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

[[Page 37296]]

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 June 22, 1996, through July 5, 1996. The

last biweekly notice was published on July 3, 1996 (61 FR 34884).

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

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By August 16, 1996, 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,

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notwithstanding the request for a hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

Services Branch, or may be delivered to the Commission's Public

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

by the above date. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to 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.

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 amendments request: April 4, 1996

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TS) to add an allowance to

complete a TS-required surveillance within 24 hours of discovery of a

missed surveillance in accordance with the guidance of Generic Letter

(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical

Specifications (STS) on the Applicability of Limiting Conditions for

Operation and Surveillance Requirements'' and NUREG-1433, ``Standard

Technical Specifications, General Electric Plants, BWR/4,'' Revision 1,

April 1995. Typographical errors are being corrected and wording

adjustments are being incorporated for consistency between plant TS

terminology and the associated Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

. The proposed amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The operational flexibility resulting from the proposed

revision to Technical Specification 3.0.4 is consistent with that

allowed by the existing individual LCO [limiting condition for

operation] and their associated ACTION requirements, which provide

an acceptable level of safety for continued operation. A delay of up

to 24 hours or the time of the surveillance interval, whichever is

less, provided by Technical Specification 4.0.3 to complete a missed

surveillance reduces the probability of a transient occurring when

the affected system or component is either out of service to allow

performance of the surveillance test, or there is a lower level of

confidence in the operability because the normal surveillance was

exceeded. The revision to Technical Specification 4.0.4 makes it

clear that Technical Specification 4.0.4 does not prevent passage

through or to OPERATIONAL CONDITIONS as required to comply with

ACTION requirements. The revision to the wording in Unit 2 Technical

Specification Table 3.12.1-1, Notation (h), revisions to the Bases

of the Technical Specifications, and the elimination of specific

exemptions to Technical Specifications 3.0.4 are administrative in

nature.

Based on the above, the proposed license amendments do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. The proposed amendments do not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The proposed license amendments do not introduce any new

equipment nor do they require any existing equipment or systems to

perform a different type of function than they are presently

designed to perform. The proposed changes result in improved

Technical Specifications by removing unnecessary restrictions on

changes in OPERATIONAL CONDITIONS and facility operation, removing

unnecessary shutdowns caused by inadvertently exceeding surveillance

intervals, and removing conflicts between various Technical

Specifications. The revision to the wording in Unit 2 Technical

Specification Table 3.12.1-1, Notation (h), revisions to the Bases

of the Technical Specifications, and the elimination of specific

exemptions to Technical Specification 3.0.4 are administrative in

nature.

Based on the above, the proposed license amendments do not

create a new or different kind of accident from any previously

evaluated.

3. The proposed license amendments do not involve a significant

reduction in a margin of safety. The operational flexibility that

results from the proposed revision to Technical Specification 3.0.4

is consistent with that allowed by the existing individual LCO and

associated ACTION requirements, which provide an acceptable level of

safety for continued operation. Therefore, there is no change in the

margin of safety associated with this change. A delay of up to 24

hours or the length of the surveillance interval, whichever is less,

provided by Technical Specification 4.0.3 to complete a missed

surveillance reduces the probability of a transient occurring when

the affected system or component is either out of service to allow

performance of the surveillance test, or there is a lower level of

confidence in the operability because the normal surveillance was

exceeded. In addition, the proposed change acknowledges that the

most common outcome of the performance of a surveillance is the

successful demonstration that acceptance criteria are met. The

proposed change provides the potential benefit of avoiding a

shutdown transient when required equipment is still capable of

performing its function, and variables are still within limits. The

revision to Technical Specification 4.0.4 makes it clear that

Technical Specification 4.0.4 does not prevent passage through or to

OPERATIONAL CONDITIONS as required to comply with ACTION

requirements. This change is considered to be a clarification to

achieve consistency with existing Technical Specification

requirements. The revision to the wording in Unit 2 Technical

Specification Table 3.12.1-1, Notation (h), revisions to the Bases

of the Technical Specifications, and the elimination of specific

exemptions to Technical Specification 3.0.4 are administrative in

nature.

The proposed changes would result in improved Technical

Specifications and eliminate unnecessary plant challenges. Based on

the above, the proposed license amendments 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: University of North Carolina

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

Wilmington, North Carolina 28403-3297

[[Page 37298]]

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

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: December 6, 1995

Description of amendment request: The proposed amendment would

relocate the crane operation and movement of heavy loads requirements

and their bases from the Technical Specifications (TS) to other plant

documents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The change moves the requirements from TS to other plant

documents controlled under 10 CFR 50.59 without affecting their

technical content. Since this change does not alter the technical

content of any requirements, the operation of the facility in

accordance with the proposed change cannot involve a significant

increase in the probability or consequences of an accident

previously evaluated, create the possibility of a new or different

kind of accident from any previous evaluated, or 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: Mark Reinhart

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: June 6, 1996

Description of amendment request: The proposed change would remove

the Engineered Safeguard (ES) signals that presently open the outlet

valves on the Low Pressure Service Water (LPSW) System coolers, LPSW-4

and LPSW-5, on high reactor coolant system pressure or high reactor

building pressure. The valves will continue to be operable from the

control room when needed. The proposed change to Technical

Specification (TS) 4.5.1.1.2.a.(2) would require that the refueling

outage test signal be applied to the LPSW pumps, but no longer to LPSW-

4 and LPSW-5, and that the operability of the valves be verified by

cycling them from the control room. A note would be added to reflect

that the refueling outage test of LPSW-4 and LPSW-5 response to the ES

signal will continue to be verified until the signal is removed from

the ES system for each unit during the specified refueling outages. In

addition, TS 4.5.1.1.2.b would be clarified to differentiate between

test acceptance criteria for automatic actuation of the appropriate

LPSW pumps and valves in response to the ES signal, and completion of

travel of LPSW-4 and LPSW-5 in response to manual operation of the

valves. A proposed change to the Bases would also reflect these

changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10CFR50.91, Duke Power Company (Duke) has made the

determination that this amendment involves a No Significant Hazards

Consideration by applying the standards established by NRC

regulations in 10CFR50.92. The following discusses the basis for our

analysis:

Will operation of the facility in accordance with the proposed

amendment:

A. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. Eliminating the automatic signal that opens Low Pressure

Service Water (LPSW) System valves, LPSW-4 and LPSW-5, upon an

Engineered Safeguards (ES) actuation does not increase the

probability of any accident previously evaluated. The proposed

change would involve a delay in providing cooling water to the Low

Pressure Injection (LPI) System coolers after a design basis

accident. Cooling water flow to the LPI coolers is isolated during

normal power operation. During normal cold shutdown conditions,

cooling water flow to the LPI coolers is normally open without

relying on the ES actuation signal. This cooling water flow is

needed to mitigate certain accidents, but a delay in providing this

cooling water flow after a design basis accident does not

significantly increase the probability of any accident previously

evaluated.

Eliminating the ES actuation signal for LPSW-4 and LPSW-5 will

not increase the consequences of an accident previously evaluated.

After a loss of coolant accident (LOCA), operators will operate the

appropriate valves from the control room in sufficient time to

provide adequate cooling water flow to maintain containment

temperature and pressure within acceptable limits. Duke has also

evaluated the delay of LPSW cooling flow's impact on core cooling

and concluded that there are no adverse impacts on the capability to

maintain core cooling. Since the containment temperature and

pressure limits after a LOCA will not be exceeded, this change will

not increase any potential off-site dose consequences after a LOCA.

Due to the time available for operator action (approximately one

hour), there is no significant increase in operator burden during

this accident scenario.

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

from the accidents previously evaluated?

No. As stated above, due to the time available for operator

action (approximately 1 hour), there is no significant increase in

operator burden during this accident scenario. Eliminating the ES

signal that automatically opens valves LPSW-4 and LPSW-5 results in

significantly lower flow demand on the LPSW pumps. If all LPSW pumps

are successfully started, this could result in a stronger pump

causing deadhead conditions on a weaker pump since the pumps feed

into the same piping system. To prevent any potential adverse

effects on the LPSW pumps due to inadequate flow during the initial

stages of a LOCA, minimum flow piping will be installed for the LPSW

pumps to provide adequate flowpaths for pump minimum flow. Testing

will be performed to validate that the LPSW pumps can operate at the

chosen design value for pump minimum flow. In addition, Duke

conducted an evaluation, based on manufacturer input, of the thermal

effects on the LPI coolers due to delaying LPSW cooling flow. This

evaluation concluded that the 30 minute delay of LPSW cooling flow

has no adverse thermal effects on the LPI coolers. Therefore,

because there is no significant increase in operator burden and

because there will be no adverse effects on the LPSW pumps, LPI

coolers, and associated piping caused by the delayed LPSW cooling

flow, the proposed change will not create the possibility of a new

or different kind of accident from the accidents previously

evaluated.

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

No. There are no safety limits or limiting safety system

settings associated with the LPSW System in the Oconee Nuclear

Station Technical Specifications. The proposed change will not

affect any existing safety limits or limiting safety system

settings. The proposed change will not affect any existing Limiting

Conditions for Operation in the Technical Specifications. The

proposed change involves an alternative method of initiating cooling

water flow to the LPI coolers after a LOCA. This alternative method

will achieve the required results since there will be no significant

change in the containment temperature and pressure after a LOCA.

Duke 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

[[Page 37299]]

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 (GGNS), Unit 1, Claiborne County, Mississippi

Date of amendment request: June 20, 1996

Description of amendment request: The amendment would redefine the

secondary containment boundary to allow the enclosure building to be

inoperable during the upcoming refueling outage 8 (RFO 8) scheduled to

begin in October 1996. The amendment would add a condition to the

license that the enclosure building may be inoperable during core

alterations and movement of non-recently irradiated fuel (i.e., fuel

that has not occupied part of a critical reactor core for 12 days)

during RFO 8 and the standby gas treatment (SGT) system may be unable

to automatically start or achieve and maintain the required vacuum,

provided the following conditions exist:

a. All dampers communicating between the auxiliary building and the

enclosure building are closed.

b. The access door between the auxiliary building and the enclosure

building is closed, except when the access opening is being used for

entry and exit.

c. The SGT system is blocked from automatic initiation.

d. SGT system is available for manual initiation or the actions for

Limiting Condition for Operation 3.6.4.3 in the Technical

Specifications for GGNS are complied with.

The non-recently irradiated fuel is spent fuel that has decayed at

least 12 days after the reactor was shut down for refueling.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated.

The equipment affected by the proposed change is not considered

an initiator to any previously analyzed accident, therefore,

inoperability of the equipment does not increase the probability of

any previously evaluated accident.

As described in Updated Final Safety Analysis Report [for GGNS,]

Chapter 15, the accidents postulated to occur during core

alterations in addition to fuel handling accidents are [the

following]: inadvertent criticality due to a control rod removal

error or continuous control rod withdrawal error during refueling

and the inadvertent loading of a fuel assembly in an improper

location. These events are not postulated to result in fuel cladding

integrity damage. The only accident postulated to occur during core

alterations that results in a significant radioactive release is the

fuel handling accident. The proposed requirements in conjunction

with existing administrative controls on light loads, bounds the

conditions of the current design basis fuel handling accident

analysis which concludes that the radiological consequences are

within the acceptance criteria of NUREG 0800, Section 15.7.4 and

General Design Criteria [GDC] 19 [of Appendix A to 10 CFR Part 50].

Therefore, the proposed changes do not significantly increase

consequences of any previously evaluated accident.

Based on the above, the proposed changes do not significantly

increase the probability or consequences of any accident previously

evaluated.

2. The proposed changes would not create the possibility of a

new or different kind of accident from any previous analyzed.

The leaktightness of the enclosure building does not affect the

function of any plant system other than the ability of the SGT

System to ensure the secondary containment is at the specified

pressure. The proposed change in [the] normal SGT System

alignment[,] by defeating the automatic start feature of the SGT

System and the inability to ensure secondary containment is at the

specified pressure[,] does not affect the operation of any [other]

plant system or component. The SGT System is not relied upon to

provide normal or accident cooling to plant systems or components.

The function of the enclosure building and the SGT System is only to

mitigate the release of radioactivity to the environment in the

event of an accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes continue to ensure that the radiological

consequences are at or below the current GGNS licensing limit.

Safety margins and analytical conservatisms have been evaluated and

are well understood. Substantial margins are retained to ensure that

the analysis adequately bounds all postulated event scenarios. The

current margin of safety is retained.

Specifically, the margin of safety for the fuel handling

accident is the difference between the 10CFR100 [dose consequence

guidelines of 300 rem thyroid and 25 rem whole- body] and the

licensing limit defined by NUREG-0800, Section 15.7.4. With respect

to the control room personnel doses, the margin of safety is the

difference between the 10CFR100 [guidelines] and the licensing limit

defined by 10CFR50 [10 CFR Part 50], Appendix A, Criterion 19 (GDC

19). The proposed applicability continues to ensure that the whole-

body and thyroid doses at the exclusion area and low population zone

boundaries[,] as well as control room doses[,] are at or below the

corresponding licensing limit. The margin of safety is unchanged;

therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

In excess to the margin of safety supplied by the licensing

limits of NUREG-0800 and GDC 19, the proposed change incorporates an

additional layer of conservative requirements. The proposed change

leaves in effect a redefined secondary containment boundary which

will provide a low leakage boundary (consisting of the primary

containment and the auxiliary building) by automatically isolating

in the event of the design basis fuel handling accident and requires

that the SGT System be available for manual initiation when desired.

These requirements will ensure that doses will be even lower than

those calculated.

Therefore, the proposed changes do not result in a significant

reduction in a margin of safety.

Based on the above evaluation, operation in accordance with the

proposed amendment involves no significant hazards considerations.

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: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

Date of amendment request: June 17, 1996

Description of amendment request: The proposed amendments would

revise Technical Specification Section 5.3.1 to allow use of fuel

assemblies containing fuel rods clad with ZIRLOTM.

Basis for proposed no significant hazards consideration

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

[[Page 37300]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The methodologies used in the accident analyses remain

unchanged. The proposed change does not change or alter the design

assumptions for the systems or components used to mitigate the

consequences of an accident. Use of ZIRLOTM fuel cladding does

not adversely affect fuel performance or impact nuclear design

methodology. Therefore, accident analysis results are not

significantly impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC-approved methodologies. Other than the changes

to the fuel assemblies cladding, there are no physical changes to

the plant associated with this Technical Specification change. A

safety analysis will continue to be performed for each specific

reload cycle to demonstrate compliance with all fuel safety design

bases.

The 10 CFR 50.46 criteria are applied to the ZIRLOTM clad

fuel rods. The use of these fuel assemblies will not result in a

change to the reload design and safety analysis limits. Since the

original design criteria are met, the ZIRLOTM clad fuel rods

will not be an initiator for any new accident. The clad material is

similar in chemical composition and has similar physical and

mechanical properties as Zircaloy-4. Thus, the cladding integrity is

maintained and the structural integrity of the fuel assembly is not

affected. ZIRLOTM cladding improves corrosion performance and

dimensional stability. Since the dose predictions in the safety

analyses are not sensitive to the fuel rod cladding material used,

the radiological consequences of accidents previously evaluated in

the safety analysis remain valid.

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

different kind of accident from any accident previously evaluated

because:

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

accident previously evaluated is not created since the fuel

assemblies containing ZIRLOTM clad fuel rods will satisfy the

same design bases as that currently used for Zircaloy-4 clad fuel

assemblies. All design and performance criteria will continue to be

met and no new single failure mechanisms have been defined. In

addition, the use of ZIRLOTM fuel assemblies does not involve

any alterations to plant equipment or procedures which would

introduce any new or unique operational mode or accident precursor.

Therefore, the possibility for a new or different kind of accident

from any accident previously evaluated is not created.

The proposed change does not involve a significant reduction in

a margin of safety because:

The margin of safety is not significantly reduced since the

ZIRLOTM clad fuel assemblies will not change the reload design

and safety analysis limits. Their use will take into consideration

the normal core operating conditions allowed for in the Technical

Specifications. Each specific cycle's reload core will continue to

be specifically evaluated using NRC approved reload design methods

and approved fuel rod design models. This will include consideration

of the core physics analysis peaking factor and core average linear

heat rate effects.

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: BurkeCounty Public Library,

412 Fourth Street, Waynesboro, Georgia 30830

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Herbert N. Berkow

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

Date of amendment request: June 17, 1996

Description of amendment request: The proposed amendments would

clarify the requirement of Technical Specification Surveillance

Requirement 4.8.1.1.2.j(2) that requires a pressure test of those

portions of the diesel fuel-oil system that are designed to Section

III, Subsection ND of the American Society of Mechanical Engineers

(ASME) Code. The system pressure test would be performed at a pressure

of 110% of the design pressure, at least once per 10 years and only on

those sections of piping that are isolable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed clarification of T/S [Technical Specification]

4.8.1.1.2.j(2) does not involve a significant hazards consideration

because operation of [the Vogtle Electric Generating Plant] with

this change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The configuration

of the diesel fuel-oil system as currently installed and operated is

such that a pressure test of 110% of design pressure would be

impractical to perform. The system contains tanks designed for

atmospheric pressure and isolation of them and their vent lines from

the specified pressure test is not practical. The ASME Code, Section

XI, provides alternate test methods to use when storage tanks are

involved in a system pressure test. By clarifying this T/S

requirement, the requirements set forth in ASME Section XI can be

utilized as guidance for testing requirements to ensure the

integrity of the diesel fuel-oil system to perform its intended

safety function.

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

from any accident previously evaluated. There are no design changes

being made that would create a new type of accident or malfunction

and the method and manner of plant operation remain unchanged. Using

ASME Section XI as guidance for pressure testing the isolable

sections of piping provides assurance that the fuel oil supply

system will perform its intended function.

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

are no changes being made to the safety limits or safety system

settings that would adversely impact plant safety. Utilizing ASME

Section XI as guidance for determining those sections of piping that

should be pressure-tested and atmospheric-tested will ensure proper

operation of the diesel generator fuel oil supply system.

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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: April 10, 1996

Description of amendment request: The proposed changes bring the

surveillance requirements to conformance with Amendment No. 196 issued

September 19, 1995. Additionally, this request changes frequency

notation for a group of surveillance requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

[[Page 37301]]

consideration (SHC), which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence of the consequences of an accident

previously evaluated.

The proposed amendment extends the interval between successive

refueling interval surveillances to once every 24 months for those

surveillances evaluated herein, and to make administrative changes

serving to conform the Technical Specifications to Amendment No.

196. Except for the administrative changes, the proposed

surveillance interval changes do not involve any change to the

actual surveillance requirements, nor does it involve any

change to the limits and restrictions on plant operations. The

reliability of systems and components relied upon to prevent or

mitigate the consequences of accidents previously evaluated is not

degraded by the proposed change to the surveillance interval.

Assurance of system and equipment availability is maintained. This

change does not involve any change to system or equipment

configuration. Therefore, this change does not increase the

probability of occurrence or the consequences of an accident

previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment extends the interval between successive

refueling interval surveillances to once every 24 months for those

surveillances evaluated herein, and to make administrative changes

serving to conform the Technical Specifications to Amendment No.

196. Except for the administrative changes the proposed surveillance

interval changes do not involve any change to the limits and

restrictions in plant operation. This change does not involve any

change to system or equipment configuration. Therefore, this change

is unrelated to the possibility of creating a new or different kind

of accident from any previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment extends the interval between successive

refueling interval surveillances to once every 24 months for the

surveillances evaluated herein, and to make administrative changes

serving to conform the Technical Specifications to Amendment No.

196. Except for the administrative changes the proposed surveillance

interval changes do not involve any change to the actual

surveillance requirements, nor does it involve any change to the

limits and restrictions on plant operation. The reliability of

systems and components is not degraded by the proposed change to the

surveillance interval. Assurance of system and equipment

availability is maintained. Therefore, it is concluded that

operation of the facility in accordance with the proposed amendment

does not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

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

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

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: John F. Stolz

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,

Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: May 2, 1996

Description of amendment request: The proposed change would remove

Technical Specification Figure 5.1, which is used in maintaining

Keff values, and substitute in its place a defined requirement for

maximum K-infinity for any fuel placed in the Millstone Unit 1 spent

fuel pool.

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:

Pursuant to 10CFR50.92, NNECO [Northeast Nuclear Energy Company]

has reviewed the proposed change and concludes that the change does

not involve a significant hazards consideration (SHC) since the

proposed change satisfies the criteria in 10CFR50.92(c). That is,

the proposed change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

There are no spent fuel pool accident conditions discussed in

Chapter 15 of the FSAR [Final Safety Analysis Report]. FSAR section

15.8 discusses a fuel handling accident which drops a fuel assembly

into the core during refueling. Changing the maximum allowed fuel

reactivity or allowing gaps in the Boraflex

panels will have no effect on the probability or consequences of

a fuel assembly drop onto the core.

Therefore, based on the above, the proposed change to the

Technical Specifications does not involve a significant increase in

the probability or consequences of any previously analyzed accident.

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

from any accident previously evaluated.

The reduction in the allowable fuel reactivity in the SFP [spent

fuel pool] is conservative and does not create the possibility of a

new or different type of accident. Allowing boraflex gaps does not

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

any accident previously evaluated.

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

The margin to safety, for this proposed technical specification

change, is to maintain the SFP Keff to be less than or equal to

0.90. As described in the HOLTEC analysis, gaps in the Boraflex of

up to 5 inches can exist in every boraflex panel of every rack with

Boraflex in the SFP, with Keff still less than 0.90. This is

true even if all of the gaps are uniformly lined up at the same

elevation. These calculations conservatively assumed 4% Boraflex

width shrinkage as well as the axial Boraflex gaps. Older fuel

designs were also considered to ensure that they had not become

limiting with the reduced allowable K-infinity limit of 1.24. With

no boraflex gaps, the maximum Keff is less than .844. With 5

inch Boraflex gaps in every panel at the same elevation, the maximum

Keff is 0.896, which is less than 0.90. NNECO has implemented a

1 year decay time requirement to minimize gamma irradiation damage

to the Boraflex, and will continue to measure via ``blackness

testing'' the actual gap size to ensure the margin of safety in

maintained.

Therefore, this change has no impact on the margin to 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 Project Director: Phillip F. McKee

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

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

Date of amendment request: March 29, 1996

Description of amendment request: The proposed amendment would add

limits associated with Departure from Nucleate Boiling (DNB) to the

Indian Point 3 (IP3) Technical Specifications.

Basis for proposed no significant hazards consideration

determination:

[[Page 37302]]

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:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no significant hazards based on the

following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response:

The proposed amendment makes no changes to the way in which the

plant is operated and has no effect on accident initiators

associated with analyzed transients. The probability of previously

analyzed accidents is not increased. The proposed amendment

clarifies the relationship between measurable parameters (RCS

[reactor coolant system] temperature, pressure, and flow rate) and

the resulting heat transfer regime in the reactor core, as

characterized by the Departure from Nucleate Boiling (DNB) ratio.

This clarification ensures that safety analysis initial conditions

regarding heat transfer remain valid, so that the consequences of

previously analyzed accidents are not increased. The changes ensure

that RCS pressure, temperature, and flow are within analytical

bounds. This ensures that the plant is operated in a manner that

will not increase the probabilities of previously analyzed accidents

nor the consequences of previously analyzed accidents.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

The proposed amendment does not involve any modifications to

plant systems, structures, or components. The proposed change

clarifies existing limits on RCS parameters and makes no changes to

plant setpoints or operating limits. The amendment does not involve

any physical mechanism which could contribute to a new or different

kind of accident. The changes ensure that RCS pressure, temperature,

and flow are within analytical bounds. This ensures that the plant

is operated in a manner that will not create the possibility of a

new [or] different kind of accident from any previously evaluated.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed amendment clarifies existing limits on the

measurable parameters (RCS temperature, pressure, and flow rate) so

that the resulting DNB value is consistent with initial condition

assumptions used in existing safety analyses. Maintaining these

limits during normal plant operation ensures that the existing

margins of safety remain valid. The proposed amendment does not

involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Jocelyn A. Mitchell, Acting

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: June 18, 1996

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 5.2.2, ``Design Pressure and

Temperature,'' by adding design parameters for Main Steam Line Break

(MSLB). The MSLB analysis results in a higher containment air

temperature than the current value in TS 5.2.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The accidents considered for this change are the Loss of Coolant

Accident (LOCA) and the Main Steam Line Break (MSLB). The proposed

change ensures the design limiting containment pressure and

temperature data specified in the TS is consistent with the [Updated

Final Safety Analysis Report] UFSAR. Since no physical changes to

the containment are being made there will be no change in the

probability of either accident occurring.

Detailed structural analysis presented in Supplement 1 of

Licensee Event Report (LER) 272/95-016 shows that the Design Basis

LOCA combination of pressure and temperature result in more severe

loading for the containment concrete structure and, therefore,

bounds the temperature and pressure scenario associated with a MSLB

accident. The pressure retaining capability of the liner is governed

by the loads generated in the MSLB. Since containment leakage is

maintained within the limits assumed in the Accident Analysis for

either scenario there is no change in the consequences of either

accident.

Therefore the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The changes proposed affect the post-accident condition of the

containment, and have no impact on the pre-accident condition. Since

there is no physical change proposed the containment and all systems

in the containment will continue to perform as designed. With no

physical changes being proposed and no change to the pre-accident

condition of the containment it can be concluded that there will be

no change in the probability of a new or different accident being

created.

Therefore the proposed change does 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.

Although calculations indicate that some yielding of the liner

plate could occur during a MSLB, loading is transferred to the

containment concrete structure and leakage from the containment is

maintained within the limits assumed in the Accident Analysis. Since

containment leakage is maintained within the limits assumed in the

Accident Analysis the proposed change does not involve a significant

change the margin of safety provided by the containment for the

MSLB.

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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington, DC 20005-3502

NRC Project Director: John F. Stolz

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

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

Date of amendment request: June 7, 1996 (TSC 95-19)

Description of amendment request: The proposed change would revise

Section 6 of the plant Technical Specifications to be more closely

aligned with the Revised Standard Technical Specifications for

Westinghouse-designed nuclear plants (NUREG-1431). Additionally, the

proposed changes would be consistent with the guidance provided in

Administrative Letter 95-06, ``Relocation of Technical Specification

Administrative Controls Related to Quality Assurance.''

Basis for proposed no significant hazards consideration

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

[[Page 37303]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

TVA [Tennessee Valley Authority] has concluded that operation of

SQN [Sequoyah Nuclear Plant] Units 1 and 2 in accordance with the

proposed changes to the TS [Technical Specification] does not

involve a significant hazards consideration. TVA's conclusion is

based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of

the three standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed TS change is administrative. TVA has evaluated the

proposed TS changes and has determined that the proposed changes are

administrative in nature. Certain sections are being relocated into

other licensee documents for which those provisions are adequately

controlled by regulatory requirements. These changes do not affect

any of the design basis accidents. They do not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any previously evaluated.

The proposed TS change is administrative. TVA has evaluated the

proposed TS changes and has determined that the proposed changes are

administrative in nature. Certain sections are being relocated into

other licensee documents for which those provisions are adequately

controlled by regulatory requirements. These changes do not affect

any of the design-basis accidents. No modifications to any plant

equipment are involved. There are no effects on system interactions

made by these changes. They do not create the possibility of a new

or different kind of accident from an accident previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed TS change is administrative. TVA has evaluated the

proposed TS changes and has determined that the proposed changes are

administrative in nature. Certain sections are being relocated into

other licensee documents for which those provisions are adequately

controlled by regulatory requirements. The margin of safety as

reported in the basis for the TSs is not reduced. The proposed

change is administrative and does not impact any technical

information contained in the bases of the TS.

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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

North Atlantic Energy Service Company, Docket No. 50-443, Seabrook

Plant Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: June 20, 1996

Description of amendment request: The proposed amendment would

increase the allowed time for an inoperable service water cooling tower

loop electrical supply to be the same as the allowed outage time for an

operable service water cooling tower loop.

Date of publication of individual notice in Federal Register: June

26, 1996 (61 FR 33142)

Expiration date of individual notice: July 26, 1996

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, New Hampshire

Northeast Utilities Service Company, Docket No. 50-336, Millstone

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

Date of amendment request: June 3, 1996

Description of amendment request: The proposed amendments would

provide a one-time change to Technical Specification 3.9.1, ``Refueling

Operations, Boron Concentration.'' This change would remove the

requirement that the boron concentration in all filled portions of the

Reactor Coolant System be ``uniform'' and would only be applicable

during Millstone 2 Cycle 13 mid-cycle core offload.

Date of publication of individual notice in Federal Register: June

12, 1996 (61 FR 29771)

Expiration date of individual notice: July 12, 1996

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut

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

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

Connecticut

Date of amendment request: May 23, 1996

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) for the Overtemperature delta

T time constants in TS Table 2.2-1 and the Steam Line Pressure Negative

Rate High Steam Line Isolation time constant in TS Table 3.3-4. Date of

publication of individual notice in Federal Register: June 17, 1996 (61

FR 30639)

Expiration date of individual notice: July 17, 1997

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut

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

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: June 10, 1996

Brief description of amendment request: The amendment proposes

changes to Technical Specification 3/4.7.6, ``Control Room Emergency

Air Conditioning System,'' to reflect a control room design in which

the common Salem Unit 1 and 2 control room envelope is supplied by 2

one hundred percent capable Control Room Emergency Air Conditioning

System trains. Date of publication of individual notice in Federal

Register: June 24, 1996 (61 FR 32468)

Expiration date of individual notice: July 24, 1996

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

West Broadway, Salem, NJ 08079

[[Page 37304]]

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendment request: June 24, 1996

Description of amendment request: The proposed amendments would

revise Technical Specification Table 4.3.1 to delete the requirement

for surveillance of the manual safety injection to the reactor trip

circuitry until the next unit shutdown, following which, this testing

will be performed prior to Mode 2 entry. This change is applicable only

to Unit 1, Cycle 14 and Unit 2, Cycle 11. Date of publication of

individual notice in Federal Register: July 3, 1996 (61 FR 34880)

Expiration date of individual notice: August 2, 1996

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, P. O. Box 1369, Dothan, Alabama

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of application for amendments: September 16, 1994, as

supplemented January 31, 1996.

Brief description of amendments: The amendments revise the

technical specifications to eliminate periodic response time testing

requirements for selected pressure and differential pressure sensors in

the reactor trip system and engineered safety features actuation

instrumentation channels.

Date of issuance: June 26, 1996

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

Amendment Nos.: 84, 84, 76 and 76

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 13, 1996 (61 FR

10393). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 26, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

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

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: November 14, 1995, as

supplemented by letters dated February 23, March 1, March 13, March 25,

March 26, May 10, June 10, June 14, two letters dated June 25 and a

letter dated June 26, 1996.

Brief description of amendments: The proposed amendments closed out

additional open items identified in the NRC staff's review of the

upgrade of the Dresden and Quad Cities Technical Specifications (TS) to

the Standard Technical Specifications (STS) contained in NUREG-0123.

The Technical Specification Upgrade Program (TSUP) is not a complete

adaptation of the STS. The TS upgrade focuses on (1) integrating

additional information such as equipment operability requirements

during shutdown conditions, (2) clarifying requirements such as

limiting conditions for operation and action statements utilizing STS

terminology, (3) deleting superseded requirements and modifications to

the TS based on the licensee's responses to Generic Letter (GL), and

(4) relocating specific items to more appropriate TS locations.

Date of issuance: June 28, 1996

Effective date: June 28, 1996

Amendment Nos.: 150, 145, 171, and 167

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications and operating

licenses.

Date of initial notice in Federal Register: November 29, 1995 (60

FR 61272) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 28, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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

Michigan

Date of application for amendment: November 22, 1995 (NRC-95-0124)

Brief description of amendment: The amendment revises the Technical

Specifications to remove accelerated testing frequencies and special

reporting requirements for Fermi 2 emergency diesel generators (EDGs)

in accordance with guidance contained in Generic Letter 94-01, dated

May 31, 1994. NRC will issue a separate safety evaluation on extending

the allowed outage time for the EDGs at a later date.

Date of issuance: June 20, 1996

Effective date: June 20, 1996, with full implementation within 60

days

Amendment No.: 107

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

Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7550) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated

[[Page 37305]]

June 20, 1996. No significant hazards consideration comments received:

No

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

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

Date of application for amendments: December 12, 1995, as

supplemented by letter dated June 10, 1996

Description of amendment request: The amendments revise the

absolute values in the Axial Flux Difference (AFD) Equations to reflect

the proper AFD limit reduction in the current Technical Specifications.

Date of issuance: July 2, 1996

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

30 days.

Amendment Nos.: 167 and 149

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 24, 1996 (61 FR

18166) The June 10, 1996, letter provided clarifying information that

did not change the scope of the December 12, 1995, application and the

initial proposed no significant hazards consideration determination.

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

Safety Evaluation dated July 2, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 11, 1995, as supplemented by

letter dated February 12, 1996

Brief description of amendment: The amendment reduced the minimum

reactor coolant cold leg temperature to 541 deg.F from 544 deg.F in

Technical Specification Section 3.2.6, ``Reactor Coolant Cold Leg

Temperature.''

Date of issuance: June 24, 1996

Effective date: June 24, 1996

Amendment No.: 120

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

Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

25706) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 24, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

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

Date of application for amendments: March 20, 1996, as supplemented

by letter date April 23, 1996.

Brief description of amendments: These amendments relocate the

requirements for surveillance testing of the water level and pressure

channel instrumentation for the reactor coolant system accumulators.

These amendments also modify the existing action statements of TS 3.5.1

for accumulators to reflect the requirements of NUREG-1431 by requiring

a 72- hour period to restore boron concentration if it is not within

the limits, and a 1-hour period to restore any other condition

rendering the accumulators inoperable.

Date of issuance: June 24, 1996

Effective date: June 24, 1996

Amendment Nos. 185 and 179Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: May 22, 1996 (61 FR

25707) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 24, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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: February 21, 1996, as

supplemented by letters dated May 1 and June 4, 1996.

Brief description of amendments: The amendments revise the

Technical Specifications to change the Drywell Air Temperature Limiting

Condition for Operation (LCO) from less than or equal to 135 deg.F to

less than or equal to 150 deg.F. The proposed change would provide a

margin for the primary containment Drywell Air Temperature LCO when

prolonged summer and high river temperatures are experienced. Also, a

strictly editorial correction to a Final Safety Analysis Report

reference would be made.

Date of issuance: 201 and 142

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

30 days.

Amendment Nos.: 201 and 142

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 24, 1996 (61 FR

18167) The May 1 and June 4, 1996, letters provided clarifying

information that did not change the scope of the February 21, 1996,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 27, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

Michigan

Date of application for amendments: May 19, 1995, and supplemented

October 20, 1995, and April 8, 1996 (AEP:NRC:1213A)

Brief description of amendments: The amendments modify the neutron

flux high setpoints for one or more main steam safety valves inoperable

in response to Westinghouse Nuclear Safety Advisory Letter 94-001. The

associated action statements are also revised and an exemption to TS

4.0.4 is added to support the operability surveillance.

Date of issuance: June 28, 1996

Effective date: June 28, 1996, with full implementation within 45

days.

Amendment Nos.: Unit 1 - 210, Unit 2 - 195

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

revised the Technical Specifications.

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

FR 65681) The April 8, 1996, submittal provided information clarifying

the location of the TS 4.0.4 exemption statement. This information was

within the scope of the original application and did not alter the

staff's no significant hazards considerations determination. Therefore

renoticing was not warranted. The Commission's related evaluation of

the amendments is contained in a Safety Evaluation dated June 28, 1996.

No significant hazards consideration comments received: No.

[[Page 37306]]

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile

Point Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: January 17, 1996

Brief description of amendment: The amendment revises the Technical

Specifications (TSs) and associated Bases by relocating certain

response time limit tables from the TSs to the Updated Safety Analysis

Report in accordance with the guidance of NRC Generic Letter 93-08. The

relocated tables are for instrumentation for the Reactor Protection

System, Isolation Actuation System, Emergency Core Cooling System, and

the Recirculation Pump Trip System.

Date of issuance: June 25, 1996

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

30 days.

Amendment No.: 73

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20850) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 25, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

Connecticut

Date of application for amendment: December 18, 1995

Brief description of amendment: The amendment changes the Reactor

Coolant Flow - Low Flow in Technical Specification Table 2.2-1,

``Reactor Instrumentation Protective Trip Setpoint Limits.'' The

proposed change increases the allowable value from greater than or

equal to 90.1% to greater than or equal to 90.9% of the reactor coolant

flow with four pumps operating. As an editorial change for

clarification, the word ``flow'' is added after ``reactor coolant'' in

the above sentence.

Date of issuance: July 2, 1996

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

within 60 days.

Amendment No.: 199

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

Technical Specifications.

Date of initial notice in Federal Register: February 14, 1996 (61

FR 5815) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 2, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385

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

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

Connecticut

Date of application for amendment: June 27, 1995, as supplemented

July 21, 1995

Brief description of amendment: The amendment revises the Technical

Specifications (TS) to relocate TS requirements for the containment

purge exhaust and supply valves, and to remove a duplicate testing

requirement for the safety injection input from engineered safety

features from the TS.

Date of issuance: June 27, 1996

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

within 60 days.

Amendment No.: 129

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

Technical Specifications.

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

62494) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 27, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385

Pennsylvania Power and Light Company, Docket No. 50-387,

Susquehanna Steam Electric Station, Unit 1, Luzerne County,

Pennsylvania

Date of application for amendment: January 26, 1996

Brief description of amendment: The amendment deletes three

residual heat removal (RHR) system relief valves from Technical

Specification (TS) Table 3.6.3-1, ``Primary Containment Isolation

Valves.'' These valves are no longer needed to support the steam

condensing mode of RHR and are being removed from the plant during the

Unit 1 ninth refueling outage.

Date of issuance: June 24, 1996

Effective date: As of date of issuance to be implemented within 60

days.

Amendment No.: 157

Facility Operating License No. NPF-14: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 27, 1996 (61 FR

13531) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 24, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

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: February 29, 1996

Brief description of amendments: These amendments relocate

Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam

Electric Station Technical Requirements Manual, a document which is

controlled under the requirements of 10 CFR 50.59.

Date of issuance: July 2, 1996

Effective date: July 2, 1996

Amendment Nos.: 158 and 129

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15992) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 2, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

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

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

Date of application for amendment: March 12, 1996

Brief description of amendment: The proposed changes would remove a

requirement to cross tie safety injection accumulators.

Date of issuance: July 3, 1996

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

30 days.

[[Page 37307]]

Amendment No.: 167

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

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20853) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 3, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: April 24, 1996

Brief description of amendment: The amendment proposes to relocate

Specification 3.11.B/4.11.B ``Crescent Area Ventilation'' and

associated Bases from the TS to an Authority controlled procedure.

Date of issuance: June 28, 1996

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

30 days.

Amendment No.: 231

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 22, 1996 (61 FR

25710) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 28, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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: February 6, 1996

Brief description of amendments: The amendments change the

Technical Specifications to lower the 125 Volt Battery Charger

surveillance amperage from at least 200 amps to at least 170 amps.

Date of issuance: June 27, 1996

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment Nos. 183 and 164

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

revised the Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7556) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 27, 1996. No significant

hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: June 26, 1995, as supplemented

by letter dated February 2, 1996.

Brief description of amendment: The amendment revised the allowed

outage time for component cooling water motor operated containment

isolation valves, moved the list of containment isolation valves from

the technical specifications to the final safety analysis report, and

allowed containment penetration check valves to be used as isolation

devices.

Date of issuance: June 28, 1996

Effective date: June 28, 1996, to be implemented within 30 days of

the date of issuance.

Amendment No.: 113

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45187) The February 2, 1996, supplemental letter provided additional

clarifying information and did not change the staff's original no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated June 28, 1996. No significant hazards consideration comments

received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Dated at Rockville, Maryland, this 10th day of July 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

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

Reactor Regulation.

[Doc. 96-18007 Filed 7-16-96; 8:45 am]

BILLING CODE 7590-O1-F