

limits from the TSs and relocate them to the proposed PTLR is an

administrative change. Consistent with the guidance provided by GL

96-03, the proposed P/T limits contained in the proposed PTLR meet

the requirements of 10 CFR [Part] 50, Appendices G and H, and were

generated using the NRC-approved methods described in WCAP-14040-NP-

A, Revision 2. The proposed changes do not result in a physical

change to the plant or add any new or different operating

requirements on plant systems, structures, or components with the

exception of limiting the number of operating RCPs [reactor coolant

pumps] at RCS temperatures below 110 deg.F. Limiting the number of

operating RCPs below 110 deg.F results in a reduction in the

[delta]P between the reactor vessel beltline and the RHRRVs, thereby

providing additional margin to limits of Appendix G. Provisions are

made to allow the start of a second RCP at temperatures below

110 deg.F in order to secure the pump that was originally operating

without interrupting RCS flow. The LTOP enable temperature exceeds

the minimum LTOP enable temperature determined using the NRC-

approved methods described in WCAP-14040-NP-A, Rev. 2, thereby

providing additional assurance that the LTOP system will be

available to protect the RCS in the event of an overpressure

transient at RCS temperatures at or below 310 deg.F. Using the

methods contained in WCAP-14040-NP-A, Rev. 2, the minimum boltup

temperature for the reactor vessel flange region is 60 deg.F which

is less than the design limits of the fuel cladding. Administrative

controls require a minimum RCS temperature of 68 deg.F when fuel is

loaded in the reactor vessel to protect against brittle failure of

the fuel cladding, and also require that the component cooling water

(CCW) temperature be maintained between 60 deg.F and 105 deg.F

during refueling operations, thus reducing the potential for the RCS

temperature to be less than the minimum boltup temperature specified

in the proposed PTLRs.

As stated in the above response, implementation of the proposed

changes do not result in a significant increase in the probability

of a new or different accident (i.e., loss of reactor vessel

integrity). The RCS P/T limits will continue to meet the

requirements of 10 CFR [Part] 50, Appendices G and H, and will be

generated in accordance with the NRC approved methodology described

in WCAP-14040-NP-A, Rev. 2. Therefore, the proposed changes do not

result in a significant increase in the possibility of a new or

different accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety is not affected by the removal of the RCS

P/T limits from the TSs and relocating them to the proposed PTLR.

The RCS P/T limits will continue to meet the requirements of 10 CFR

[Part] 50, Appendices G and H. To provide additional assurance that

the P/T limits continue to meet the requirements of Appendices G and

H, TS 6.9.1.15 will require the use of the NRC-approved methodology

described in WCAP-14040-NP-A, Rev. 2, to generate P/T limits. The

RCS LTOP requirements will be retained in TS 3.4.10.3 due to use of

the RHRRVs for LTOP, consistent with the guidance provided by GL 96-

03. The LTOP enable temperature exceeds the LTOP enable temperature

determined in accordance with the NRC-approved methodology, thus

protecting the RCS in the event of a low temperature overpressure

transient over a broader range of temperatures than required by

WCAP-14040-NP-A, Rev. 2. Administrative procedures preclude

operation of the RCS at temperatures below the minimum boltup

temperature for the reactor vessel head, thus precluding the

possibility of tensioning the reactor vessel head at RCS

temperatures below the minimum boltup temperature. Operation of the

plant in accordance with the RCS P/T limits specified in the PTLR

and continued operation of the LTOP system in accordance with TS

3.4.10.3 will continue to meet the requirements of 10 CFR [Part] 50,

Appendices G and H, and will therefore, assure that a margin of

safety is not significantly decreased as the result of the proposed

changes.

Based on the preceding analysis, SNC [Southern Nuclear Operating

Company, Inc.] has determined that removal of the RCS P/T limits

from the TS and relocation to the proposed PTLR will not

significantly increase the probability or consequences of an

accident previously evaluated, create the possibility of a new or

different kind of accident from any accident previously evaluated,

or involve a significant reduction in a margin of safety. SNC

therefore concludes that the proposed change meets the requirements

of 10 CFR 50.92(c) and does

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not involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration. Local

Public Document Room location: Houston-Love Memorial Library, 212 W.

Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302. Attorney

for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office

Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201. NRC

Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: June 20, 1997 (TS-97-004).

Description of amendment request: The proposed amendment would be

an administrative change that would revise the analytical methodology

used to determine the low temperature overpressure protection (LTOP)

event heatup and cooldown curves. This revised methodology would be

incorporated by reference in the Watts Bar Nuclear Plant (WBN), Unit 1

Technical Specification (TS) 5.9, ``Reporting Requirements,'' Section

5.9.6, ``Reactor Coolant System (RCS) Pressure and Temperature Limits

Report (PTLR),'' upon approval for use by the U.S. Nuclear Regulatory

Commission (NRC). The revised methodology extends the current LTOP

requirements through the end of 7 effective full power years (EFPY).

The only technical change being proposed is the substitution of the 7

EFPY American Society of Mechanical Engineering (ASME), Appendix G,

heatup and cooldown curves adjusted by ASME Code Case N-514, ``Low

Temperature Overpressure Protection'' in place of the current 1.5 EFPY

curves as the bounding curves for the LTOP setpoints. This change will

not impact the current 10 CFR 50, Appendix G, pressure/temperature (P/

T) limit curves used for heatup and cooldown that are based on 7 EFPY.

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:

The Nuclear Regulatory Commission has provided standards for

determining whether a significant hazards consideration exists (10

CFR 50.92). A proposed amendment to an operating license for a

facility involves no significant hazards consideration if operation

of the facility, in accordance with the proposed amendment, would

not: (1) Involve a significant increase in the probability or

consequences of an accident previously evaluated: or (2) create the

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

previously evaluated: or (3) involve a significant reduction in a

margin of safety. Each standard is discussed below for the proposed

amendment.

(1) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The LTOP setpoints (identified as the cold overpressure

mitigation system (COMS) for WBN), adjusted for instrument

inaccuracy, pressure differential, and setpoint overshoot by the

scaling and setpoint documents (SSDs), ensure that the 10 CFR 50,

Appendix G P/T [pressure and temperature] limits based on 7 EFPY are

not exceeded by more than the provisions of ASME Code Case N-514,

and therefore, ensure that the RCS integrity is maintained.

The change does not modify the RCS pressure boundary, nor make

any physical changes to the facility design, material, construction

standards, or setpoints. The LTOP enabling temperature based on TS

3.4.12, ``Cold Overpressure Mitigation System (COMS),'' is [less

than or equal to] 350 degrees F and is more conservative than a

value of 271.1 degrees F (RTNDT + 90 degrees F) based on

7 EFPY. This temperature would be acceptable based on NRC Branch

Technical Position-Reactor Systems Branch (BTP-RSB)-5.2,

``Overpressurization Protection of Pressurized Water Reactors While

Operating at Low Temperatures.'' The LTOP enabling temperature

remains unchanged by this proposed amendment. The probability of a

LTOP event occurring is independent of the P/T limits for the RCS

pressure boundary; therefore, the probability of an LTOP event

occurring remains unchanged.

The calculation of the P/T limits in accordance with approved

regulatory methods based on 7 EFPY provides assurance that reactor

pressure vessel fracture toughness requirements are met and the

integrity of the RCS pressure boundary is maintained. LTOP setpoints

based on 1.5 EFPY P/T limits have provided margin such that a

pressure excursion exceeding the 7 EFPY limits would not exceed the

1.5 EFPY limits. This margin between the 7 EFPY curves and the LTOP

setpoints is maintained by changing the bounding curves for the LTOP

setpoints to 7 EFPY curves adjusted by the provisions of ASME Code

Case N-514. The only technical change being made is the bounding

curves which provide the basis for the current LTOP setpoints.

The use of theoretical fluence for generating the P/T curves to

be used for the first 7 EFPY is appropriate and was submitted July

31, 1995, with the WBN Unit 1 PTLR, Revision 4 and WCAP-13829,

Revision 2, ``Heatup and Cooldown Limit Curves for Normal Operation

for Watts Bar Unit 1.'' The present 7 EFPY curves are generated

using a theoretical value for fluence calculated by Westinghouse in

accordance with NRC approved methodology since WBN had no

surveillance capsule data available at the time of plant startup.

This value for fluence is conservative, and the actual fluence to

the intermediate shell forging (the controlling beltline material)

is expected to be significantly less than the theoretical value used

to generate the initial 7 EFPY curves since WBN is transitioning to

a low-leakage core. The LTOP bounding curves are based on 7 EFPY

curves adjusted in accordance with ASME Code Case N-514 which were

generated using the same theoretical fluence as used for the P/T

curves. The significance of using the theoretical value of fluence

in generating these curves is the additional margin that exists

between the 7 EFPY theoretical curves and curves that would be

generated using actual fluence values from capsule data. This

additional margin reduces the significance of changing the LTOP

basis from the 1.5 EFPY curves to the 7 EFPY curves adjusted for

ASME Code Case N-514.

This change does not adversely affect the integrity of the RCS

such that its function in the control of radiological consequences

is affected. In addition, the change does not affect any fission

barrier. The change does not degrade or prevent the LTOP power

operated relief valves (PORVs) or other safety related systems from

responding to accidents described in Chapter 15 of the Final Safety

Analysis Report (FSAR). In addition, the change does not alter any

assumptions previously made in the radiological consequences of an

accident described in the FSAR. Therefore, the consequences of an

accident previously evaluated in the FSAR are not increased. Thus,

the operation of WBN Unit 1 in accordance with this proposed

amendment does not involve a significant increase in the probability

or consequences of any accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The Appendix G P/T limitations were prepared using methods

derived from the ASME Boiler and Pressure Vessel Code Section III

and the criteria set forth in NRC Regulatory Standard Review Plan

5.3.2, ``Pressure-Temperature Limits.'' The use of ASME Code Case N-

514 and the theoretical fluence value for 7 EFPY does not modify the

RCS pressure boundary, nor make any physical changes to the LTOP

setpoints or system design. The proposed change was prepared in

accordance with regulatory requirements and provides evaluation of

LTOP events based on 7 EFPY theoretical fluence which is more

limiting than actual expected neutron exposure for that same period.

This proposed change is an administrative change which

incorporates by reference the use of an NRC approved methodology;

therefore, the change does not cause the initiation of any accident

nor create any new creditable limiting failure for safety-related

[[Page 47701]]

systems and components. The change does not result in an event

previously deemed incredible being made credible. As such, it does

not create the possibility of an accident different than any

evaluated in the FSAR.

The change does not have any effect on the ability of the

safety-related systems to perform their intended safety functions.

The change does not create failure modes that could adversely impact

safety-related equipment. Therefore, it will not create the

possibility of a malfunction of equipment important to safety

different than previously evaluated in the FSAR. Thus, the proposed

amendment does not create the possibility of a new or different kind

of accident from any accident previously evaluated.

Therefore, the proposed change does not create the possibility

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

evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in margin of

safety.

The 10 CFR 50, Appendix G P/T limitations were prepared using

methods derived from ASME Section III and criteria set forth in NRC

Regulatory Standard Review Plan 5.3.2. These documents along with

the calculational limitations specified in 10 CFR 50.61 are an

acceptable method for implementing the requirements of 10 CFR 50

Appendices G and H. Inherent conservatisms in the P/T limits

resulting from these documents include:

a. An assumed defect in the reactor vessel wall with a depth

equal to \1/4\ of the thickness (T) of the vessel wall and a length

equal to 1\1/2\ times the thickness of the vessel wall.

b. Assumed reference flaw oriented in both longitudinal and

circumferential directions and limiting material property. At WBN,

the only weld in the core region is oriented in the circumferential

direction.

c. A factor of safety of 2 is applied to the membrane stress

intensity factor.

d. The limiting toughness is based upon a reference value

(KIM) which is the lower bound of the dynamic crack

initiation and arrest toughness.

e. A 2-sigma margin term is applied in determining the adjusted

reference temperature (ART) that is used in calculating the limiting

toughness.

Beyond the conservatisms described above, WBN has the following

additional margin:

a. The value of fluence used in the calculation of the WBN Unit

1 Appendix G P/T limits is a theoretical value calculated by NRC

approved methodology.

b. The ART for 7 EFPY is based on the theoretical value for

fluence and therefore is conservative. The LTOP enabling temperature

of [less than or equal to] 350 degrees F in accordance with TS

3.4.12 is conservative with respect to (RTNDT + 90

degrees F) which based on an ART of 181.1 degrees F would equal

271.1 degrees F. An enabling temperature of (RTNDT + 90

degrees F) is based on NRC BTP-RSB 5.2.

The ASME Working Group for Operating Plant Criteria developed

Code Case N-514 as an alternative methodology to the safety margin

requirements of Appendix G to 10 CFR 50. The Code Case provides

criteria to determine pressure limits during LTOP events that avoid

certain operational restrictions, provide adequate margins against

failure of the reactor vessel, and reduce the potential for

unnecessary activation of the relief valves used for LTOP.

Specifically, the N-514 Code Case allows determination of the LTOP

setpoints such that for LTOP events the maximum pressure in the

reactor vessel would not exceed 110% of the P/T limits of the

existing ASME Appendix G curves, and redefines the enabling

temperature as a coolant temperature less than 200 degrees F or a

reactor vessel metal temperature less than RTNDT + 50

degrees F. Code Case N-514 has been approved by the ASME Code

Committee and its content has been incorporated in Appendix G of

ASME Section XI and published in the 1993 Addenda and 1995 Edition.

Code Case N-514 has not been approved for use in Regulatory Guide

1.147, ``Inservice Inspection Code Case Acceptability, ASME Section

XI;'' however, it has been included in the Draft Regulatory Guide

1.147 (Task DG-1050) which is currently out for public review and

comment. As stated above, WBN Unit 1 uses Appendix G for the P/T

limits for plant operation and an LTOP enabling temperature greater

than RTNDT + 90 degrees F which is more conservative than

the alternative methodology contained in Code Case N-514.

The need for implementation of Code Case N-514 at WBN involves

the avoidance of certain operational restrictions associated with

low temperature operation of the plant. Use of Appendix G P/T limits

to determine the PORV setpoints would result in pressure setpoints

within the operating window; consequently, no margin would be

available for normal operating pressure surges. Therefore, operating

with these limits could result an unnecessary challenge to the PORVs

and cavitation of the reactor coolant pumps (RCP) during normal

operation. Additionally, the need to raise the RCS inventory by

external heating methods to a temperature high enough to avoid PORV

activation when starting a RCP from a RCS cold shutdown condition

could result in undesirable thermal transients in the RCS.

Utilizing the methodology set forth in the ASME Boiler and

Pressure Vessel Code Section XI, Appendix G, which includes the

provisions of Code Case N-514, NRC Regulatory Standard Review Plan

5.3.2, 10 CFR 50.61, and 10 CFR 50, Appendices G and H with the

above additional margins ensures that proper limits and conservative

safety factors are maintained. Thus the proposed change does not

significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document rooms for

the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of application for amendments: March 18, 1997.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) to increase the High Pressure Coolant

Injection (HPCI)

[[Page 47702]]

system low pressure isolation setpoint from greater than 80 psig to

greater than 100 psig.

Date of issuance: August 21, 1997.

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

Amendment Nos.: 161, 156.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17228).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 21, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: June 12, 1997.

Brief description of amendments: The amendments change the name

``Duke Power Company'' to ``Duke Energy Corporation'' in the Catawba

operating licenses and appendices as a result of Duke Power Company's

recent name change.

Date of issuance: August 22, 1997.

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

30 days.

Amendment Nos.: 161 and 153.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Facility Operating Licenses.

Date of initial notice in Federal Register: July 2, 1997 (62 FR

35848).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 22, 1997, and an Environmental

Assessment dated July 31, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: June 12, 1997.

Brief description of amendments: The amendments change the name

``Duke Power Company'' to ``Duke Energy Corporation'' in the McGuire

operating licenses and appendices as a result of Duke Power Company's

recent name change.

Date of issuance: August 26, 1997.

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

30 days.

Amendment Nos.: 176 and 158.

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Licenses.

Date of initial notice in Federal Register: July 2, 1997 (62 FR

35848).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 26, 1997. An Environmental

Assessment was issued and dated August 15, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: October 26, 1995 and supplemented by

letters dated April 7 and July 30, 1997.

Brief description of amendment: The amendment revised the technical

specifications for 16 editorial changes and deletes the reuirement for

a program to prevent and detect Asiatic Clams (Corbicula) in the

service water system (SWS). The Corbicula program is no longer needed

because the facility has been modified and SWS no longer takes water

from the Mississippi River; source of the larvae and infestation.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 95.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 6, 1995 (60 FR

62492).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 26, 1997.

No significant hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: November 15, 1996, as supplemented May 9

and August 15, 1997.

Brief description of amendment: The amendment revises the technical

specifications to increase the two recirculation loop Minimum Critical

Power Ratio (MCPR) from 1.07 to 1.10 and the single recirculation loop

MCPR limit from 1.08 to 1.12.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 96.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 2, 1997 (62 FR

127).

The May 9 and August 15, 1997, submittal provided clarifying

information that 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 August 26, 1997.

No significant hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: January 20, 1997 as supplemented by

letter dated July 7, 1997.

Brief description of amendment: The amendment revises the technical

specifications to allow the use of flow control spectral shift

strategies to increase cycle energy. The revision is based on a Maximum

Extended Load Line Limit (MELLL) analysis for the River Bend Station.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 97.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications/operating license.

Date of initial notice in Federal Register: February 26, 1997 (62

CFR 8799).

The July 7, 1997 submittal provided clarifying information and did

not change the initial no significant hazards consideration

determination.

[[Page 47703]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 26, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana

Date of amendment request: November 6, 1996, as supplemented by

letter dated July 31, 1997.

Brief description of amendment: The amendment revises the Technical

Specifications to delete the requirement for the Penetration Valve

Leakage Control System. The licensee requested deferal of the proposal

to increase the allowed leakage by main steam isolation valves and to

delete the requirement for the Main Steam Positive Leakage Control

System.

Date of issuance: August 26, 1997.

Effective date: August 26, 1997.

Amendment No.: 98.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 2, 1997 (62 FR

125).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 26, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

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

Mississippi Electric Power Association, and Mississippi Power & Light

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

Claiborne County, Mississippi

Date of application for amendment: October 22, 1996, as

supplemented by letter dated June 26, 1997.

Brief description of amendment: The amendment revises Figure

3.4.11-1, ``Minimum Reactor Vessel Metal Temperature vs. Reactor Vessel

Pressure,'' in Limiting Condition for Operation 3.4.11, ``RCS [Reactor

Coolant System] Pressure and Temperature (P/T) Limits,'' of the

Technical Specifications. The previous figure was only up to 10

Effective Full Power Years (EFPYs) and this amendment revises the

figure up to 32 EFPYs. There are now five curves of Figure 3.4.11-1 for

five different EFPY periods: up to 16, 16 to 20, 20 to 24, 24 to 28,

and 28 to 32. The licensee submitted two sets of curves. The first set

replaced TS Figure 3.4.11-1. The second set were duplicates of the

first set except the second set also contained detailed information

used in development of the curves and would be included in the next

update of the Updated Final Safety Analysis Report. There were also

minor additions to Surveillance Requirements (SRs) 3.4.11.1 and

3.4.11.2 to have the SRs reference the ``applicable Figure 3.4.11-1

based on the current effective full power year (EFPY).''

Date of issuance: August 27, 1997.

Effective date: August 27, 1997.

Amendment No: 132.

Facility Operating License No. NPF-29: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8797).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 27, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 11, 1997.

Brief description of amendment: The amendment modifies Technical

Specifications 3.3.3.7.3, and Surveillance Requirements (SR) 4.3.3.7.3

for the broad range gas detection system. Also it includes some changes

to the Bases in Section 3/4.3.3.7 to incorporate information associated

with the proposed modifications. The licensee is planning to replace

the existing toxic gas monitors in the system with a new, more advanced

gas monitors.

Date of issuance: August 19, 1997.

Effective date: August 19, 1997, to be implemented within 90 days.

Amendment No.: 133.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 7, 1997 (62 FR

24987)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 19, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin

I. Hatch Nuclear Plant, Unit 1, Appling County, Georgia

Date of application for amendment: April 29, 1997, as supplemented

by letter dated May 28, 1997.

Brief description of amendment: The amendment revises Hatch Unit 1

reactor vessel pressure and temperature limits to reflect data

collected from the material sample recovered during the March 1996 Unit

1 outage.

Date of issuance: August 19, 1997.

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

30 days.

Amendment No.: 207.

Facility Operating License No. DPR-57: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38138).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 19, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 4, 1996, as supplemented

June 10 and August 15, 1997 (TSCR 250).

Brief description of amendment: The amendment changes the Safety

Limit Minimum Critical Power Ratio and as a result, the operating

Minimum Critical Power Ratio. The amendment also capitalized certain

definitions and provided a uniform type font for Sections 2.1 and 3.10.

Date of Issuance: August 26, 1997.

Effective date: August 26, 1997, with full implementation within 30

days.

Amendment No.: 192.

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 6, 1996 (61 FR

57484).

The Commission's related evaluation of this amendment is contained

in a

[[Page 47704]]

Safety Evaluation dated August 26, 1997.

The June 10 and August 15, 1997, submittals provided clarifying

information that did not alter the staff's initial proposed no

significant hazards considerations determination.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: April 21, 1997, as supplemented

July 17, 1997.

Brief description of amendment: The amendment reduces the required

volume of borated water in each core flood tank from 1040 ft \3\ to 940

ft \3\, reduces the required high pressure injection pump flowrate from

500 gallons per minute (gpm) to 431 gpm, and deletes the local manual

valve operability option for decay heat system valves DH-V-6A and DH-V-

6B.

Date of issuance: August 27, 1997.

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

30 days.

Amendment No.: 203.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27795).

The July 17, 1997, submittal provided clarifying information that

did not alter the initial no significant hazards determination. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated August 27, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: April 22, 1997.

Brief description of amendments: The proposed amendment revised

Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6, Core

Operating Limits Report, to allow use of an alternate zirconium-based

fuel cladding, ZIRLO, and limited substitution of fuel rods by ZIRLO

filler rods.

Date of issuance: August 19, 1997.

Effective date: August 19, 1997.

Amendment Nos.: Unit 1--Amendment No. 89; Unit 2--Amendment No. 76.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27795).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 19, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: November 20, 1995.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by providing clarifications to the applicability

and action statements in TS Table 3.3-12 relating to the Steam

Generator Blowdown Monitor and the Condensate Polishing Facility Waste

Neutralizing Sump radiation monitor.

Date of issuance: August 26, 1997.

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

60 days.

Amendment No.: 207.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 20, 1995 (60

FR 65683).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 26, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: May 1, 1997.

Brief description of amendment: Technical Specifications 3/4.8.2.2

and 3/4.8.3.2 specify which electrical power systems are required to be

operable in Modes 5 and 6. The amendment clarifies the requirements by

identifying the specific equipment required and their alignments in

Modes 5 and 6.

Date of issuance: August 21, 1997.

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

within 60 days.

Amendment No.: 146.

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30637).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 21, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: May 5, 1997.

Brief description of amendment: Technical Specification

Surveillance 4.5.2.b.1 requires that the emergency core cooling system

piping be verified full of water at least once per 31 days. The

amendment revises the surveillance to exempt the operating charging

pump(s) and associated piping from the requirement to be verified full

of water and moves the description of the verification method from the

surveillance to the Bases section.

Date of issuance: August 28, 1997.

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

within 60 days.

Amendment No.: 147.

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30638).

The Commission's related evaluation of the amendment is contained

in a

[[Page 47705]]

Safety Evaluation dated August 28, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope.

PECO Energy Company, Public Service Electric and Gas Company Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: March 31, 1997, as supplemented

by letter dated June 25, 1997.

Brief description of amendments: These amendments extend the APRM

flow bias instrumentation surveillance interval from 18 months to 24

months. This will eliminate the need to perform on-line APRM

surveillance testing, which requires plant operators to place an

operating unit in a half scram configuration.

Date of issuance: August 19, 1997.

Effective date: Units 2 and 3 effective as of date of issuance.

Amendments Nos.: 219 and 222.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 7, 1997 (62 FR

24988).

The supplemental letter provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 19, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Southern Nuclear Power Company, Inc., Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric Authority of Georgia, City of

Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: June 13, 1997, as supplemented

by letter dated July 18, 1997.

Brief description of amendments: The amendments revise the

pressurizer safety relief valve setpoint specified in Technical

Specification 3.4.10.

Date of issuance: August 26, 1997.

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

Unit 1 prior to or after initial entry into Mode 3 (in accordance with

the provisions of the note to the Applicability for LCO 3.4.10)

following the fall 1997 refueling outage; for Unit 2 prior to or after

initial entry into Mode 3 (in accordance with the provisions of the

note to the Applicability for LCO 3.4.10) following the spring 1998

refueling outage.

Amendment Nos.: 98 and 76.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38139).

The supplemental material did not change the no significant hazards

finding or expand the scope of the Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 26, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: December 7, 1994 (TXX-94326), as

supplemented by letter dated June 21, 1996 (TXX-96384).

Brief description of amendments: These changes revised Section

3.7.1.5 of the Technical Specification to increase the Allowed Outage

Time for one inoperable Main Steam Isolation Valve (MSIV) while in Mode

1, and to clarify requirements related to inoperable MSIVs while in

Modes 2 and 3.

Date of issuance: August 18, 1997.

Effective date: August 18, 1997, to be implemented within 60 days.

Amendment Nos.: 54 and 40.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 1995 (60 FR

6312).

The additional information contained in the supplemental letter

dated June 21, 1996, was clarifying in nature and thus, within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 18, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: April 14, 1997 (TSCR 197), as

supplemented on August 11, 1997.

Brief description of amendments: These amendments revise Technical

Specifications (TS) Sections 15.6.2, ``Organization,'' TS 15.6.5.1,

``Manager's Supervisory Staff,'' TS 15.6.6, ``Reportable Event

Action,'' TS 15.6.7, ``Actions To Be Taken If A Safety Limit Is

Exceeded,'' and TS 15.7.8, ``Administrative Controls,'' by changing the

title of the corporate officer responsible for nuclear operations from

the ``Vice President-Nuclear Power,'' to the ``Chief Nuclear Officer.''

Date of issuance: August 25, 1997.

Effective date: August 25, 1997, with full implementation within 45

days.

Amendment Nos.: 177 and 181.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27802), as corrected May 29, 1997 (62 FR 29163) The August 11, 1997,

submittal provided a corrected TS page. This information was within the

scope of the action noticed and did not change the staff's initial

proposed no significant hazards considerations determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 25, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Dated at Rockville, Maryland, this 3rd day of September 1997.

For The Nuclear Regulatory Commission.

Bruce E. Boger,

Director, Division of Reactor Projects--I/II Office of Nuclear Reactor

Regulation.

[FR Doc. 97-23820 Filed 9-9-97; 8:45 am]

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