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 October 28, 1996, through November 7, 1996.

The last biweekly notice was published on November 6, 1996.

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 December 20, 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

[[Page 58901]]

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: 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.

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

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

Date of amendment request: October 2, 1996

Description of amendment request: The amendment would change

Figures 3.1.A-1, 3.1.A-2, and 3.1.A-3, Section 3.1.B and its Bases,

Figures 3.1.B-1 and 3.1.B-2, and the Bases of Section 4.3 and Figure

4.3-1 of the Technical Specifications by providing new pressure/

temperature limit curves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1)Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Response:

Neither the probability nor the consequences of an accident

previously analyzed is increased due to the proposed changes. The

adjusted reference temperature of the most limiting beltline

material was used to correct the pressure-temperature (P-T) curves

to account for irradiation effects. Thus, the operating limits are

adjusted to incorporate both the initial fracture toughness

conservatism present when the reactor vessel was new and the effect

of fluence. The adjusted reference temperature calculations were

performed utilizing the guidance contained in RG [Regulatory Guide]

1.99, Revision 2. Overpressure Protection System (OPS) curves and

tables were regenerated to be consistent with the new P-T curves.

The updated curves provide assurance that brittle fracture of the

reactor vessel is prevented.

2) Does the proposed license amendment create the possibility of

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

Response:

The updated P-T and OPS limits will not create the possibility

of a new or different kind of accident. The revised operating limits

merely update the existing limits by taking into account the effects

of radiation embrittlement, utilizing criteria defined in RG 1.99,

Revision 2. The updated curves are conservatively adjusted to

account for the effect of irradiation on the limiting reactor vessel

material.

No change is being made to the way the pressure-temperature

limits provide plant protection. No new modes of operation are

involved. Incorporating this amendment does not necessitate physical

alteration of the plant.

3) Does the proposed amendment involve a significant reduction

in the margin of safety?

Response:

The proposed amendment does not involve a significant reduction

in the margin of safety. The pressure-temperature operating limits

and OPS setpoints are designed to maintain an appropriate margin of

safety. The required margin is specified in ASME [American Society

of Mechanical Engineers] Boiler and Pressure Vessel Code, Section

III, Appendix G and 10 CFR [Part] 50 Appendix G. The revised curves

are based on the latest NRC guidelines along with actual neutron

fluence data for the reactor vessel. The new limits retain a margin

of safety equivalent to the original margin when the vessel was new

and the fracture toughness was slightly greater. The new operating

limits account for irradiation embrittlement effects, thereby

maintaining a conservative margin of safety.

The removal of the pressure-temperature limits for criticality

does not reduce the plant safety margin because these limits are

conservatively encompassed and bounded by the requirements of the

proposed Technical Specification 3.1.C.2.

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

this

[[Page 58902]]

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 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Acting

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: September 6, 1996

Description of amendment request: The proposed amendments would

revise Item 7.c of BVPS-1 Technical Specifications (TSs) Table 3.3-3

and Item 7.d of BVPS-2 TS Table 3.3-3 to reflect that a safety

injection (SI) signal starts all auxiliary feedwater (AFW) pumps. The

notation on BVPS-1 TS Table 3.3-5 would be revised to state that the

response time is for all AFW pumps on all SI signal starts. Items 7.d

of BVPS-2 TS Tables 3.3-4 and 4.3-2 would be revised to reflect that an

SI signal starts all AFW pumps.

The proposed amendments would also revise and reformat TSs 3/

4.7.1.2 to more closely resemble the wording contained in the NRC's

``Standard Technical Specifications Westinghouse Plants,'' (NUREG-1431,

Revision 1). These changes would require three AFW trains to be

operable and would provide what constitutes an operable train. The mode

applicability for these TSs would expand to include Mode 4 when the

steam generator(s) is relied upon for heat removal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed revisions to reflect that a Safety Injection (SI)

signal starts the turbine driven Auxiliary Feedwater (AFW) pump, in

addition to both motor driven AFW pumps, will ensure that plant

operability requirements for the AFW system actuation signals are

maintained at a level consistent with current safety analyses. The

proposed revisions to Limiting Condition for Operation (LCO) 3.7.1.2

will require that the AFW pumps and associated flow paths are

maintained operable to ensure that the AFW system can mitigate the

consequences of a Design Basis Accident (DBA) with a loss of normal

feedwater. The addition of the Mode 4 applicability will ensure that

a safety related source of cooling water is available to remove

decay heat.

The proposed change will ensure that the plant is placed in Mode

4 when the number of operable feedwater injection headers is

insufficient to ensure that at least two steam generators are

supplied during a feedline break accident.

The proposed addition of footnote (2) to action statement ``c''

will limit plant thermal cycles following a refueling outage due to

turbine driven AFW pump inoperability. During the additional time

period provided by footnote (2) to reach Hot Shutdown, the two

remaining motor driven AFW pumps will provide sufficient flow to the

steam generators to mitigate the consequences of a DBA assuming no

single failures during this time period. Since there is negligible

decay heat following a refueling outage prior to entry into Mode 2,

the performance capabilities of the two remaining motor driven AFW

pumps to remove decay heat will not be challenged.

Changing the AFW pump surveillance test frequencies for Beaver

Valley Power Station (BVPS) Unit No. 2 to quarterly, as specified in

the Inservice Testing (IST) Program, will continue to assure that

the AFW system will be capable of performing its intended functions.

The proposed change to the current Surveillance Requirement

4.7.1.2, for BVPS Unit No. 2 only, will not lower the pump

performance operability criteria for the AFW pumps. The required

values for developed pump head and flow will continue to satisfy

accident mitigation requirements and will be maintained and

controlled in the BVPS Unit No. 2 IST Program. Future changes to the

AFW pump head and flow requirements will be made under the 10 CFR

50.59 process to ensure that the AFW design requirement to remove

sufficient decay heat continues to be met.

Based on the above factors, the probability of an accident

previously evaluated is not significantly increased.

The proposed changes do not affect the ability of the AFW system

to perform as assumed in the safety analyses. The proposed changes

will not result in any additional challenges to plant equipment.

Because the plant design limits will continue to be met, the fuel

and reactor coolant system pressure boundary integrity is not

challenged for the assumptions employed in the calculation of the

offsite radiological doses. The additional time to reach Mode 4 from

Mode 3 provided by footnote (2) does not result in increased

radiological consequences. The potential for a radioactivity release

due to the uncontrolled heatup of [the] reactor coolant system[s]

are enveloped by the releases postulated in the DBA Loss of Coolant

Accident (LOCA) analysis in the Updated Final Safety Analysis

Report. The DBA LOCA analysis assumes 102% power operation prior to

the event and assumes that core melt occurs. Therefore, there is no

increase in the radiological consequences as a result of allowing

additional time to repair/test the turbine driven AFW pump. Hence,

the consequences of a DBA previously evaluated is not significantly

increased.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated.

The proposed change does not alter the method of operating the

plant. The AFW system is an accident mitigation system and is

normally in standby. System operation is initiated in response to a

DBA. The AFW pumps will continue to provide sufficient flow to

mitigate the consequences of a DBA. AFW operation continues to

fulfill the safety function for which it was designed and no changes

to plant equipment will occur. As a result, an accident which is new

or different than any already evaluated in the Updated Final Safety

Analysis Report will not be created due to this change.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

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

of safety?

The proposed changes will not affect the heat removal capability

of the AFW system to a value less than assumed in the safety

analysis. The proposed changes will not result in any additional

challenges to the plant equipment including the fuel and reactor

coolant system pressure boundary. The additional time period to

reach Hot Shutdown provided by footnote (2) will not significantly

reduce the decay heat removal capability provided by the AFW system.

The two remaining motor driven AFW pumps will continue to provide

sufficient flow to the steam generators as assumed in the safety

analysis to mitigate the consequences of a DBA assuming no single

failure during this time period. The plant will continue to operate

within the bounds of the safety analysis.

The AFW system will continue to be tested in a manner and at a

frequency which will ensure acceptable system performance should it

be relied upon to remove decay heat following a DBA.

The AFW pumps' performance requirements will continue to be

controlled in a manner to ensure safety analysis assumptions are

met.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001

[[Page 58903]]

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

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

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

Date of amendment request: July 25, 1996

Description of amendment request: The proposed change modifies

Technical Specification (TS) 3/4.7.4 Ultimate Heat Sink (UHS) by

incorporating more restrictive fan operability requirements and lower

basin temperature. Several other administrative changes are

incorporated to improve the humanfactors associated with this TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed change modifies the UHS TS by revising [Wet Cooling

Tower] WCT basin water temperature from less than or equal to 95

Degrees Fahrenheit to less than or equal to 89 Degrees Fahrenheit

and incorporating more restrictive cooling tower fan operability

requirements. These changes are necessary to adequately preserve the

assumptions and limits of the revised UHS design basis calculations.

These calculations conclude that the UHS is capable of dissipating

the maximum peak heat load resulting from the limiting design bases

accident (i.e., large break LOCA) and the most severe natural

phenomena (i.e., tornado event). Other changes are purely

administrative in nature. The proposed change does not directly

affect any material condition of the plant that could directly

contribute to causing an accident. The proposed change ensures that

the mitigating effects of the UHS will be consistent with the design

basis analysis. Therefore, the proposed change will not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

accident from any accident previously evaluated?

Response: No

The proposed change modifies the UHS TS to be consistent with

revised design basis calculations. These new calculations adjust

margin to incorporate an additional allowance for fouling in the

[Component Cooling Water] CCW heat exchangers and more restrictive

UHS minimum fan requirements that were not adequately addressed in

the initial design basis. This change also incorporates

administrative changes that are intended to improve the application

and use of this specification. The proposed change will not alter

the operation of the plant or the manner in which the plant is

operated. Therefore, the proposed change will not create the

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

previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No

The proposed change modifies the UHS TS by revising WCT basin

water temperature from less than or equal to 95 Degrees Fahrenheit

to less than or equal to 89 Degrees Fahrenheit and incorporating

more restrictive cooling tower fan operability requirements.

Modifying the UHS meteorological design bases reduced WCT basin

temperature requirement for operability, thus, providing an

allowance for fouling in the CCW heat exchangers. The proposed

change better preserves the margin of safety by ensuring that the

UHS will maintain the CCW accident analysis temperature limit of 115

Degrees Fahrenheit. Increased cooling tower fan operability

requirements will ensure that the expected cooling efficiency is

actually available and not unknowingly degraded due to fouling.

Other changes requested herein are purely administrative in nature,

do no affect safety margins and intended to improve the use and

application of this specification. Therefore, the proposed change

will 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: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 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: October 4, 1996

Description of amendment request: The proposed amendments would

incorporate the requirements necessary to change the basis for

prevention of criticality in the fuel storage pool. This change would

eliminate credit for Boraflex as a neutron absorbing material in the

fuel storage pool criticality analysis and would support the storage of

fuel with enrichments up to and including 5.0 weight percent U-235

rather than the current value of 4.5 weight percent U-235.

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.

There is no increase in the radiological consequences of

accidents previously evaluated in the Vogtle FSAR [Final Safety

Analysis Report] with the use of 5.0 weight percent U-235 fuel.

Increasing the enrichment up to and including 5.0 weight percent U-

235 affects the radiological source terms and subsequently the

potential releases both normal and accidental. Evaluations performed

(WCAP-12610-P-A, Reference 6) considered the source term, gap

fraction, normal operating plant releases and the accident doses for

a maximum fuel enrichment of 5.0 weight percent U-235. It was

concluded that operating with and storing fuel with 5.0 weight

percent U-235 enrichment may result in minor increases in the normal

annual releases of long half-life fission products that are not

significant. Also, the radiological consequences of accidents are

minimally affected due to the very small changes in the core

inventory and the fact that the currently assumed gap fractions

remain bounding.

The use of the slightly higher enrichment for VEGP [Vogtle

Electric Generating Plant] fuel will not result in burnups in excess

of those currently allowed for VEGP. The cycle design methods and

limits will remain the same as are currently licensed. Therefore the

use of fuel with the higher enrichment is not expected to result in

operating conditions outside those currently allowed for VEGP.

There is no increase in the probability of a fuel assembly drop

accident in the fuel storage pool when considering the presence of

soluble boron in the pool water for criticality control. The

handling of the fuel assemblies in the fuel storage pool has always

been performed in borated water.

Fuel assembly placement will be controlled pursuant to approved

fuel handling procedures and will be in accordance with the spent

fuel rack storage configuration limitations in the COLR [Core

Operating Limit Report]. The consequences of a misplaced assembly

have been included in the analysis supporting this revision to the

Technical Specifications.

There is no increase in the consequences of the accidental

misloading of a spent fuel assembly into the fuel storage pool racks

because criticality analyses demonstrate that

[[Page 58904]]

the pool will remain subcritical following an accidental misloading

of an assembly even considering a dilution event. The proposed

Technical Specifications and COLR limitations will ensure that an

adequate fuel storage pool boron concentration will be maintained.

There is no increase in the probability of the loss of normal

cooling to the fuel storage pool water due to the presence of

soluble boron in the pool water for subcriticality control, because

a high concentration of soluble boron has been maintained in the

fuel storage pool water.

The loss of normal cooling to the fuel storage pool will cause

an increase in the temperature of the fuel storage pool water. This

will cause a decrease in water density which would normally result

in an addition of negative reactivity. However, since Boraflex is

not considered to be present, and the fuel storage pool water has a

high concentration of boron, a density decrease causes a positive

reactivity addition. The amount of soluble boron required to offset

this postulated accident was evaluated for the allowed storage

configurations. The amount of soluble boron necessary to mitigate

these accidents and ensure that the Keff will be maintained

less than or equal to 0.95 has been included in the fuel storage

pool boron concentration. Because adequate soluble boron will be

maintained in the pool water, the consequences of a loss of normal

cooling to the fuel storage pool will not be increased.

Therefore, based on the conclusions of the above analysis, the

proposed changes will 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 analyzed.

The potential for criticality accidents in the fuel storage pool

are not new or different types of accidents. It has been reanalyzed

in the Criticality Analysis report (Enclosure 5 [of the proposed

amendment request]).

Because soluble boron has been maintained in the fuel storage

pool water, the possibility of a fuel storage pool dilution has

previously existed. Therefore, the implementation of Technical

Specification controls for the soluble boron will not create the

possibility of a new or different kind of accidental pool dilution.

With credit for soluble boron now a major factor in controlling

criticality, an evaluation of fuel storage pool dilution events was

completed. A generic methodology was applied... to identify

potential events which would dilute the soluble boron contained in

PWR [pressurized water reactor] fuel storage pools, and to quantify

the frequency of those events. This methodology utilized a

probabilistic assessment of a composite plant model to calculate the

event frequency of a dilution event. The results of the assessment

concluded that the event frequency remained less than the NRC Safety

Goal Policy Statement target risk objective of IE-6/reactor year.

Differences between the composite plant described in WCAP-14181

and Vogtle relative to the potential sources of pool dilution were

addressed in an individual analysis of the Vogtle pool. This

analysis was conducted with methodology which closely paralleled

that employed in WCAP-14181. That analysis, found in Enclosure 6 [of

the licensee's proposed amendment request], concluded that the

frequency of pool dilution to the 0.95 Keff boron concentration

(1250 ppm) is on the same order of magnitude as reported in WCAP-

14181 and less than the NRC Safety Goal Policy Statement criterion

of 1.0E-6/reactor year.

Proposed Technical Specifications 3.7.17 and 3.7.18 which ensure

the maintenance of the fuel storage pool boron concentration and

storage configuration, do not represent new concepts. The actual

boron concentration in the fuel storage pool has been maintained at

a higher value than the proposed limits for the Unit 1 and 2 fuel

storage pools for refueling purposes. The criticality analysis

(Enclosure 5 [of the licensee's proposed amendment request])

determined that a boron concentration of 1,100 ppm (Unit 1) and,

1,250 ppm (Unit 2) results in a Keffeff calculations to ensure that the spent fuel rack Keff

will be less than 1.0 with no soluble boron under normal storage

conditions and assuming nominal fuel assembly parameters and fuel

rack dimensions. Soluble boron credit is used to offset

uncertainties, tolerances and off-normal conditions (such as a

misplaced assembly) and to provide subcritical margin such that the

fuel storage pool Keff is maintained less than or equal to

0.95.

The loss of a considerable amount of soluble boron in the fuel

storage pool which could lead to exceeding a Keff of 0.95

during accidents and under adverse conditions has been evaluated and

shown to be very improbable.

The combination of the probabilistic evaluation which shows that

the dilution of the fuel storage pool is a low probability

occurrence, the maximum feasible Keff calculation which shows

that the Keff will remain less than 1.0 when flooded with

unborated water and assuming nominal fuel assembly parameters and

fuel rack dimensions, and the unavailability of the large volumes of

water which are necessary to dilute the fuel storage pool, provide a

level of safety comparable to the conservative criticality analysis

methodology...

Therefore, the proposed changes in this license amendment will

not result in a significant reduction in the plant's 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: 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

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 amendment request: September 25, 1996

Description of amendment request: The proposed amendment would (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).

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 58905]]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The function of the reactor coolant system leakage detection

systems is to detect leakage from the reactor coolant pressure

boundary so that appropriate actions can be taken before the

integrity of the reactor coolant pressure boundary is impaired. In

the plant accident analysis, no credit for mitigation of an accident

is taken for the reactor coolant system leakage detection systems.

These proposed changes do not alter this function, therefore, these

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated?

The function of the reactor coolant system leakage detection

systems is to detect leakage from the reactor coolant pressure

boundary so that appropriate actions can be taken before the

integrity of the reactor coolant pressure boundary is impaired.

These proposed changes do not alter this function; therefore, these

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

accident previously evaluated.

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

The change to allow both the gaseous and particulate

radioactivity monitoring systems to be inoperable at the same time

provided a grab sample is obtained and analyzed at least once per 12

hours is predicated on the availability of the primary leak

detection system (drywell floor sump level monitor system). Since

the gaseous and particulate radioactivity monitoring systems are

backups to the drywell floor sump level monitoring system, allowing

grab samples every 12 hours provides periodic information that is

adequate to detect leakage. The addition of the action to require an

orderly shutdown of the unit for the loss of all reactor coolant

system leakage detection systems does not affect the margin of

safety. Therefore, these proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

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

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

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: October 24, 1996

Description of amendment request: The proposed amendments would

change Technical Specification 3/4.7.1.2, ``Auxiliary Feedwater

System.'' The changes would revise the 18-month surveillances performed

on the system's pumps and valves because testing of the turbine driven

Auxiliary Feedwater pump (TDAFWP) can only be performed in higher modes

when there is sufficient secondary steam pressure.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The changes proposed on the testing of components in the AFW

[Auxiliary Feedwater] System do not affect the operation of the

equipment during conditions when they are required to perform their

safety function. No physical changes to the plant result from the

proposed changes made to the surveillance requirements. The AFW

System is used as a backup system upon loss of main feedwater which

is analyzed as a Condition II event in the UFSAR [Updated Final

Safety Analysis Report] and as such, does not impact the probability

of an accident.

Testing is being performed with the plant in the condition in

which the automatic initiation signals would result, that is, with

the plant in Hot Standby. The changes do not impact the availability

of the AFW System in providing feedwater to the steam generators.

The 24 hour duration to perform testing is sufficiently short that

it is considered unlikely that a condition requiring AFW initiation

would occur with the TDAFWP unable to feed the generators. For such

an occurrence, however, the motor driven AFW pumps would be

available to mitigate the consequences of the event. This time is

less than the 72 hour allowed outage time for an inoperable TDAFWP

in Modes 1-3.

Therefore, the consequences of an accident previously evaluated

are not significantly increased.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve any modifications to

existing plant equipment, do not alter the function of any plant

systems, do not introduce any new operating configurations or new

modes of plant operation, nor change the safety analyses. Testing of

the TDAFWP in Mode 3, Hot Standby, will not impact auxiliary

feedwater capability or impact the ability to maintain Reactor

Coolant temperature. The proposed changes will, therefore, 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.

The changes to the valve surveillance does not decrease the

scope of the existing testing, but will clarify the automatic valves

to be included.

The time in which testing is performed, within 24 hours of

reaching 680 psig steam generator pressure, ensures that testing is

performed in a timely manner after attaining the required steam

pressure. This does not impose a significant safety impact since the

testing is performed within the plant at the zero load conditions

prior to increasing reactor power.

Elimination of the wording ``during shutdown,'' in reference to

the time in which the surveillance is performed, is considered

editorial and is proposed for consistency with the change made to

the pump surveillance requirement.

All changes are consistent with the intent of Salem's current TS

and with the 18 month surveillances specified in NUREG-1431,

Revision 1.

The proposed change, therefore, 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: 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

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

Point Beach Power Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: September 30, 1996 (TSCR 192)

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) Section 15.3.3, ``Emergency Core

Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan

Coolers, and Containment Spray,'' TS 15.3.7, ``Auxiliary Electrical

Systems,'' and the TS Bases to reflect proposed changes to the limiting

conditions for operation, action statements, allowable outage times,

and design specifications for the Point Beach Nuclear Plant (PBNP) TS

associated with the containment

[[Page 58906]]

accident fan coolers, service water equipment, and normal and emergency

power supplies.

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 this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents. Initiating

events for accidents previously evaluated for Point Beach include:

Control rod withdrawal and drop, CVCS [chemical volume and control

system] malfunction (Boron Dilution), startup of an inactive reactor

coolant loop, reduction in feedwater enthalpy, excessive load

increase, losses of reactor coolant flow, loss of external

electrical load, loss of normal feedwater, loss of all AC power to

the auxiliaries, turbine overspeed, fuel handling accidents,

accidental releases of waste liquid or gas, steam generator tube

rupture, steam pipe rupture, control rod ejection, and primary

coolant system ruptures.

This license amendment request proposes to change the limiting

conditions for operation, action statements, allowable outage times,

and design specifications for the Point Beach Nuclear Plant

Technical Specifications associated with the containment accident

fan coolers, service water equipment, and normal and emergency power

supplies.

These proposed changes do not cause an increase in the

probabilities of any accidents previously evaluated because these

changes will not cause an increase in the probability of any

initiating events for accidents previously evaluated. In particular,

these changes affect accident mitigation systems and equipment which

do not cause accidents.

The consequences of the accidents previously evaluated in the

PBNP FSAR [final safety analysis report] are determined by the

results of analyses that are based on initial conditions of the

plant, the type of accident, transient response of the plant, and

the operation and failure of equipment and systems. The changes

proposed in this license amendment request provide appropriate

limiting conditions for operation, action statements, and allowable

outage times for service water, containment cooling and normal and

emergency power supplies.

The proposed changes affect components that are required to

ensure the proper operation of engineered safety features equipment.

The proposed changes do not increase the probability of failure of

this equipment or its ability to operate as required for the

accidents previously evaluated in the PBNP FSAR. The proposed

changes that increase the allowed outage times for engineered safety

features equipment continue to provide appropriate limitations for

these conditions because sufficient equipment is still required to

be operable for accident mitigation and the proposed allowed outage

times are consistent with currently accepted time periods for these

situations.

Therefore, this proposed license amendment does not affect the

consequences of any accident previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors that are used to determine

the consequences of accidents are not being changed.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

New or different kinds of accidents can only be created by new

or different accident initiators or sequences. New and different

types of accidents (different from those that were originally

analyzed for Point Beach) have been evaluated and incorporated into

the licensing basis for Point Beach Nuclear Plant. Examples of

different accidents that have been incorporated into the Point Beach

licensing basis include anticipated transients without scram and

station blackout.

The changes proposed by this license amendment request do not

create any new or different accident initiators or sequences because

these changes to limiting conditions for operation, action

statements, allowable outage times, and design specifications for

service water, containment cooling and normal and emergency power

supplies will not cause failures of equipment or accident sequences

different than the accidents previously evaluated. Therefore, these

proposed Technical Specification changes do not create the

possibility of an accident of a different type than any previously

evaluated in the Point Beach FSAR.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design

and operation of the reactor and containment and the safety systems

that provide their protection.

The changes proposed by this license amendment request provide

the appropriate limiting conditions for operation, action

statements, allowable outage times, and design specifications for

service water, containment cooling and normal and emergency power

supplies. This ensure that the safety systems that protect the

reactor and containment will operate as required. The design and

operation of the reactor and containment are not affected by these

proposed changes. Therefore, the margins of safety for Point Beach

are not being reduced because the design and operation of the

reactor and containment are not being changed and the safety systems

and limiting conditions of operation for these safety systems that

provide their protection that are being changed will continue to

meet the requirements for accident mitigation for Point Beach

Nuclear Plant.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: John N. Hannon

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.

[[Page 58907]]

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

324, Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County,

North Carolina

Date of amendment request: April 2, 1996 (BSEP 96-0123), as

supplemented by an earlier submittal dated November 20, 1995 (BSEP 95-

0535), and by subsequent submittals dated July 1, 1996 (BSEP 96-0242),

July 30, 1996 (BSEP 96-0287), August 7, 1996 (BSEP 96-0300), September

13, 1996 (BSEP 96-0340), September 20, 1996 (BSEP 96-0348), October 1,

1996 (BSEP 96-0362), October 22, 1996 (BSEP 96-0392), October 22, 1996

(BSEP 96-0403), and October 29, 1996 (BSEP 96-0412).

Brief description of amendment: The proposed amendment would modify

Facility Operating Licenses Nos. DPR-71 and DPR-62 and the Technical

Specifications (TS) for the Brunswick Steam Electric Plant, Units 1 and

2, respectively, to authorize an increase in the maximum power level

from 2436 megawatts thermal (MWt) to 2558 MWt.

Date of issuance: November 1, 1996

Effective date: November 1, 1996

Amendment No.: 183 (Unit 1); 214 (Unit 2)

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

revises

Facility Operating License Nos. DPR-71 and DPR-62 and the Technical

Specifications.

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

25698) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 1, 1996.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.

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

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

Date of application for amendment: August 5, 1994, as supplemented

by letters dated November 17, 1994, December 2, 1994, and August 1,

1996.

Brief description of amendment: The amendment revises surveillance

intervals for various systems, components and instruments to

accommodate a 24-month refueling cycle. These revisions are being made

in accordance with the guidance provided by Generic Letter 91-04,

``Changes in Technical Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle.''

Date of issuance: October 30, 1996

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

30 days.

Amendment No.: 187

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 7, 1994 (59 FR

63117) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 30, 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.

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

Buren County, Michigan

Date of application for amendment: December 11, 1995, as

supplemented by letters dated January 15, September 3, October 2,

October 18, and October 25, 1996.

Brief description of amendment: The amendment revises the

Administrative Controls section of the TS by deleting or relocating

requirements that are adequately controlled by existing regulatory

requirements, adding requirements, and editorially restructuring the TS

to be consistent with NUREG-1432, ``Standard Technical Specifications,

Combustion Engineering Plants.'' In addition, containment leak rate

testing requirements are revised to allow the Type A integrated leak

rate test to be scheduled in accordance with Option B of 10 CFR Part

50, Appendix J. Review of several changes proposed by the licensee have

not yet been completed by the staff. The NRC will issue an evaluation

of these changes upon completion of staff review.

Date of issuance: October 31, 1996

Effective date: October 31, 1996

Amendment No.: 174

Facility Operating License No. DPR-20 Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 20, 1996 (61

FR 49493). The October 2, October 18, and October 25, 1996, letters

provided clarifying information and updated TS pages that were within

the scope of the initial application and did not affect the staff's

initial proposed no significant hazards consideration determination.

Therefore, renoticing was not warranted.The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

October 31,1996. No significant hazards consideration comments

received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423

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

County Station, Units 1 and 2, LaSalle County, IllinoisDate of

application for amendments: April 8, 1996, as supplemented on

October 14, 1996.

Brief description of amendments: The amendments revise various

sections of the Technical Specifications (TS) to reflect the transition

of fuel supplier from General Electric (GE) to Siemens Power

Corporation (SPC). The amendments revise the definitions, limiting

conditions for operation, required actions, or surveillance

requirements related to the following fuel thermal limits: Linear Heat

Generation Rate, Critical Power Ratio, Minimum Critical Power Ratio,

and Average Planar Linear Heat Generation Rate. The previous GE

terminology is replaced with vendor independent terms and new, NRC-

approved methodologies are incorporated. The amendments also include

changes to Section 6.0 of the TS to include SPC references, relocate

the requirements for the traversing in-core probe system from the TS to

the Core Operating Limits Report, and revise the fuel description in TS

Section 5.0.

Date of issuance: October 29, 1996

Effective date: Immediately, to be implemented prior to startup of

Cycle 9 for Unit 1 and Cycle 8 for Unit 2.

Amendment Nos.: 116, 101

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

revised the Technical Specifications.

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

25699) The October 14, 1996, submittal provided additional clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

October 29, 1996.No significant hazards consideration comments

received: No

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, IllinoisDate of

application for amendments: August 16, 1996, as supplemented on

October 4, 1996.

Brief description of amendments: The amendments revise the

definition of the F* distance by removing the uncertainty

[[Page 58908]]

term from the specified distance and removing the footnote which

specifies the time frame for which it is applicable.

Date of issuance: November 6, 1996

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

Amendment Nos.: 174, 161

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47968) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 6, 1996No significant

hazards consideration comments received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station Units 1 and 2, Lake County, Illinois

Date of application for amendments: September 3, 1996

Brief description of amendments: The amendments incorporate revised

installation procedures for steam generator tube sleeves designed by

ABB Combustion Engineering (ABB/CE).

Date of issuance: October 29, 1996

Effective date: October 29, 1996

Amendment Nos.: 173 and 160

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 11, 1996 (61

FR 47966) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 29, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

Michigan Date of application for amendment: September 5, 1996 (NRC-

96-0075), as supplemented by letters dated October 14, October 23,

October 29, and October 31, 1996

Brief description of amendment: The amendment revises Technical

Specification (TS) 2.1.2 to incorporate cycle-specific safety limit

minimum critical power ratios (SLMCPRs) for the core that will be

loaded for Cycle 6. In addition, TS 3.4.1.1 is revised to delete the

specific SLMCPR number and replace it with a reference to TS 2.1.2.

Date of issuance: November 5, 1996

Effective date: November 5, 1996, with full implementation within

45 days

Amendment No.: 109

Facility Operating License No. NPF-43 Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: September 25, 1996 (61

FR 50342) The letters of October 14, 23, 29, and 31, 1996, provided

clarifying information and were not outside the scope of the initial

proposed no significant hazards consideration determination. The

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

Safety Evaluation dated November 5, 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, et al., Docket Nos. 50-413 and 50-414, Catawba

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

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

supplemented by letters dated May 16 and August 29, 1996

Brief description of amendments: The amendments modify the

Technical Specifications for diesel generators to incorporate guidance

and recommendations contained in NRC Generic Letter (GL) 93-05, ``Line-

Item Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' GL 94-01, ``Removal

of Accelerated Testing and Special Reporting Requirements for Emergency

Diesel Generators,'' NUREG-1431, ``Revised Standard Technical

Specifications for Westinghouse PWRs,'' and NUREG-1366, ``Improvements

to Technical Specifications Surveillance Requirements.''

Date of issuance: October 30, 1996

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

30 days

Amendment Nos.: 155 and 147

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

revised the Technical Specifications.

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

31175) The August 29, 1996, submittal provided additional information

that did not change the scope of the December 14, 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 October 30, 1996. No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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

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

CarolinaDate of application for amendments: August 12, 1996, as

supplemented by letter dated September 10, 1996

Brief description of amendments: The amendments revise the

Technical Specifications associated with the containment leak-rate

tests by implementing 10 CFR Part 50, Appendix J, Option B, for Type A

leak-rate testing.

Date of issuance: October 30, 1996

Effective date: As of the date of issuance to be implemented 30

days from the date of issuance.

Amendment Nos.: 218, 218, 215

Facility Operating License Nos. DPR-38, DPR-47 and DPR-55:

Amendments revise the Technical Specifications.

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

44356) The September 10, 1996, letter provided additional information

that did not change the scope of the August 12, 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 October 30, 1996.No significant hazards

consideration comments received: No

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: April 15, 1996 (TSCR No. 244)

Brief description of amendment: The amendment revises Specification

5.3.1.B to allow the shield plug and the associated lifting hardware to

be moved over irradiated fuel assemblies that are in a dry shielded

canister within the transfer cask in the cask drop protection system.

Date of Issuance: November 7, 1996, to be implemented within 30

days of issuance

Effective date: November 7, 1996

Amendment No.: 187

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

Technical Specifications

[[Page 58909]]

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

20849) The Commission's related evaluation of this amendment and final

determination of no significant hazards consideration addressing

comments received on the proposed no significant hazards consideration

determination are contained in a Safety Evaluation dated November 7,

1996.No significant hazards consideration comments received: Yes.

Local Public Document Room location: Ocean County Library,

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

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear

Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

Date of application for amendment: February 6, 1995

Brief description of amendment: This amendment revised the

Technical Specifications by extending the surveillance interval to

demonstrate operability of the containment airlocks from quarterly to

annually and to decrease the personnel exposure with implementing the

surveillance.

Date of issuance: October 24, 1996

Effective date: October 24, 1996

Amendment No.: 51Possession-Only License No. DPR-73: The amendment

revised the Technical Specifications.

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

28616) The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

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

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

Matagorda County, Texas

Date of amendment request: May 1, 1995, as supplemented by letters

dated June 22, August 28, November 22, and December 19, 1995, and

January 4, 8 (two letters), and 23, June 27, July 9, August 8, and

September 23, 1996.

Brief description of amendments: The amendments allowed extension

of the standby diesel generator allowed outage time to 14 days, and

extension of the essential cooling water loop and the essential chilled

water loop allowed outage times to 7 days. The amendments also added to

Administrative Controls a description of the Configuration Risk

Management Program (CRMP) used to assess changes in core damage

probability resulting from applicable plant configurations.

Date of issuance: October 31, 1996

Effective date: October 31, 1996, to be implemented within 30 days

Amendment Nos.: 85 and 72

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 31, 1996 (61 FR

40019) The additional information contained in the supplemental letters

dated August 8 and September 23, 1996, were clarifying in nature and

thus, within the scope of the initial notice and did not affect the

staff's proposed no significant hazards consideration determination.The

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

Safety Evaluation dated October 31, 1996.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

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: February 22, 1996, and

supplemented July 22, 1996

Brief description of amendments: The amendments revise the

administrative controls section of the technical specifications to

change the operator license requirements for operations management.

Date of issuance: October 29, 1996

Effective date: October 29, 1996, with full implementation within

45 days

Amendment Nos.: 212 and 197

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

revised the Technical Specifications.

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

13527) The July 22, 1996, submittal was more restrictive than the

original submittal and did not change the staff's original no

significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated October 29, 1996.No significant hazards consideration

comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

ConnecticutDate of application for amendment: August 27, 1996

Brief description of amendment: The Technical Specification (TS)

amendment clarifies the limiting condition for operation and

surveillance requirements to ensure that the appropriate number of

charging pumps and high pressure safety injection pumps are operable

for reactivity control and reactor coolant system (RCS) makeup

requirements, while also limiting the number of operable pumps to

ensure that the low temperature overpressure limits will not be

exceeded in the event of a mass addition to the RCS during shutdown

conditions. The TS Bases remain unchanged as the result of this

amendment.

Date of issuance: October 25, 1996

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

within 30 days.

Amendment No.: 205

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

Technical Specifications.

Date of initial notice in Federal Register: September 20, 1996 (61

FR 49498) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 25, 1996No 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

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

A. FitzPatrick Nuclear Power Plant, Oswego County, New YorkDate of

application for amendment: March 22, 1996, as supplemented October

11, 1996

Brief description of amendment: The amendment proposed changes to

the Technical Specifications to establish operability requirements for

avoidance and protection from thermal hydraulic instabilities to be

consistent with Boiling Water Reactor Owners Group long-term solution

Option I-D. Editorial changes are also made to support the revised

specifications, improve readability of Bases sections, and enhance the

presentation of requirements for single loop operation.

Date of issuance: October 30, 1996

[[Page 58910]]

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

30 days.

Amendment No.: 236

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

Technical Specifications.

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

20854) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 30, 1996No 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.

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: May 30, 1996, as supplemented by

letter dated October 11, 1996

Brief description of amendment: The amendment proposes to eliminate

selected response time testing requirements for certain sensors and

specified loop instrumentation.

Date of issuance: October 28, 1996

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

30 days.

Amendment No.: 235

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

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34896) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 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: July 12,

1996, as supplemented September 12, 1996

Brief description of amendments: The amendments revise Technical

Specification Table 3.3-3, ``Engineered Safety Feature Actuation System

Instrumentation,'' to clarify the setpoint for the interlock designated

P-12.

Date of issuance: November 4, 1996

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

implemented within 30 days.

Amendment Nos. 185 and 167

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 23, 1996 (61 FR

38229) The supplemental letter provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination nor the Federal Register notice.The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

November 4, 1996.No significant hazards consideration comments

received: No

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

West Broadway, Salem, NJ 08079

Dated at Rockville, Maryland, this 13th day of November 1996.

FOR THE NUCLEAR REGULATORY COMMISSION

Steven A. Varga,

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

Regulation

[FR Doc. 96-29584 Filed 11-18-96; 8:45 am]

BILLING CODE 7590-01-F