Biweekly Notice

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 August 29, 1997, through September 12, 1997.

The last biweekly notice was published on September 10, 1997 (62 FR

47696).

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

[[Page 50001]]

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 24, 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 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.

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: July 15, 1997

Description of amendment request: The proposed amendments would:

(1) add Technical Specification (TS) 3.5.7, ``Main Steam Line Break

Detection and Feedwater Isolation,'' to identify operability

requirements and Bases for the main steamline break (MSLB) detection

isolation circuitry, the feedwater isolation circuitry, the main

feedwater main control valves, and the main feedwater startup control

valves; (2) revise TS 3.5.1, ``Operation Safety Instrumentation'' to

add a reference to TS 3.5.7; (3) revise Table 3.5.1-1, ``Instruments

Operating Conditions,'' to reflect operability requirements for the

main steam header pressure and MSLB detection channels, the feedwater

isolation channels, and the feedwater isolation channels manual

pushbuttons; and (4) revise Table 4.1-1, ``Instrument Surveillance

Requirements,'' and Table 4.1-2, ``Minimum Equipment Test Frequency,''

to include surveillance requirements for the subject circuitry and

components.

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:

[[Page 50002]]

A. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

NO

This proposed Technical Specification amendment does not create

any conditions or events which lead to accidents (events) previously

evaluated in the UFSAR [Updated Final Safety Analysis Report], other

than a loss of Main Feedwater (FDW). The new MSLB detection and

feedwater isolation circuitry addressed by this change is designed

so that a credible single failure will not cause a loss of FDW to

the steam generator unless [an] MSLB is detected. Single failures

are not assumed if entry into a Technical Specification action

statement occurs.

During [an] MSLB, the circuitry is intentionally stopping and

isolating FDW. Operators are currently instructed to isolate FDW on

indication of [an] MSLB. The new circuitry will automatically stop

FDW to eliminate the need for this operator action. Thus the

probability of the stopping (loss) of FDW is not increased. The NRC

has also stated that the stopping of FDW to mitigate [an] MSLB is an

acceptable response to address the concerns of Inspection and

Enforcement Bulletin 80-04.

The Emergency Feedwater (EFW) System is an accident mitigation

system. The MSLB modification and associated Technical Specification

to keep the turbine driven emergency feedwater pump (TDEFW) pump

from starting following [an] MSLB will not initiate any accidents.

The potential for containment overpressurization currently

exists without the installed modification and associated Technical

Specification. The new MSLB detection and feedwater isolation

circuitry will assist in reducing the potential for the

overpressurization of containment. The EFW circuitry is designed so

that the TDEFWP will still auto start for any event other than [an]

MSLB. The TDEFWP can still be manually started during [an] MSLB or

FDW line break accident as needed. This action is similar to other

manual actions to align EFW for the MSLB scenarios that are already

described in the ONS [Oconee Nuclear Station] UFSAR. This new

circuitry and associated Technical Specification creates no new

credible single failures that could prevent the TDEFWP from auto

starting (except for the MSLB). The motor driven EFW pumps and EFW

flow control valves are not adversely affected by this change and

will provide EFW flow for scenarios other than Station Blackout.

Both FDW and EFW will still provide their design functions of

supplying feedwater to the steam generators, as evaluated in the

UFSAR. The ability to shut down following a 10CFR50 Appendix R fire

is not adversely affected. This Technical Specification change does

not adversely affect containment integrity and radiological release

pathways.

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

from the accident previously evaluated?

NO

No accidents different than already evaluated in the UFSAR are

postulated. The FDW System will still perform its design function of

supplying feedwater to the steam generators as evaluated in the

UFSAR. The EFW System will still provide its function of supplying

feedwater to the steam generators, as evaluated in the UFSAR for

events resulting in the loss of the FDW System.

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

NO

The design pressure of containment is specified to be 59 psig in

the bases to several Technical Specifications. With the potential

for unrestricted FDW and EFW flow during [an] MSLB inside

containment, the design pressure of the containment could be

exceeded. The proposed Technical Specifications address equipment

which will function to isolate FDW in the unlikely event of [an]

MSLB accident. Therefore, the proposed Technical Specifications do

not increase the potential for the containment to be pressurized or

increase the expected pressure of containment following [an] MSLB.

No plant safety limits, set points, or design parameters are

adversely affected. The fuel, fuel cladding, and Reactor Coolant

System are not impacted.

Duke [Duke Energy Corporation] has concluded based on the above

that there are no significant hazards considerations involved in

this amendment request.

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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee

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

Date of amendment request: August 28, 1997 (TSC 96-09)

Description of amendment request: The proposed changes would add

new limiting conditions for operation and new surveillance requirements

for the Emergency Condenser Circulating Water System, the Essential

Siphon Vacuum System, and the Siphon Seal Water System to reflect

design changes and modifications to these systems.

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?

NO.

This Technical Specification change does not create any

conditions or events which lead to accidents previously evaluated in

the UFSAR [Updated Final Safety Analysis Report]. The new ECCW

[Emergency Condenser Circulating Water] System Technical

Specification 3.19, along with the new ECCW Surveillance

requirements specified in Technical Specification Table 4.1-2, are

conservative in nature. No existing Technical Specification

requirements are being deleted with this revision. Surveillance and

operability requirements are being added for the upgraded ECCW

System.

The ECCW System is only required following the occurrence of

loss of offsite power (LOOP) events. The most limiting of these LOOP

events is the loss of coolant accident concurrent with the LOOP

(LOCA/LOOP). Therefore, the ECCW System is not considered to be an

accident initiator. As a result, the proposed new ECCW Technical

Specification requirements will not result in any increase in the

probability of any design basis accidents or events evaluated in the

UFSAR.

The credit for restarting a CCW [Condenser Circulating Water]

pump within 1.5 hours following a LOOP, to ensure suction to LPSW

[Low Pressure Service Water] is maintained, is being replaced by

credit for maintaining the ECCW siphon using the new siphon support

systems (ESV [Essential Siphon Vacuum] System and SSW [Siphon Seal

Water] System) in conjunction with the upgraded ECCW System.

Therefore, obsolete requirements specified in Selected Licensee

Commitments (SLCs) 16.9.7 and 16.9.8 will be revised or deleted

accordingly. Replacement of the CCW pump restart during a LOOP with

the ability to maintain ECCW siphon flow will not create any

conditions or events which lead to accidents previously evaluated in

the UFSAR.

The modifications to upgrade the ECCW System were performed to

improve the reliability of the ECCW System. The proposed new ECCW

Technical Specification provides additional surveillance and

operability requirements to ensure that the upgraded ECCW System

will function reliably during the design basis events which require

its operation. Therefore, these proposed new Technical Specification

requirements will not increase the consequences of any accidents

previously evaluated in the UFSAR.

[2. Will the change] create the possibility of a new or

different kind of accident from the accident previously evaluated?

NO.

No accidents different than those already evaluated in the UFSAR

are postulated. The upgraded ECCW System will more reliably perform

its design function of supplying water to the suction of the Low

Pressure Service Water (LPSW) System as evaluated in the UFSAR. The

new Technical Specification requirements will increase the

reliability of the upgraded ECCW System. In addition, the ECCW

System is not an accident initiator since it is used following

certain design basis events such as a LOCA/LOOP.

[[Page 50003]]

[3. Will the change] involve a significant reduction in a margin

of safety?

NO.

The proposed Technical Specifications address equipment which

will function in certain design basis events, such as a LOCA/LOOP,

to ensure a reliable water supply to the LPSW System. The LPSW

System must function to remove decay heat from primary systems and

the reactor building during a LOCA/LOOP. The proposed Technical

Specifications addressing the upgraded ECCW System will further

enhance the reliability of the ECCW System and will result in

greater assurance that the LPSW System can perform its safety

functions. No plant safety limits, setpoints, or design parameters

are adversely affected. The fuel, fuel cladding, and Reactor Coolant

System are not impacted. The proposed Technical Specifications

provide additional, conservative, operational requirements beyond

the current Technical Specifications which address the ECCW System.

Duke [Duke Energy Corporation] has concluded based on this

information that there are no significant hazards considerations

involved in this amendment request.

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

thisreview, 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

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: September 4, 1997

Description of amendment request: The proposed changes would

incorporate changes to the Oconee Final Safety Analysis Report and

Technical Specification Bases to address a potential unreviewed safety

question associated with implementation of revised small break loss-of-

coolant accident analysis. The proposed changes would address operation

of the facility and single failure criteria related to the high

pressure injection 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) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

No. None of the proposed changes [have] any impact upon the

probability of any accident which has been evaluated in the UFSAR

[Updated Final Safety Analysis Report].

None of these changes have any impact upon the ability of the

HPI [high-pressure injection] System to mitigate the consequences of

a small break LOCA [loss-of-coolant accident], which is addressed

below. The small break LOCA is the limiting design basis accident

with respect to the HPI System operability requirements.

The proposed changes to the Bases of Specification 3.3.1 and

Chapter 15 of the Oconee UFSAR include operator actions that have

not previously been reviewed and approved by the [NRC] staff for

licensing basis small break LOCA analyses. However, these operator

actions have been included in the Emergency Operating Procedure for

over 10 years and crediting these actions in the safety analyses

does not result in any change to the operator's response to a small

break LOCA. These actions are simply changes to the assumptions

contained in the licensing basis small break LOCA analyses. The

operability requirements for the HPI System contained in

Specification 3.3.1 are supported by a spectrum of small break LOCA

analyses based on the approved Evaluation Model described in FTI

[Framatome Technologies, Inc.] topical report BAW-10192P. These

small break LOCA analyses demonstrate that the acceptance criteria

of 10CFR 50.46 are satisfied.

The operability requirements in Technical Specification 3.3.1.c

assure that the HPI System can withstand the worst single failure

and still result in two HPI pumps injecting through two trains. The

full power small break LOCA analyses supporting this proposed

license amendment have been performed in accordance with the

approved Evaluation Model described in FTI topical report BAW-

10192P.

When at or below 75% FP [full power], one HPI train provides

sufficient flow to mitigate a small break LOCA. The 60% power level

currently in Specification 3.3.1 is justified by analyses using the

Evaluation Model described in FTI topical report BAW-10192P,

considering the worst case break location and size described in LER

[Licensee Event Report] 269/90-15 and Attachment 2 to this

submittal. The proposed changes to the Bases of Technical

Specification 3.3.1 describe the operator actions credited to

justify the adequacy of the current specification and eliminate the

need for the administrative restrictions imposed by LER 269/90-15.

These requirements ensure that, following the worst single failure,

one train of HPI would remain available to mitigate a small break

LOCA.

In summary, the technical analyses described in this license

amendment justify the adequacy of this specification and assure that

operability of the HPI System is maintained in a manner consistent

with the requirements of the design basis accidents. Therefore, it

is concluded that this amendment request will not significantly

increase the probability or consequences of an accident previously

evaluated.

(2) Create the possibility of a new or different kind of

accident from any kind of accident previously evaluated:

No. The proposed changes to the Bases of Technical Specification

3.3.1 and Chapter 15 of the Oconee UFSAR do not result in any new

operator actions or changes in plant operation. The proposed changes

involve crediting operator actions in the licensing basis small

break LOCA analyses that have been included in the Emergency

Operating Procedure for years. No new initiating events or

potentially unanalyzed conditions have been created. Therefore, this

proposed amendment will not create the possibility of any new or

different kind of accident.

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

No. The HPI System requirements associated with the proposed

UFSAR and Technical Specification Bases changes are supported by

analyses which demonstrate that the acceptance criteria of 10 CFR

50.46 are not violated for any small break LOCA. These analyses were

performed in accordance with the Evaluation Model described in FTI

topical report BAW-10192P. Therefore, it is concluded that the

proposed amendment request will not result in a significant decrease

in the margin of safety.

Duke [Duke Energy Corporation] has concluded, based on the

above, that there are no significant hazards considerations involved

in this amendment request.

The NRC 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: August 6, 1997

Description of amendment request: The proposed amendment would

eliminate the provisions in Technical Specification 3.8.1, ``AC Sources

- Operating,'' for accelerated testing of the emergency diesel

generators (DG). The proposed changes are the following: (1) the

frequency of verifying DG starts and operation in Surveillance

Requirements 3.8.1.2 and 3.8.1.3, respectively, would be changed to 31

days, from the present reference to Table 3.8.1-1, and (2) Table 3.8.1-

1, ``Diesel Generator Test

[[Page 50004]]

Schedule,'' would be deleted. The emergency DG provide emergency AC

power to the site with the loss of offsite AC power.

Basis for proposed no significant hazardsconsideration

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. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

[These] change[s] will provide flexibility to structure the

standby diesel generator maintenance program based on the risk

significance of the structures, systems, and components [(SSCs)]

that are within the scope of the Maintenance Rule [(10 CFR 50.65,

``Requirements for Monitoring the Effectiveness of Maintenance at

Nuclear Power Plants)]. The removal of the diesel generator

accelerated testing is acceptable as the maintenance rule applies

site and system specific performance criteria to monitor diesel

generator performance. This criteria includes a running availability

and reliability goal as well as specific goals to monitor

maintenance preventable functional failures. The performance

criteria for the diesel generator reliability and availability

established by the maintenance rule and the causal determinations

and corrective actions required for maintenance preventable

functional failures are considered to be an acceptable method for

monitoring diesel generator performance.

The proposed change[s] [have] no effect on the probability of

the initiation of an accident, because the emergency diesel

generators do not serve as the initiator of any event. Additionally,

as diesel generator performance will continue to be [ensured] by the

maintenance rule, the proposed changes do not affect the ability to

mitigate the consequences of an accident previously evaluated. The

changes do not impact the diesel [generator]'s design sources,

operating characteristics, system functions, or system

interrelationships. The failure mechanisms for the accident

previously evaluated are not affected and no additional failure

modes are created that could cause an accident that has been

previously evaluated. Since the diesel generator performance and

reliability will continue to be [ensured] by the maintenance rule,

the proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. This request does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

[These] proposed change[s] [do] not involve a change to the

plant design or operation. As a result, the proposed change[s] [do]

not affect any of the parameters or conditions that could contribute

to the initiation of any accidents. The proposed changes only affect

the methods used to monitor and [ensure] diesel generator

performance. The performance criteria for both the diesel generator

reliability and unavailability established by the maintenance rule,

and the causal determinations and corrective actions required for

maintenance preventable functional failures, [are] considered by

[the Nuclear Regulatory Commission (NRC) in] GL [(Generic Letter)]

94-01[, ``Removal of Accelerated Testing and Special Reporting

Requirements for Emergency Diesel Generators,'' issued May 31,

1994,] to be an acceptable method for monitoring diesel generator

performance.

No SSC, method of operation, or system interface is altered by

[these] change[s]. The changes do not impact the diesel

[generator]'s design sources, operating characteristics, system

functions, or system interrelationships. The failure mechanisms for

the accidents are not affected, and no additional failure modes are

created. Because the diesel generator performance and reliability

will continue to be [ensured] by the maintenance rule, the proposed

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

accident from any accident previously evaluated.

3. This request does not involve a significant reduction in a

margin [of] safety.

The proposed changes only affect the methods used to monitor and

[ensure] diesel generator performance and reliability. The

performance criteria for the diesel generator reliability and

availability established by the maintenance rule, and the causal

determinations and corrective actions required for maintenance

preventable functional failures, [are] considered by [NRC in] GL 94-

01 to be an acceptable method for monitoring diesel generator

performance. No margin [of] safety as defined in the bases for any

technical specification is impacted by these changes. [These]

change[s] [do] not impact any uncertainty in the design,

construction, or operation of any SSC. Diesel generator response to

accident initiators is unchanged. No SSC, method of operating, or

system interface is altered by [these] change[s]. The changes do not

impact the diesel [generator]'s design sources, operating

characteristics, system functions, or system interrelationships.

Because the diesel generator performance and reliability will

continue to be [ensured] by the maintenance rule, the proposed

changes do not involve a significant reduction in the margin [of]

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Judge George W. Armstrong

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

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: James W. Clifford, Acting

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: August 26, 1997

Description of amendment request: The proposed amendment would

revise the Crystal River Unit 3 (CR3) Technical Specifications Bases

(TSB) to change the design basis of the Emergency Diesel Generator

(EDG) Air Handling System. Specifically, TSB Sections B 3.8.1 and B

3.8.2 would be revised to indicate that a single or dual fan operation

depending upon fan supply air temperature, would maintain the

temperature of the EDG engine and control rooms within the EDG

manufacturer's limits.

Basis for proposed no significant hazardsconsideration

determination:

The EDG Air Handling System provides continuous ventilation, and

dissipates internal heat gains in the EDG engine and control rooms when

the diesel is operating. Presently, the CR3 plant documentation

requires operation of only one cooling fan per room to maintain the EDG

room temperature within the manufacturer's limit and is inconsistent

with the Final Safety Analysis Report (FSAR) which requires operation

of two fans.

As part of its EDG upgrade to increase their service ratings and

associated cooling analysis, the licensee has determined that operation

of either a single or dual cooling fans depending upon fan supply air

temperature, would achieve the required room cooling limits. The

licensee has determined that reliance on the operation of two cooling

fans instead of one involves an unreviewed safety question and requires

a license amendment.

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in

the probability of an accident previously evaluated. The EDG room

cooling fans support operation of the EDGs which are used to

mitigate design basis accidents. Although EDG availability is a

contributor to the risk of station blackout, the CR-3 licensing

basis assumes a station blackout without regard to EDG reliability.

Therefore, the probability of previously evaluated accidents is not

significantly increased.

For design basis accidents, the proposed change does not involve

a significant increase in the consequences of an accident previously

evaluated. The proposed change to operate both cooling fans for each

EDG to

[[Page 50005]]

provide adequate ventilation potentially increases the probability

of malfunction of equipment important to safety. However, the

proposed changes do not affect the independence of the EDGs or the

independence of the EDG Air Handling System and, based on single

failure criteria, one EDG will be fully operable and capable of

meeting its mission at all times as required by the CR-3 Technical

Specifications. Therefore, no significant increase in the

consequences of an accident previously evaluated, including the

offsite radiological dose exists.

Based on the above, the probability of an accident previously

evaluated has not been significantly increased, and this change does

not involve a significant increase in the consequences of an

accident previously evaluated.

2. Does not create the possibility of a new or different kind of

accident from any accident previously evaluated

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Neither the fans nor the EDGs are initiators of any new accidents.

The EDG room cooling fans support operation of the EDGs, which are

used to mitigate design basis accidents. Reliance on two fans rather

than one has reduced the redundancy of the EDG Air Handling System

and increased the probability of a malfunction of an EDG. However,

the proposed changes do not affect the independence of the EDGs or

the independence of the EDG Air Handling System and, based on single

failure criteria, one EDG will be fully operable and capable of

meeting its mission at all times as required by the CR-3 Technical

Specifications. Results of analyses to evaluate the failure of an

EDG to operate following a design basis accident are documented in

the FSAR. Therefore, this change does not create the possibility of

a new or different kind of accident.

3. Does not involve a significant reduction in the margin of

safety

The proposed change does not involve a significant reduction in

the margin of safety. The EDG room cooling fans support operation of

the EDGs. Following this change, two fans will be required to

maintain the EDG engine room and EDG control room temperatures

within the design basis limit when the fan supply air temperature is

greater than or equal to 85 deg.F. Reliance on two fans rather than

one has reduced the redundancy of the EDG Air Handling System and

slightly increased the probability of malfunction of an EDG, but

only after it has run for some period of time. However, the proposed

changes do not affect the independence of the EDGs or the

independence of the EDG Air Handling System and, based on single

failure criteria, one EDG will be fully operable and capable of

meeting its mission at all times as required by the CR-3 Technical

Specifications. Therefore, this change does not result in a

significant reduction to the margin safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J. Hebdon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: September 9, 1997

Description of amendment request: The proposed amendment would

revise the Crystal River 3 (CR3) Final Safety Analysis Report (FSAR) to

reflect the revised analysis for the hypothetical Makeup System Letdown

Line Failure Accident. In the original analysis, the event was modeled

as being terminated by an automatic isolation of the failed letdown

line on low reactor coolant system pressure. The revised analysis has

modeled the event as being terminated by manual operator action to

isolate the line. The licensee has determined that reliance on a manual

operator action in place of the automatic action involves an unreviewed

safety question (USQ) and requires prior Nuclear Regulatory Commission

(NRC) approval. Other FSAR changes are being proposed to clarify that

this accident is a hypothetical event that is presented only to

demonstrate that the dose consequences are below 10 CFR Part 100

limits. The licensee submitted its proposed FSAR changes which, upon

NRC approval, will be incorporated in the next revision to the FSAR.

Basis for proposed no significant hazardsconsideration

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change involves a revision to the analysis for the Makeup

System Letdown Line Failure Accident. The revised analysis assesses

the resultant change in consequences of this event based on the

actions specified in EOP-3 [Emergency Operating Procedure - 3] to

manually isolate the letdown line failure. No changes have been made

to any precursors to this event. Therefore, the probability of an

accident previously evaluated has not been increased.

This change has resulted in an increase in the calculated doses

due to the greater release of reactor coolant prior to termination

of the leak. Although the doses have increased, they remain

significantly less than the limits of 10 CFR 100. These doses also

remain lower than the resultant doses for the design basis LOCA

[loss-of-coolant-accident].

The revised analysis evaluates the consequences of this accident

based on the replacement of the automatic isolation of the letdown

line with a manual operator action to isolate the letdown line. This

action was added to EOP-3 when it was identified that the manual

initiation of the HPI [high pressure injection] system directed by

the EOP would interfere with the automatic isolation signal assumed

to terminate this event. Manual initiation of the HPI system for a

LSCM [loss of subcooling margin] event is consistent with the

symptomatic philosophy of the EOPs. This philosophy is utilized in

order to manage a wide range of event/leaks that would be indicated

by a LSCM. Early initiation of the HPI system is intended to ensure

adequate core cooling as the primary concern during a LSCM event.

Prior to the addition of the EOP step to manually isolate the

letdown line, the EOP directed actions towards locating and

isolating the source of the leak resulting in the LSCM. However, due

to the potential significance of the letdown line failure which can

result in RCS [reactor coolant system] leakage outside the reactor

building, the manual action was added early in EOP-3 to isolate the

letdown line. This action is proactive in ensuring early isolation

of the potential leakage path and is consistent with the concept of

a ``simple'' operator action (Reference 9) [NRC to Florida Power

Corporation letter, Long-term modifications regarding emergency core

cooling system Small Break Analysis problem, dated September 26,

1978].

Crediting a manual operator action instead of the automatic

isolation introduces the possibility of a malfunction of a different

type (i.e., operator error). The revised analysis assumes that

operator action to isolate the letdown line occurs 10 minutes

following a LSCM. Although the probability of operator error during

this action may be greater than the probability of the failure of

the automatic function, the consequences of this error would be

small. Several indications would be available to the operator to

identify the continued loss of coolant through this line. As

discussed above, the radiological dose calculated by this event

remains a small fraction of the limits of 10 CFR Part 100.

Therefore, adequate time would exist for the identification of an

operator error and correction of this error before any significant

increase in the consequences of this event would occur.

Additionally, the probability for operator error in this event

is considered to be small due to the extensive training plant

operators receive regarding the EOPs and the simple nature of the

action. Validation of the required actions in the EOPs, including

isolation of the letdown line, is performed on the plant simulator

to ensure the validity of the EOPs as well as to ensure that these

actions can be performed as required.

[[Page 50006]]

The clarification added to FSAR Section 5.4.4.2 and 14.2.2.6.1

reflects the previously approved evaluation for pipe rupture

criteria outside the reactor building for CR-3. A break in the high

energy portion of the letdown line outside containment is not

considered a credible event. This accident is presented only to

demonstrate that the dose consequences from a postulated break in

the letdown line outside containment remain below the 10 CFR Part

100 limits.

Based on the above, this change does not involve a significant

increase in the consequences of an accident previously evaluated.

2. Does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

This change does not involve any modification to the plant nor a

change in the operation of the plant prior to the postulated failure

of the letdown line. This change only evaluates the radiological

dose consequences of the actions taken following the line failure.

The addition of the action to manually isolate the letdown line for

a LSCM event is consistent with the need to isolate potential RCS

leakage paths and replaces the automatic isolation that was

previously assumed to occur. Therefore, this change does not create

the possibility of a new or different kind of accident.

3. Does not involve a significant reduction in the margin of

safety.

This change does not result in a reduction to the margin of

safety as defined in the Bases for any Technical Specifications. As

discussed above, the radiological doses for the revised analysis

have increased but remain a small fraction of the 10 CFR Part 100

limits.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J. Hebdon

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: August 22, 1997

Description of amendment request: The proposed amendment revises

Technical Specification (TS) 4.0.5, Surveillance Requirements for

Inservice Inspection and Testing of ASME Code Class 1, 2, and 3

components, to relocate the Inservice Testing Program requirements from

TS 4.0.5 to the Administrative Controls Section 6.8, Procedures and

Programs. The proposed amendment also provides conforming changes to

several Surveillance Requirements to change the reference from TS 4.0.5

to the Inservice Testing Program.

Basis for proposed no significant hazardsconsideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

There are no changes to the testing and evaluation related to pumps

and valves in the Inservice Testing Program. The only substantive

change allows the implementation of alternate testing provisions

where Code-requirements are impractical and the NRC has not formally

provided written approval. Since impractical testing would not be

performed in any event, the actual testing program is unaffected.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The use of the modified specifications cannot create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

implementation of this administrative change since the proposed

changes do not involve the addition or modification of equipment,

nor do they alter the design or operation of affected plant systems,

structures, or components.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components remain unchanged by the proposed

amendments, therefore, these changes do not involve a significant

reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420

NRC Project Director: Frederick J. Hebdon

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

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

Dates of amendment request: August 27, 1997

Description of amendment request: The licensee proposed modifying

the Turkey Point Units 3 and 4 Technical Specifications (TS) to delete

a sentence from section 6.2.2.f and add clarification to section

6.2.2.f of the Administrative section of TS to allow the use of up to

12 hour shifts without routine heavy use of overtime.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change does not involve a physical or procedural

change to any structure, system or component that significantly

effects the probability or consequences of any accident or

malfunction of equipment important to safety. The proposed changes

will allow the use of 12 hour shifts for a nominal 40 hours per

week.

This change is only administrative in nature and has no

significant impact on the probabilities or consequences of any

evaluated accident or malfunction of equipment important to safety.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or

modes of plant operation defined in the Turkey Point Units 3 and 4

operating license. The proposed amendment will not involve addition

or modification of permanent equipment for any systems structures or

components at Turkey Point.

The change does modify the controls on working shift hours for

operating personnel without significantly changing the hours worked

per week and retains the current limitations on excessive overtime.

The changes are administrative in nature.

Consequently, operation of either unit in accordance with the

proposed amendment would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

[[Page 50007]]

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

amendments would not involve a significant reduction in a margin of

safety.

The proposed amendment will allow the use of 12 hour shifts by

virtue of the administrative change. This will result in fewer

turnovers per day and will allow more contiguous days off between

work shifts. The sum of these 12 hour work shift features will be

more rested crews with better communications between shifts. The

proposed change will not alter the basis for any Technical

Specification that is related to the establishment of, or

maintenance of, a nuclear safety margin.

Consequently, operation of Turkey Point Units 3 and 4 in

accordance with this proposed amendment would 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 three standards of 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

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

NRC Project Director: Frederick J. Hebdon

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: August 18, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification 3.7.1.6, Atmospheric Steam Relief

Valves, to ensure the automatic feature of the steam generator power

operated relief valve remains operable during Modes 1 and 2. In

addition, the proposed change adds a surveillance requiring that a

channel calibration on the steam generator power operated relief valve

be performed every 18 months.

Basis for proposed no significant hazardsconsideration

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 South Texas Project proposed to revise Technical

Specification 3.7.1.6 to ensure the automatic feature of the Steam

Generator Power Operated Relief Valve remains operable during Modes

1 and 2. The South Texas Project has evaluated this proposed

amendment and determined that it involves no significant hazards

considerations based on the following:

A. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The methodologies used in the accident analyses remain

unchanged. The automatic actuation of the Steam Generator Power

Operated Relief Valves is not a new design feature. The effects of

the inadvertent opening of a Steam Generator Power Operated Relief

Valve are currently analyzed as described in Section 15.1.4 of the

Updated Final Safety Analysis Report. The radiological consequences

for the SBLOCA [small-break loss-of-coolant accident] event

presented in the Updated Final Safety Analysis Report remain

unchanged. The calculated Peak Clad Temperature remains

substantially below the 2200 deg.F acceptance limit of

10[]CFR[]50.46.

B. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The automatic actuation of the steam generator power operated

relief valves is not an accident initiator for the SBLOCA event. The

automatic actuation of the steam generator power operated relief

valves currently exists at the South Texas Project and is not a new

design feature. The description of the Steam Generator Power

Operated Relief Valves currently exists in the Updated Final Safety

Analysis Report. This change does not represent a change to the

facility and does not affect the safety functions and reliability of

systems, structures, or components in any new manner. Operating

procedures have a temporary administrative control to ensure the

automatic actuation of the Steam Generator Power Operated Relief

Valves remains operable in Modes 1 and 2. This condition will become

permanent with the approval of the Technical Specification Amendment

proposal.

C. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change results in the calculated Peak Clad

Temperature remaining well below the acceptance limit of

10[]CFR[]50.46 and comparable to the results currently described in

the Updated Final Safety Analysis Report.

Therefore, the South Texas Project has concluded that the

proposed change does not involve a significant hazards

considerations.

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.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: James W. Clifford, Acting

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

Millstone Nuclear Power Station, Unit No. 2, New London,

Connecticut

Date of amendment request: September 2, 1997

Description of amendment request: The proposed changes to the

Technical Specifications (TSs) would modify the maximum allowed

containment pressure specified in TS 3.6.1.4, ``Containment Systems

Internal Pressure,'' from 2.1 psig to 1.0 psig. The TS Bases, Section

3/4.6.1.4, would also be revised to reflect the new maximum allowed

containment pressure.

Basis for proposed no significant hazardsconsideration

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 involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will reduce the maximum allowed value for

containment pressure specified in Technical Specification 3.6.1.4,

``Containment Systems Internal Pressure.'' This change will improve

the margin between the peak containment pressure following a main

steam line break (most limiting accident for peak containment

pressure at Millstone Unit No. 2) and the containment design

pressure limit of 54 psig. Reducing the initial containment pressure

will result in a reduction in peak containment pressure.

To ensure the assumption of a lower initial containment pressure

is maintained, a change to Technical Specification 3.6.1.4 is

necessary.

The proposed change to Technical Specification 3.6.1.4 will

allow one of the initial assumptions used in the analysis for peak

containment pressure following a main steam line break to be

changed. However, this change will not affect how any of the plant

systems function to mitigate design basis accidents and will not

require any changes to mitigation procedures. The acceptance

criteria of a peak containment pressure less than the design limit

of 54 psig remains the same. Therefore, this change does not

significantly increase 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.

The proposed change does not alter the way any structure,

system, or component functions and does not alter the manner in

which the plant is operated. It does not

[[Page 50008]]

introduce any new failure modes and conservatively alters an

assumption made in the main steam line break safety analysis.

Therefore, the change will 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.

This proposed change will reduce the maximum allowed value for

containment pressure specified in Technical Specification 3.6.1.4,

``Containment Systems Internal Pressure.'' This change will improve

the margin between the peak containment pressure following a main

steam line break (most limiting accident for peak containment

pressure at Millstone Unit No. 2) and the containment design

pressure limit 54 psig. Starting at a lower initial containment

pressure will result in a lower peak containment pressure. To ensure

the assumption of a lower initial containment pressure is

maintained, a change to Technical Specification 3.6.1.4 is

necessary.

This more restrictive change in the maximum allowed containment

pressure will result in the use of a lower initial containment

pressure in the analysis of a main steam line break accident.

However, the analysis acceptance criteria of a peak accident

containment pressure less than 54 psig, will remain the same.

Therefore, there is no significant reduction in a margin of safety

as defined in the Bases of Technical Specification 3.6.1.4.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

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

Connecticut

Date of amendment request: September 2, 1997

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) to: (1) Combine TS 3.6.2.1,

``Containment Spray System,'' and TS 3.6.2.2, ``Containment Air

Recirculation System,'' into one specification which would reduce the

allowed outage time for one inoperable containment spray (CS) train or

one inoperable containment air recirculation (CAR) cooler from 30 days

to 7 days; increase the allowed outage time for two inoperable CAR

coolers from 48 hours to 7 days; add an allowed outage time of 48 hours

(instead of entering TS 3.0.3) for one inoperable CS train and two

inoperable CAR coolers or three or four inoperable CAR coolers; provide

specific guidance on when to enter TS 3.0.3; and expand the applicable

TS Bases to reflect these changes; (2) Modify the definition of

containment integrity and TS 3.6.1.1, ``Containment Integrity,'' to

indicate that the operability of the automatic isolation valve system

is satisfied by the use of the containment isolation trip push buttons

in Mode 4, and expand the TS Bases to reflect these change; (3) Add an

exception to the reactor coolant flow rate surveillance requirement, TS

4.1.1.3, whenever there is a reduction in reactor coolant system

boration while in Modes 2 and 3 because the reactor coolant pumps are

required to be in operation; (4) Delete the reactor coolant system

leakage surveillance requirements, TS 4.4.6.2.a and TS 4.4.6.2.b, which

require monitoring the containment atmosphere particulate radioactivity

and containment sump inventory, respectively; (5) Modify emergency core

cooling system surveillance requirement, TS 4.5.2.e, to allow the use

of alternative methods to verify that the throttle valves in Table 4.5-

1 are in the correct position and expand the TS bases to address the

alternative methods; (6) Modify TS 5.5.1,'' Emergency Core Cooling

Systems,'' by deleting the word ``original'' since the design has been

modified; and (7) Make editorial changes to terminology and item

numbering.

Basis for proposed no significant hazardsconsideration

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 changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1.Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to combine Technical Specifications 3.6.2.1

and 3.6.2.2 into one specification reduces the allowed outage time

for one inoperable containment spray (CS) train or one inoperable

containment air recirculation (CAR) cooler from 30 days to 7 days;

increases the allowed outage time for two inoperable CAR coolers

from 48 hours to 7 days; adds an allowed outage time of 48 hours

(instead of entering Technical Specification 3.0.3) for one

inoperable CS train and two inoperable CAR coolers, or three or four

inoperable CAR coolers; and provides specific guidance when it is

necessary to enter Technical Specification 3.0.3 will not affect how

these systems function to mitigate design basis accidents.

Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed changes to modify the definition of containment

integrity, modify the Technical Specification 3.6.1.1, ``Containment

Integrity,'' and expand the Bases to explain why automatic

containment isolation valves are operable in Mode 4 have no affect

on any containment isolation valve or Engineered Safety Feature

Actuation System (ESFAS) component. These components will still

function as designed to mitigate design basis accidents. Therefore,

this change does not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed change to provide an exception to Surveillance

Requirement 4.1.1.3 when the plant is in Modes 1 and 2 will not

result in any new approach to plant operation, it simply removes the

requirement to perform an unnecessary surveillance. The minimum

coolant flow through the core during a reduction in Reactor Coolant

System (RCS) boron concentration will still be met. Therefore, this

change does not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed change to delete Surveillance Requirements (SRs)

4.4.6.2.a and 4.4.6.2.b does not reduce the operability requirements

for any equipment used to monitor RCS leakage. The equipment covered

by these 2 SRs, containment atmosphere particulate radioactivity

monitors and containment sump inventory monitor, provide early

indication that RCS leakage exists, but do not provide the specific

information (amount of leakage) necessary to verify operation within

the leakage limits contained in Technical Specification 3.4.6.2,

``Reactor Coolant System Leakage.'' Operability of the containment

atmosphere particulate radioactivity monitors and containment sump

inventory monitor is verified by SRs 4.4.6.1.a and 4.4.6.1.b.

Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed change to Surveillance Requirement 4.5.2.e. to

allow the use of alternate methods does not reduce operability or

surveillance requirements for any of the Emergency Core Cooling

System (ECCS) throttle valves. Therefore, these ECCS throttle valves

will continue to function as designed to mitigate design basis

accidents. Therefore, this change does not significantly increase

the probability or consequences of an accident previously evaluated.

The proposed change to Technical Specification 5.5.1 has no

affect on how the ECCS operates. The ECCS will still function as

designed to mitigate design basis accidents. Therefore, this change

does not significantly increase the probability or consequences of

an accident previously evaluated.

[[Page 50009]]

The proposed changes to add information to the Bases of the

affected Technical Specifications, and make editorial changes to

terminology and item numbering will have no affect on equipment

operation. Therefore, all associated equipment will continue to

function as designed to mitigate design basis accidents. Therefore,

this change does not significantly increase the probability or

consequences of an accident previously evaluated.

Thus, this License Amendment Request does not impact the

probability of an accident previously evaluated nor does it involve

a significant increase in the consequences of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The proposed changes do not

introduce any new failure modes. They will not alter assumptions

made in the safety analysis and licensing basis. The affected

components and systems will still function as designed to mitigate

design basis accidents.

Therefore, these changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

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

The proposed changes will not reduce the margin of safety since

they have no impact on any safety analysis assumption. The proposed

changes do not decrease the scope of equipment currently required to

be operable or subject to surveillance testing, nor do the proposed

changes affect any instrument setpoints or equipment safety

functions. The requirement to check containment radiation and

containment sump level every 12 hours has been eliminated. However,

this equipment is still required to be operable, and the

surveillance requirements to verify operability have not been

changed. Therefore, this equipment will be available to provide

early indication of RCS leakage.

The effectiveness of Technical Specifications will be maintained

since the changes will not alter the operation of any component or

system. In addition, the changes are consistent with the new,

improved Standard Technical Specifications (STS) for Combustion

Engineering plants (NUREG-1432).

Therefore, there is not significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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

Millstone Nuclear Power Station, Unit No. 2, New London,

Connecticut

Date of amendment request: September 3, 1997

Description of amendment request: The proposed amendment would

revise the Updated Final Safety Analysis Report (UFSAR) by changing the

length of time the emergency diesel generators (EDGs) would operate

following a loss-of-coolant accident (LOCA) based on the capacity of

the onsite diesel fuel oil supply required by the current Technical

Specifications (TSs). The UFSAR indicates that the diesel fuel oil

supply tanks contain a sufficient amount of fuel to operate one EDG for

about 7 days and the other EDG 1 hour following a LOCA based on the TS

minimum limit of 24,000 gallons of diesel fuel oil stored onsite.

Northeast Nuclear Energy Company (the licensee) has performed

calculations indicating that both EDGs can initially operate, following

a LOCA, for 24 hours and one EDG can continue to operate for an

additional 3.5 days based on the TS requirement to have a minimum of

24,000 gallons of fuel oil stored onsite. The licensee has determined

that the difference in the EDGs operating time, as a result of the new

calculations, constitutes an unreviewed safety question and requests

approval to revise the UFSAR.

Specifically, the proposed license amendment would revise the

UFSAR, Section 8.3, ``Emergency Generators,'' to reflect the operating

times for the EDGs based on the TS-required onsite fuel oil supply.

Additional requirements would also be added indicating that the

existing nonsafety-related underground fuel oil storage tank would be

required to maintain about 17,700 gallons of fuel oil when the unit is

operating in Modes 1 through 4. This requirement would be included in

the Technical Requirements Manual, which also will require that the

amount of stored fuel oil be verified by surveillance requirements

similar to the TS-required surveillances for the safety-related fuel

oil supply. This change will increase the total time that one EDG can

continue to operate following a LOCA from 3.5 to 7 days. The Emergency

Plan (EP) procedures require that an evaluation be performed within 4

hours following a LOCA or loss of normal power (offsite power) to

determine if additional fuel oil is needed from an offsite source. The

licensee has a contract with a supplier for the delivery of fuel oil to

the Millstone site. The EP procedures also require that load shedding

recommendations be made within 24 hours. The recommendations will vary

depending on the situation and are another way to extend the operating

times for the EDGs.

Basis for proposed no significant hazardsconsideration

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 involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change expands FSAR Section 8.3, ``Emergency

Generators,'' to discuss the length of time the emergency diesel

generators (EDGs) will operate following a loss of coolant accident

(LOCA) and a loss of normal power (LNP), utilizing only onsite

diesel fuel oil sources. The onsite sources include the Technical

Specification required volume of 12,000 gallons in each diesel oil

supply tank and an additional approximate 17,700 gallons that will

be maintained in the underground diesel oil storage tank. This

onsite volume of diesel fuel oil is sufficient to allow two EDGs to

operate at rated load (2750 KW) for 24 hours following a design

basis LOCA and LNP. The remaining diesel fuel oil will be sufficient

for one EDG to continue operation at rated load for a total of 7

days from event initiation.

The proposed change to the FSAR has no effect on EDG operation

and reliability. The EDGs will continue to operate as designed to

supply the electrical loads assumed to mitigate the design basis

accidents. Therefore, there is no 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.

The proposed change will not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. Plant operating procedures will be

changed. However, the changes will not require the performance of

any task not currently performed by the plant operators. Emergency

Plan procedures already specify the action to provide load shedding

recommendations within 24 hours of a LOCA and LNP, and to evaluate

the need to order additional fuel from offsite sources within four

hours after the accident.

The proposed change does not alter the way any structure,

system, or component

[[Page 50010]]

functions and does not alter the manner in which the plant is

operated. It does not introduce any new failure modes and does not

alter assumptions made in the safety analysis.

Therefore, the change will 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 length of time the emergency diesel generators (EDGs) will

operate following a Loss of Coolant Accident and a Loss of Normal

Power, utilizing only the onsite diesel fuel oil sources required by

Technical Specifications has been recalculated. The new EDG run

times do not agree with the current EDG run times contained in the

Millstone Unit No. 2 Final Safety Analysis Report (FSAR), and

therefore do not agree with the current Technical Specification

Bases for 3.8.1.1, ``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C.

Sources - Shutdown.''

This deviation does result in a reduction in the margin of

safety as defined in the Technical Specification Bases for 3.8.1.1,

``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C. Sources -

Shutdown.'' However, this proposed change will require additional

diesel fuel oil to be maintained onsite in the non-seismic

underground diesel oil storage tank. This will ensure sufficient

diesel fuel oil will be maintained onsite to provide a 7 day supply,

assuming a seismic event does not occur. Therefore, this is not a

significant reduction in the margin of safety as defined in the

Technical Specification Bases for 3.8.1.1, ``A.C. Sources -

Operating,'' and 3.8.1.2, ``A.C. Sources - Shutdown.''

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: : Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: August 20, 1997

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) to provide for: (1) the

relocation of suppression pool volume references in Limiting Condition

for Operation (LCO) 3.5.3 to the Hope Creek (HC) Updated Final Safety

Analysis Report (UFSAR) and TS Bases as appropriate; (2) the revision

of the suppression pool volume currently listed in LCO 3.5.3.b; (3) the

relocation of the suppression pool volume references in LCO 3.6.2.1.a.1

to the UFSAR and TS Bases; and (4) the revision to the suppression pool

volume reference in TS 5.2.1 to reference the TS Bases section where

this information will reside.

Basis for proposed no significant hazardsconsideration

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 TS revisions involve: 1) no changes to the

operation of any systems or components in normal or accident

operating conditions; and 2) no significant changes to existing

structures, systems or components. The installation of the new

strainers will be justified separately using the provisions of

10CFR50.59. The relocation of Technical Specification references to

suppression pool volume to the UFSAR and/or TS Bases will not

adversely impact the safety-related functions of the suppression

pool or its supported systems since any changes to suppression pool

volume will be subject to 10CFR50.59 provisions. The impact of the

new strainers on ECCS [emergency core cooling system] performance in

Operational Conditions 4 and 5 has been determined to be negligible,

with less than a 0.3% decrease in suppression pool water volume at

the minimum specified suppression pool water level limit. In

addition, suppression pool volume is not a parameter involved in the

initiation of any accident. Therefore these changes will not

significantly increase the probability of an accident previously

evaluated. To the extent practicable, these proposed changes were

developed consistent with the changes approved by the NRC when

developing NUREG-1433, ``Standard Technical Specifications, General

Electric Plants, BWR/4'', with the intent of having the relocated

information controlled in other plant documents subject to

10CFR50.59 provisions. Since the plant systems associated with these

proposed changes will still be capable of: 1) meeting all applicable

design basis requirements; and 2) retain the capability to mitigate

the consequences of accidents described in the HC UFSAR, the

proposed changes were determined to be justified. Therefore, these

changes will not involve a significant increase in the 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.

Neither the relocation of Technical Specification references to

suppression pool volume nor the revision of the suppression pool

volume references for Operational Conditions 4 and 5 (COLD SHUTDOWN

and REFUELING) will adversely impact the operation of any safety

related component or equipment. Since the proposed changes involve:

1) no changes to the operation of any systems or components; and 2)

no significant changes to existing structures, systems or

components, there can be no impact on the occurrence of any

accident. To the extent practicable, these proposed changes were

developed consistent with the changes approved by the NRC when

developing NUREG-1433, ``Standard Technical Specifications, General

Electric Plants, BWR/4'', with the intent of having the relocated

information controlled in other plant documents subject to

10CFR50.59 provisions. Furthermore, there is no change in plant

testing proposed in this change request which could initiate an

event. Therefore, these changes will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Removal and relocation of the Technical Specification references

to suppression pool volume is consistent, to the extent practicable,

with the changes approved by the NRC when developing NUREG-1433,

``Standard Technical Specifications, General Electric Plants, BWR/

4''. The information retained in the Technical Specifications for

minimum suppression pool water level and the information retained in

the UFSAR and Technical Specification Bases will ensure that the

suppression pool and supported components will remain capable of

performing their intended safety functions. Any changes to

suppression pool volume information retained in the UFSAR or

Technical Specification Bases will be subject to the provisions of

10CFR50.59 and a separate safety evaluation would be developed to

support any proposed changes that would subsequently be made. The

impact of the new strainers on ECCS performance in Operational

Conditions 4 and 5 has been determined to be negligible, with less

than a 0.3% decrease in suppression pool water volume in the minimum

specified suppression pool water level limit. By retaining the 5

inch minimum suppression pool water level limit within the TS,

adequate provisions for: 1) NPSH [net-positive suction head] for

ECCS pump suction; 2) recirculation volume; and 3) vortex prevention

are maintained. Therefore, the changes contained in this request do

not result in a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

Unit - N21,

[[Page 50011]]

P.O. Box 236, Hancocks Bridge, NJ 08038

NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: August 19, 1997

Description of amendment request: The proposed amendment would

revise the Ginna Station Improved Technical Specifications (ITSs) by

adding a note to the Containment Spray (CS) Limiting Condition for

Operation (LCO) 3.6.6 which would allow the CS pumps in MODE 4 to be

placed in pull-stop, and motor-operated valves (MOVs) 896A and 896B to

have their DC control power restored with the valves placed in the

closed position in order to perform interlock and valve testing of MOVs

857A, 857B, and 857C. A time limit of 2 hours is placed on this

configuration for each test.

Basis for proposed no significant hazardsconsideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. The change is

to add a note to LCO 3.6.6 which allows the CS pumps to be placed in

pull-stop and MOVs 896A and 896B to have power restored and closed

in MODE 4. This does not increase the probability of any accident

previously evaluated since the CS system provides mitigation

capability only (i.e., does not initiate any accident). In addition,

there is no design basis accident previously evaluated in MODE 4

which would require the use of CS. Therefore, these changes do not

involve a significant increase in the probability or consequences of

an accident previously analyzed.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

changes do not involve a physical alteration of the plant (i.e., no

new or different type of equipment will be installed) or changes in

the methods governing normal plant operation. Thus, this change does

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The proposed changes will not reduce a margin of plant

safety because the CS function is not required for any design basis

accident in MODE 4. In addition, time restraints [are] placed on the

proposed plant configuration. As such, no question of safety is

involved, and the change does not involve a significant reduction in

a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005

NRC Project Director: Alexander W. Dromerick, Acting Director

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

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: August 2, 1996 (TXX-96434)

Brief description of amendments: The proposed changes would

increase the allowed outage time (AOT) for a centrifugal charging pump

from 72 hours to 7 days.

Basis for proposed no significant hazardsconsideration

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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

There is no effect on the probability of an event; the only

potential effect is on the capability to mitigate the event. The

centrifugal charging pumps are credited in the Final Safety Analysis

Report Chapter 15 LOCA analysis for ECCS injection and for the

containment sump recirculation mode for the design-basis LOCA.

Increasing the AOT for the centrifugal charging pumps does not

affect analysis assumptions regarding functioning of required

equipment designed to mitigate the consequences of accidents.

Further, the severity of postulated accidents and resulting

radiological effluent releases will not be affected by the increased

AOT.

A reliability analysis of the charging system found the change

to have no significant impact on normal operation or on the RCP seal

cooling function. Therefore, the change would not significantly

increase in the probability of a seal LOCA.

The change potentially affects only the availability of the

charging system for accident mitigation and has no effect on the

ability of other ECCS systems to perform their functions. Through

the use of a probabilistic risk assessment, it was determined that

the proposed change would have an insignificant effect on the core

damage frequency.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Do the proposed changes create the possibility of a new or

different type of accident from any accident previously evaluated?

Unavailability of one centrifugal charging pump for a finite

period of time is currently allowed by the Technical Specifications.

Increasing the AOT from 72 hours to 7 days would not change the

method that TU Electric operates CPSES, thus would not create a new

condition. Further, the proposed change would not result in any

physical alteration to any plant system, and there would not be a

change in the method by which any safety related system performs its

function. The ECCS would still be capable of mitigating the

consequences of the design-basis accident LOCA with the one

centrifugal charging pump operable. No new unanalyzed accident would

be created.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

The proposed change does not impact either the physical

protective boundaries or performance of safety systems for accident

mitigation. There is no safety analysis impact since the extension

of the centrifugal charging pump AOT interval will have no effect on

any safety limit, protection system setpoint, or limiting condition

of operation. There is no hardware change that would impact existing

safety analysis acceptance criteria, therefore there is no

significant change in the margin of safety.

In summary, the proposed change would not have a significant

impact on 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: University of Texas at

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

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, N.W., Washington, DC 20036

NRC Project Director: James W. Clifford, Acting

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the

[[Page 50012]]

action involved exigent circumstances. They are repeated here because

the biweekly notice lists all amendments issued or proposed to be

issued involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

Michigan

Date of amendment request: September 5, 1997 (NRC-97-0107)

Description of amendment request: The proposed amendment would add

Special Test Exception 3/4.10.7, ``Inservice Leak and Hydrostatic

Testing,'' that allows the performance of pressure testing at a reactor

coolant temperature up to 212 deg.F while remaining in Operational

Condition 4. This special test exception would also require that

certain Operational Condition 3 specifications for Secondary

Containment Isolation, Secondary Containment Integrity, Secondary

Containment Automatic Isolation Dampers, and Standby Gas Treatment

System operability be met. This change would also revise the Index,

Table 1.2, ``Operational Conditions,'' and the Bases to incorporate the

reference to the proposed special test exception. The licensee

requested that this amendment be reviewed under exigent circumstances.

Date of individual notice in the Federal Register: September 12,

1997 (62 FR 48113)

Expiration date of individual notice: October 14, 1997 NSHC

comments: September 29, 1997

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: John N. Hannon

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: August 14, 1997

Brief description of amendment request: The proposed amendments

would revise the allowed tolerance of the reactor coolant system volume

provided in Technical Specification 5.4.2 to account for steam

generator tube plugging.

Date of individual notice in the Federal Register: August 26, 1997

(62 FR 45278)

Expiration date of individual notice: September 25, 1997

Local Public Document Room location: Wharton County Junior College,

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

Notice Of Issuance Of Amendments ToFacility 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.

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 application for amendments: January 15, 1997, as

supplemented on August 22, 1997.

Brief description of amendments: The amendments revise the minimum

and maximum allowed values in Technical Specification 3.6.2.1 for

suppression chamber water volume. The amendments correct an error

identified by Carolina Power & Light Company in the previous

calculation of water volume and correct an error in the value listed in

the associated TS Bases for Unit 1 for primary system operating

pressure.

Date of issuance: August 28, 1997

Effective date: August 28, 1997

Amendment Nos.: 186 and 217

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

change the Technical Specifications

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

14458) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 28, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

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 application for amendments: December 4, 1996

Brief description of amendments: The amendments revise the approach

in Technical Specification 3/4.1.2 for determining a reactivity anomaly

by changing from control rod density comparison to direct comparison of

reactivity status.

Date of issuance: September 5, 1997

Effective date: September 5, 1997

Amendment Nos.: 187 and 218

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

change the Technical Specifications

Date of initial notice in Federal Register: March 12, 1997 (62 FR

11484) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 5, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at

[[Page 50013]]

Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

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: January 24, 1997

Brief description of amendments: The amendments revise the

Technical Specification (TS) required surveillance calibration to be

performed on the reactor water level instrumentation to reflect the

modifications made to the Unit 3 instrumentation. The modifications

were made during the recent Unit 3 refueling outage to improve the

reliability of emergency core cooling system (ECCS) initiation on low

low reactor water level. The surveillance requirement for calibration

of the new level instrumentation is consistent with the ECCS low

reactor water level initiation transmitter calibration requirements of

NUREG 1433, ``Standard Technical Specifications, General Electric

Plants, BWR/4'' for similar instrumentation. The same TS change for

Unit 2 has been previously reviewed and approved by the NRC staff in

Amendment No. 145 dated June 28, 1996. In addition minor editorial

changes were made to the TS.

Date of issuance: September 10, 1997

Effective date: September 10, 1997, with full implementation within

60 days.

Amendment Nos.: 162 and 157

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 18, 1997 (62 FR

19143) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 10, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450

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

Michigan

application for amendment: December 15, 1994, as revised July 25,

1996, and supplemented December 13, 1996, and June 18, 1997

Brief description of amendment: The amendment revises Technical

Specification (TS) Section 6.0, Administrative Controls, by (1)

removing requirements that are adequately controlled by existing

regulations other than 10 CFR 50.36 and the TS and (2) relocating

selected requirements from TS Section 6.0 to licensee-controlled

documents or programs.

Date of issuance: September 10, 1997

Effective date: September 10, 1997, with full implementation within

90 days. Implementation of this amendment shall include the relocation

of the TS requirements to the appropriate licensee-controlled

documents, as described in the licensee's application dated December

15, 1994, as revised July 25, 1996, and supplemented December 13, 1996,

and June 18, 1997, and evaluated in the staff's safety evaluation dated

September 10, 1997.

Amendment No.: 113

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

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

29873) and August 14, 1996 (61 FR 42279). The December 13, 1996, and

June 18, 1997, letters provided clarifying information within the scope

of the original application and did not change the staff's initial

proposed no significant hazards considerations determination. The

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

Safety Evaluation dated September 10, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley

Power Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: March 10, 1997

Brief description of amendment: The amendment modifies the

Technical Specifications (TSs) by reducing the reactor coolant system

specific activity limits in accordance with the NRC's guidance provided

in Generic Letter 95-05, ``Voltage-Based Repair Criteria for

Westinghouse Steam Generator Tubes by Outside Diameter Stress Corrosion

Cracking.'' The definition of DOSE EQUIVALENT I-131 is replaced with

the Improved Standard TS definition in the first sentence and an

equation is added based on dose conversion factors derived from the

International Commission on Radiation Protection (ICRP) ICRP-30. TS

3.4.8, Specific Activity, is revised by reducing the DOSE EQUIVALENT I-

131 limit from 1.0 micro Ci/gram to 0.35 micro Ci/gram for the 48-hour

limit and from 60 micro Ci/gram to 21 micro Ci/gram for the maximum

instantaneous limit. Item 4.a in TS Table 4.4-12, Primary Coolant

Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the

Bases for TS 3/4.4.8 are also modified to reflect the reduced DOSE

EQUIVALENT I-131 limit.

Date of issuance: September 10, 1997

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

within 60 days.

Amendment No: 205

Facility Operating License No. DPR-66. Amendment revised the

Technical Specifications.

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

24985) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 10, 1997. No

significant hazards consideration comments received: No

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: August 4, 1997, as supplemented

August 16, 1997.

Brief description of amendment: Temporary change to Technical

Specification Surveillance Requirement (SR) 3.3.8.1. The change will

allow the licensee to extend the frequency of SR 3.3.8.1 from 31 to 60

days.

Date of issuance: August 29, 1997

Effective date: August 29, 1997

Amendment No.: 157

Facility Operating License No. DPR-72. Amendment temporarily

revises Technical Specifications Surveillance Requirement 3.3.8.1.

Date of initial notice in Federal Register: August 12, 1997 (62 FR

43189) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 29, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

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: May 17, 1996, as supplemented June 14,

1996, March 17, July 29, and July 30, 1997

Brief description of amendments: The amendments modify Technical

[[Page 50014]]

Specification Section 3/4.4.5 Steam Generators, 3/4.4.6 Reactor Coolant

System Leakage, and associated Bases to allow the installation of tube

sleeves as an alternative to plugging to repair defective steam

generator tubes.

Date of issuance: September 4, 1997

Effective date: September 4, 1997

Amendment Nos.: Unit 1 - Amendment No. 90; Unit 2 - Amendment No.

77

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 29, 1996 (61 FR

25938) and April 9, 1997 (62 FR 17235). The June 14, 1996, and July 29,

and July 30, 1997, submittals provided additional information that did

not affect the staff's initial no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 4, 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. 50-423,

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

Connecticut

Date of application for amendment: June 30, 1997

Brief description of amendment: Technical Specification

Surveillance Requirements 4.7.1.5.1 and 4.7.1.5.2 require the periodic

testing of the main steam isolation valves (MSIVs) to demonstrate

operability. The amendment (1) clarifies when the MSIVs are partial

stroked or full closure tested, (2) adds a note to the Mode 4

applicability of Technical Specification 3.7.1.5 to require that the

MSIVs be closed and deactivated at less than 320 degrees F, (3) makes

editorial changes, and (4) makes changes to the associated Bases

sections.

Date of issuance: September 3, 1997

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

within 60 days.

Amendment No.: 148

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

Technical Specifications.

Date of initial notice in Federal Register: July 30, 1997 ( 62 FR

40853) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 3, 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 14, 1997, as supplemented by

letter dated July 30, 1997

Brief description of amendment: Technical Specification

Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger

be tested to verify that it can supply a specified current at 125

volts. The amendment increases the required test voltage.

Date of issuance: September 5, 1997

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

within 60 days.

Amendment No.: 149

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

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33130) The July 30, 1997, letter provided clarifying information that

did not change the scope of the May 14, 1997, application and the

initial proposed no significant hazards consideration determination.

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

Safety Evaluation dated September 5, 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

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: May 9, 1997, as supplemented by

letter dated July 14, 1997

Brief description of amendments: The proposed change revises the

Peach Bottom Atomic Power Station, Units 2 and 3, technical

specifications to extend the interval for replacing the primary

containment purge and exhaust valve inflatable seals.

Date of issuance: September 4, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendments Nos.: 220 and 223

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

revised the Technical Specifications.

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

35851) The supplemental letter provided clarifying information that did

not change the original no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 4, 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

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-

388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne

County, Pennsylvania

Date of application for amendments: September 25, 1996

Brief description of amendments: These amendments (1) revise the

required number of operable gaseous radioactivity monitoring system

channels and particulate radioactivity monitoring system channels from

one in each of the monitoring systems to one in either of the

monitoring systems, (2) allow both the gaseous radioactivity monitoring

system and the particulate monitoring system to be inoperable for up to

30 days provided that grab samples are obtained and analyzed at least

once per 12 hours, and (3) add an action for the loss of all reactor

coolant system leakage detection systems (drywell floor sump level

monitoring system, gaseous radioactivity monitoring system and

particulate radioactivity monitoring system).

Date of issuance: September 3, 1997

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

within 30 days of issuance.

Amendment Nos.: 168 and 142

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 19, 1996 (61

FR 58904) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated

[[Page 50015]]

September 3, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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 application for amendments: April 25, 1997, as supplemented

June 6, 1997

Brief description of amendments: The amendments revise Technical

Specification 3.5.2 to eliminate reference to the flow path from the

residual heat removal system to the reactor coolant system hot legs.

This flow path is being eliminated to prevent excessive flow through

the residual heat removal system during all hot leg recirculation

configurations assuming worst-case single failures that could result in

excessive flow during hot leg recirculation following a loss-of-coolant

accident.

Date of issuance: September 11, 1997

Effective date: Both units, as of the date of issuance, to be

implemented within 60 days of issuance.

Amendment Nos.: 200 and 184

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 14, 1997 (62 FR

26574) The June 6, 1997, supplement 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 September 11,

1997. No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079Sacramento Municipal Utility District,

Docket No. 312, Rancho Seco Nuclear Generating Station, Sacramento

County, California

Date of application for amendment: December 9, 1993, as superseded

December 19, 1995, and as supplemented on January 22, 1996.

Brief description of amendment: This amendment changes the

Technical Specifications to incorporate the revised 10 CFR Part 20,

Standards for Protection Against Radiation. The amendment corrects

references from Semiannual Radioactive Effluent Release Report to

Annual Radioactive Effluent Release Report. The amendment also corrects

references from NRC Region V to NRC Region IV.

Date of issuance: August 22, 1997

Effective date: August 22, 1997

Amendment No.: 125

Facility Operating License No. NPF-1: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 2, 1994 (59 FR

10015) The information provided in the licensee's letters of December

19, 1995 and January 22, 1996 contained editorial changes and did not

involve significant changes to the original Federal Register notice.

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

Safety Evaluation dated August 22, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: : Central Library, Government

Documents, 828 I Street, Sacramento, California 95814

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

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366,

Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,

Georgia Date of application for amendments: January 7, 1997, as

supplemented July 2, 1997

Brief description of amendments: The amendments revise plant

Technical Specifications associated with surveillance requirements

testing that requires manually actuating every safety/relief valve

during each unit startup from a refueling outage.

Date of issuance: September 5, 1997

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

30 days

Amendment Nos.: 208 and 150

Facility Operating License Nos. DPR-57 and NPF-5: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4350) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 5, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

Tennessee Valley Authority, Docket No. 50-260 Browns Ferry Nuclear

Plant, Unit 2, Limestone County, Alabama

Date of application for amendment: June 2, 1995, revised March 3,

1997, as supplemented May 13 and August 20, 1997 (TS 353)

Brief description of amendment: The amendment provides technical

specification (TS) changes for an upgrade of the power range neutron

monitor instrumentation. Changes to thermal limits specifications were

also proposed to implement average power range monitor and rod block

monitor ts improvements, and maximum extended load line limit analyses.

Date of issuance: September 11, 1997

Effective Date: September 11, 1997

Amendment No.: 249

Facility Operating License No. DPR-52: Amendment revised the TS.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42609) The March 3, 1997 revision, as supplemented May 13 and August

20, 1997, does not affect the staff's proposed finding of no

significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 11, 1997. No significant hazards

consideration comments received: None.

Local Public Document Room location: Athens Public library, 405 E.

South Street, Athens, Alabama 35611

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

Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: March 27, 1997, as supplemented

May 28, June 4, and July 30, 1997.

Brief description of amendment: The amendment pertains to Cycle 2

core design changes and provides operational enhancements for reactor

trip setpoints. Part 1 addresses an increase in the containment sump

boron concentration during a large break loss-of-coolant accident and

describes changes to Technical Specification (TS) 3.5.1 and 3.5.4

regarding boron concentration. Part 2 addresses changes to TS Figure

2.1.1-1, TS Table 3.3.1-1, and TS 3.4.1 on safety limits, the trip

system and pressure, temperature and flow limits, respectively.

Date of issuance: September 11, 1997

Effective date: Sepember 11, 1997

Amendment No.: 7

Facility Operating License No. NPF-90: Amendment revises the TS.

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

35852) The July 30, 1997 submittal provided clarifying information

which did not affect the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated September 11, 1997. No significant hazards

consideration comments received: None

[[Page 50016]]

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,

Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No.

50-440 Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: January 31, 1997, supplemented

August 6, 1997.

Brief description of amendment: The amendment approves the use of

Option B, ``Performance-Based Requirements,'' to 10 CFR Part 50,

Appendix J, ``Primary Reactor Containment Leakage Testing for Water-

Cooled Power Reactors.''

Date of issuance: September 9, 1997

Effective date: September 9, 1997

Amendment No.: 86

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 12, 1997 (62 FR

11492). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 9, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,

Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No.

50-440 Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: May 2, 1997

Brief description of amendment: The amendment allows the leakage

rate of one or more main steam lines to be up to 35 standard cubic feet

per hour (scfh), as long as the total leakage rate through all four

main steam lines is less than or equal to 100 scfh.

Date of issuance: September 11, 1997

Effective date: September 11, 1997

Amendment No.: 87

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33136). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 11, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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 198)

Brief description of amendments: These amendments revise Technical

Specification Section 15.3.1, ``Reactor Coolant System,'' to eliminate

the provisions for operation of the units at below 3.5 percent rated

power with a single reactor coolant pump.

Date of issuance: September 3, 1997

Effective date: September 3, 1997, with full implementation within

45 days

Amendment Nos.: 178 and 182

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) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 3, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241

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: January 24, 1997, as

supplemented on May 15 and August 5, 1997 (TSCR 193)

Brief description of amendments: These amendments revise TS 15.5.4,

``Fuel Storage,'' to increase fuel assembly enrichment limits to 5.0

weight percent uranium-235 while maintaining Keff in the

storage pools (spent fuel pool and new fuel storage racks) less than

0.95. Date of issuance: September 4, 1997

Effective date: September 4, 1997, with full implementation within

45 days

Amendment Nos.: 179 and 183

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

revised the Technical Specifications.

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

30647) The August 5, 1997, submittal provided clarifying information

within the scope of the original application and did not affect 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 September 4, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the

[[Page 50017]]

plant's licensed power level, the Commission may not have had an

opportunity to provide for public comment on its no significant hazards

consideration determination. In such case, the license amendment has

been issued without opportunity for comment. If there has been some

time for public comment but less than 30 days, the Commission may

provide an opportunity for public comment. If comments have been

requested, it is so stated. In either event, the State has been

consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective 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

application for amendment, (2) the amendment to Facility Operating

License, 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 room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By October 24, 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. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 the

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

[[Page 50018]]

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo

Verde Nuclear Generating Station, Unit No. 2, Maricopa County,

Arizona

Date of application for amendment: August 28, 1997, as supplemented

by letter dated September 3, 1997.

Brief description of amendment: The amendment revises Technical

Specification Table 4.3-2 to allow for a one-time, five-day extension

of the required surveillance interval for the main steam isolation

system portion of the engineered safety feature actuation system logic.

Date of issuance: September 4, 1997

Effective date: September 4, 1997

Amendment No.: 105

Facility Operating License No. NPF-51: The amendment revised the

Technical Specifications. Press release issued requesting comments as

to proposed no significant hazards consideration: Yes. September 1,

1997. Arizona Republic Newspaper (Arizona). Comments received: No. The

Commission's related evaluation of the amendment, finding of exigent

circumstances, consultation with the State of Arizona and final

determination of no significant hazards consideration are contained in

a Safety Evaluation dated September 4, 1997.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: William H. Bateman

Public Service Electric & Gas Company, Docket No. 50-311, Salem

Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

Date of application for amendment: August 19, 1997, as supplemented

August 20, 1997.

Brief description of amendment: This amendment to the Technical

Specifications increases the allowable band for control and shutdown

rod demanded position versus indication position from plus or minus 12

steps to plus or minus 18 steps when the power level is not greater

than 85% rated thermal power.

Date of issuance: September 10, 1997

Effective date: As of date of issuance, to be implemented within 7

days.

Amendment No. 183

Facility Operating License No. DPR-75: This amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration, and requested that any comments on

the proposed no significant hazards consideration be provided to the

staff by the close of business on September 3, 1997, and stated that,

should circumstances change during the notice period, such that a

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 notice period, provided that its

final determination is that the amendment involves no significant

hazards consideration. The notice was published in the Wilmington News

Journal on August 22, 1997, and in Today's Sunbeam on August 24, 1997.

No public comments were received. The Commission's related evaluation

of the amendment, finding of exigent circumstances, consultation with

the State of New Jersey and final no significant hazards consideration

determination are contained in a Safety Evaluation dated September 10,

1997.

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

Dated at Rockville, Maryland, this 17th day of September 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation

[Doc. 97-25210 Filed 9-23-97; 8:45 am]

BILLING CODE 7590-01-F