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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 12, 1994, through January 5, 1995.

The last biweekly notice was published on January 4, 1995 (60 FR 494).

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 [[Page 3670]] 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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, 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 February 17, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves

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, 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, 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

[[Page 3671]] 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: November 22, 1994.

Description of amendment request: The proposed amendment would

revise the allowable leak rate for the main steam isolation valves

(MSIVs) from the current 11.5 standard cubic feet per hour (scfh) for

each valve, to a maximum combined main steam line leak rate of 46 scfh.

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 operation of Pilgrim Station in accordance with the

proposed Amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment does not involve a change to structures,

components, or systems which would affect the probability of an

accident previously evaluated in the Pilgrim Updated Final Safety

Analysis Report (UFSAR). The proposed amendment results in no change

in radiological consequences of the design basis LOCA [loss-of-

coolant accident] as currently analyzed for Pilgrim Station. These

analyses were calculated using the combined total leakage factor of

46 scfh for determining acceptance to the regulatory limits for the

offsite, control room, and Technical Support Center (TSC) doses as

contained in 10CFR100 and 10CFR50, Appendix A, GDC 19. The proposed

change does not compromise existing radiological equipment

qualification, since the combined total leakage rate of 46 scfh has

been factored into our existing equipment qualification analyses for

10 CFR 50.49.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

There is no modification to the MSIVs or other plant system or

structure associated with this amendment which could impact their

capability to perform their design function. The total MSIV leakage

rate of 46 scfh is included in the current radiological analyses for

the assessment of dose exposure following an accident. This proposal

changes the allowable leakage rate from a per valve to a total

combined line leakage acceptance criteria but does not change the

cumulative allowable value. Therefore, the proposed change does not

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

any accident previously analyzed.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

The allowable leak rate limit specified for the MSIVs is used to

quantify the maximum amount of bypass leakage assumed in the LOCA

radiological analysis. Results of the analysis are evaluated against

the dose guidelines contained in GDC [General Design Criteria] 19

and 10CFR100. The margin of safety in this context is considered to

be the difference between the calculated dose exposures and the

guidelines provided by the GDC 19 and 10CFR100. Therefore, since the

maximum allowable leakage for each valve was assumed and used as the

total allowable leakage for the purpose of calculating potential

dose, the margin of safety is not affected because the dose levels

remain the same.

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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: November 22, 1994.

Description of amendment request: The proposed amendment would

revise the mode conditions under which the Scram Discharge Instrument

Volume-Scram Trip Bypass in Table 3.2.C.1 is required to be operable

and changes the associated functional test frequency from quarterly to

once per operating cycle in Table 4.2.C.

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 operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to Table 3.2.C.1, and the associated change

to Table 4.2.C, removes incorrect reactor modes listed for the Scram

Discharge Instrument Volume (SDIV)--Scram Trip Bypass function. The

Pilgrim control rod block logic for the SDIV Bypass is not operable

nor is it required by design when in the Run and Startup modes. The

control logic and the FSAR [Final Safety Analysis Report] (section

7.2.3.10) specifies SDIV--Scram Trip Bypass operability only in the

Refuel and Shutdown modes.

This change will not result in any physical modification or

operation of the control rod block system. The change conforms the

technical specifications to the actual design of the SDIV Scram Trip

Bypass as described in the FSAR. Changing the functional

surveillance frequency from quarterly to once per operating cycle

also conforms the technical specifications to the applicable mode

for the function.

The change is classified as an administrative change because it

corrects an administrative requirement that does not reflect the

logic design. It improves safety by removing the need to install

jumpers during reactor operations to perform unnecessary and

potentially risky functional surveillances.

Therefore, because this is an administrative change, operation

of Pilgrim will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated

because it is administrative and requires no physical alteration of

the plant configuration, changes to setpoints, or operating

parameters.

3. The operation of Pilgrim in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety.

The proposed change serves to enhance the margin of safety by

eliminating the potential for error caused by installing jumpers to

the control logic during reactor operation. Changing the functional

surveillance frequency from quarterly to once per operating cycle

also enhances the margin of safety by allowing test performance off-

line, the mode for which the SDIV scram trip bypass control rod

blocks are designed to be operable.

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: Plymouth Public Library, 11

[[Page 3672]] North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: November 22, 1994

Description of amendment request: The proposed amendment would

revise the suppression chamber water level operating range, increasing

it 2 inches, and revise the water level recorder range.

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 operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously identified.

The probability of an accident is not increased by this proposed

change because there is no relation between the Suppression Chamber

water level operating range and the probability of an accident.

The consequences of an accident identified are not increased.

The Suppression Chamber is an accident mitigating device. Increasing

the water level operating range has been analyzed and does not

significantly increase the structural loads and the calculated

stress levels remain within Mark 1 Acceptance Criteria.

We have reviewed the FSAR [Final Safety Analysis Report]

Containment Analyses and concluded that the safety margin is not

affected. An increase in water level enhances the Suppression Pool's

ability to mitigate an accident by providing more water for use by

emergency cooling systems. The higher water level increases the sink

capabilities resulting in lower torus water temperatures from steam

blowdowns. There is a minor reduction in the free air volume of the

torus which has a negligible effect on containment post accident

pressures. Therefore, there is no significant increase in the

probability or consequences of an accident previously identified.

The change in water level recorder range does not involve an

increase in the probability or consequence of an accident because

the new recording range accounts for instrument loop uncertainties

and is thus more conservative than the previous range.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously analyzed.

An increase in the Suppression Chamber water level operating

range does not create a new or different kind of accident from any

accident previously analyzed because the Suppression Chamber is an

accident mitigating device. The Suppression Chamber serves as the

heat sink for any postulated transient or accident condition when

the primary heat sink (main condenser) is unavailable and as a

source of water for the Core Standby Cooling Systems. The structural

affects of the increase in water volume have been analyzed and do

not significantly effect the Mark 1 containment loads.

Revising the water level recording range is more conservative

than that previously used and does not create the possibility of a

new or different kind of accident.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in the

margin of safety.

Operation with an increased Torus water level does not affect

the structure and attached piping of the Pilgrim Suppression Chamber

and does not significantly affect the calculated stress levels;

therefore, there is no significant reduction in the margin of

safety.

The change in the water level recording range is due to

replacing the transmitter with a smaller span. The change from 0 to

32 inches to -7 to +7 inches enhances resolution and accuracy of the

water level instrument loop.

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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: August 30, 1994.

Description of amendment request: The proposed amendment relocates

refueling cycle specific parameters from the technical specifications

to the Core Operating Limits Report as per recommendations promulgated

by NRC Generic Letter 88-16. Additionally, the amendment adds a 24 hour

limit on operations when only one reactor coolant pump is operating in

each loop. 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:

Criterion 1--Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The relocation of cycle-specific variables from the Technical

Specifications to the Core Operating Limits Report (COLR) is

considered to be administrative in nature and has no impact on plant

operation or safety. The Technical Specifications will continue to

require operation within the core operational limits for each cycle

reload as calculated by the NRC approved reload methodologies. The

values and setpoints placed in the COLR are addressed in the reload

report for each particular fuel cycle. The reload report presents

the results of evaluations of accidents addressed in the ANO-1

Safety Analysis Report. These evaluations demonstrate that changes

in the fuel cycle design and the corresponding COLR do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The revision of Specification 3.1.1.1.a and addition of the

footnote to Table 2.3-1 result in additional restrictions on

operation with one reactor coolant pump in each loop with the

reactor critical. This more restrictive specification limits

operation with one reactor coolant pump in each loop to a 24 hour

period when the reactor is critical. This change incorporates a more

restrictive control and does not affect any previously analyzed

event.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously.

Criterion 2--Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

This relocation of cycle-specific variables from the Technical

Specifications to the COLR does not create the possibility of a new

or different kind of accident from any previously analyzed. The

cycle-specific variables will continue to be calculated using NRC

approved methodologies. Technical Specifications will continue to

require operation within the required core operating limits and

appropriate actions will be taken if the limits are exceeded.

Because plant operation continues to be limited in accordance with

the values of cycle-specific parameter limits that are established

using NRC approved methodologies, the relocations included in this

submittal are considered to be administrative in nature and have no

impact on plant safety as a consequence.

The revision of Specification 3.1.1.1.a and addition of the

footnote to Table 2.3-1 result in additional restrictions on

operation with one reactor coolant pump in each loop with the

reactor critical. This more restrictive specification limits

operation with one reactor coolant pump in each loop to a 24 hour

period when the reactor is critical. This proposed change introduces

no new mode of plant operation.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the

Margin of Safety. [[Page 3673]]

The proposed relocations are considered to be administrative in

nature and do not involve a significant reduction in the margin of

safety since they only involve transferring limits from the

Technical Specifications to the COLR. The values and setpoints

placed in the COLR are addressed in the reload report for each

particular fuel cycle. The development of limits for future reloads

will continue to conform to methodologies described in NRC approved

documentation. Each future reload involves a 10CFR50.59 safety

review to assure that operation of the unit within the cycle-

specific limits will not involve a significant reduction in the

margin of safety.

The revision of Specification 3.1.1.1.a and addition of the

footnote to Table 2.3-1 result in additional restrictions on

operation with one reactor coolant pump in each loop with the

reactor critical. This more restrictive specification limits

operation with one reactor coolant pump in each loop to a 24 hour

period when the reactor is critical. This change does not involve a

significant reduction in the margin of safety, rather, it

constitutes an additional limitation not previously included in the

Technical Specifications.

Therefore, this change does not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502.

NRC Project Director: William D. Beckner.

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

Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: November 9, 1994.

Description of amendment request: The proposed amendment revises

those specifications associated with various engineered safety feature

systems following a design basis fuel handling accident. The proposed

changes affect conditions where irradiated fuel is handled in the

primary or secondary containment and when fuel is handled over the

reactor vessel with fuel in the vessel. These changes are based on a

recent re-analysis of the fuel handling accident for Grand Gulf Nuclear

Station (GGNS).

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 proposed definition of RECENTLY IRRADIATED fuel is used to

establish operational conditions where specific activities represent

situations where significant radioactive releases can be postulated.

These operational conditions are consistent with the design basis

analysis. Because the equipment affected by the revised operational

conditions is not considered an initiator to any previously analyzed

accident, inoperability of the equipment cannot increase the

probability of any previously evaluated accident. The proposed

applicability 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 19.

Therefore, the proposed changes do not significantly increase the

probability or 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 proposed definition is used to establish operational

conditions where specific activities represent situations where

significant radioactive releases can be postulated. These

operational conditions are consistent with the design basis

analysis. The proposed changes do not introduce any new modes of

plant operation and do not involve physical modifications to the

plant. Therefore, the proposed changes do not create the possibility

of a new or different kind of accident from any previous analyzed.

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 revised definition is used to establish operational

conditions where specific activities represent situations where

significant radioactive releases can be postulated. These

operational conditions are consistent with the design basis analysis

and are established such 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

proposed change only eliminates the excess margin from the analysis.

The current margin of safety is retained.

Specifically, the margin of safety for the fuel handling

accident is the difference between the 10 CFR 100 limits 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 10 CFR 100 limits and the licensing limit

defined by 10 CFR 50, Appendix A, Criterion 19 (GDC 19). Excess

margin is the difference between the postulated doses and the

corresponding licensing limit.

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.

Therefore, the proposed changes do not result in a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: William D. Beckner.

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: December 6, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to allow the use of the Combustion

Engineering sleeving process for repairing steam generator tubes. (The

current requirement specifies that degraded steam generator tubes be

repaired by plugging.)

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. The NRC has reviewed the licensee's analysis against the

standard of 10 CFR 59.92(c). The staff's review is presented below:

1. The proposed amendment would not involve a significant increase

in the [[Page 3674]] probability or consequences of an accident

previously evaluated.

With the sleeve dimensions, materials, and connecting joints

designed to the applicable American Society of Mechanical Engineers

Boiler and Pressure Vessel Code (ASME Code), the proposed sleeving

repair becomes an in-kind substitution for the steam generator tube

being repaired. The design criteria for the sleeves conform to the

stress limits and safety margins of Code Section III. Safety factors of

3 (normal operation) and 1.5 (accident conditions) were applied to the

sleeve design. Mechanical testing using Code stress allowables also has

been performed in support of the sleeve design. Based on the results of

vendor test and analysis programs, the sleeves fulfill their intended

function as leak tight structural members and meet or exceed all design

criteria.

Evaluation of the steam generator tubes and proposed sleeves

indicates no detrimental effects on the sleeve or sleeve-tube assembly

from reactor coolant system flow, reactor or steam generator coolant

chemistry, or thermal or pressure conditions (including transients)

that may be experienced by the Maine Yankee plant. Corrosion testing of

sleeve-tube assemblies indicates no evidence of sleeve or steam

generator tube corrosion considered detrimental under anticipated

service conditions.

Installation of the proposed sleeves will be controlled via

Combustion Engineering's proprietary equipment and process. The process

has been used 24 separate times since 1984 to install approximately

4100 steam generator sleeves in nuclear facilities worldwide. The Maine

Yankee steam generator design has been reviewed and found compatible

with the sleeve installation equipment and process. Installation of the

proposed sleeves will have no significant effect on either plant

configuration or operation.

The licensee therefore concludes that implementation of the

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The proposed amendment would not create the possibility of a new

or different kind of accident from any accident previously evaluated.

As discussed above, the structural integrity, thermal

characteristics, and material properties of the proposed sleeves are

compatible with Maine Yankee's steam generators. Therefore, the

functions of the steam generators will not be significantly affected by

installation of the proposed sleeves. In addition, the proposed sleeves

do not interact with any other plant systems. Finally, the continued

integrity of installed sleeves is periodically verified by the steam

generator inspections required by plant Technical Specifications.

The licensee therefore concludes that implementation of the

proposed change will not create a new or different kind of accident

from any accident previously evaluated.

3. The proposed amendment would not involve a significant reduction

in a margin of safety.

Repair of degraded steam generator tubes via the use of the

proposed sleeves has been confirmed to restore the structural integrity

of faulted tubes under normal operating and postulated accident

conditions. The design safety factors used for the sleeves are

consistent with ASME Code safety factors required in the design of

Maine Yankee's steam generators. The repair limit for the proposed

sleeves is consistent with that established for Maine Yankee's steam

generators. The design of the sleeve-to-tube joint has been verified by

testing to preclude significant leakage during normal and postulated

accident conditions. Use of the previously identified design safety

factors design verification testing assures that margin to safety with

respect to installation of the proposed sleeves is not significantly

different from the original steam generator tubes.

The licensee therefore concludes that implementation of the

proposed change would not involve a significant reduction in a margin

of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

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

NRC Project Director: Walter R. Butler.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: December 23, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) 2.1.2, ``Fuel Cladding

Integrity,'' 3.6.2/4.6.2, ``Protective Instrumentation,'' and

associated Bases to extend the calibration frequency of the reactor

recirculation flow transmitters from once per quarter to once per

operating cycle and for the square rooters and summers from once per

quarter to once per year. The proposed amendment would revise the flow

biased average power range monitor (APRM) scram and rod block,

recirculation flow comparator, and flow unit upscale setpoints and the

associated Bases of TSs 2.1.2, 2.2.2, and 3.6.2/4.6.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes extend the calibration interval for the

recirculation flow square rooters, summers and transmitters and

revise the setpoints for the recirculation flow upscale and

comparator rod block trips. The associated analytical limits for

APRM flow biased scram and rod block increase by 2% and 8%

respectively. Setpoints are for plant protective functions (i.e.,

scram and rod block) which respond to an accident or transient. The

scram and rod block function responds to mitigate the consequences

of an accident or transient. Therefore, a change to the setpoints

cannot increase the probability of these accidents or transients.

Likewise, changes to surveillance intervals for the protective

functions which respond to an accident or transient cannot increase

the probability. In fact, the proposed increase in the surveillance

intervals reduce the probability of an inadvertent scram by reducing

the duration that the plant is in the one-half scram condition.

The new surveillance intervals, setpoints and allowable setpoint

deviations are calculated using the approved GE [General Electric

Company] setpoint methodology documented in NEDC-31336. The

methodology in NEDC-31336 provides assurance that safety system

actuation (i.e., reactor scram or control rod withdrawal block) will

occur prior to the associated system parameters (neutron flux and

recirculation flow) exceeding their analytical limits. Based upon

re-evaluation of NMP1 [Nine Mile Point Nuclear Station Unit No. 1]

accidents and transients, it has been shown that the fuel thermal

limits are not significantly impacted. Therefore, the consequences

of an accident or transient has not significantly increased.

Thus, plant response to previously analyzed accidents remains

within previously determined limits. Therefore, the operation of

Nine Mile Point Unit 1, in accordance with the proposed amendment,

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

[[Page 3675]]

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes to extend the calibration frequency do not

represent a physical change to the plant as described in the NMP1

Final Safety Analysis Report (Updated). However, this change results

in increasing the analytical limits for the APRM flow based scram

and rod block by 2% and 8% respectively. The proposed changes do not

alter the plant configuration and the initial conditions used for

the design basis accident analysis are still valid. Thus, no

potential initiating events are created which would cause any new or

different kinds of accidents. As such, the plant initial conditions

utilized for the design basis accident analysis are still valid.

Therefore, operation of Nine Mile Point Unit 1 in accordance with

the proposed change will not create the possibility of a new or

different kind of accident from any previously assessed.

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The analytical limits for the APRM flow biased scram and rod

block increase by 2% and 8% respectively. The trip units in the APRM

and recirculation flow instrumentation systems will continue to be

calibrated every three months. In addition, the entire APRM and

recirculation flow instrumentation systems will still be subject to

Instrument Channel Tests every three months. These tests, together

with the calibration of the flow square rooters and summers once per

year and the flow transmitters once per operating cycle, will assure

that system reliability and availability are maintained at their

current levels. Reanalysis of the design basis transients was

performed utilizing these new values. The results showed that the

increase had an insignificant effect on the consequences of these

events. Therefore, the proposed amendment will not involve a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Michael J. Case

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: December 13, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification Table 3.6.1.2-1, ``Allowable Leak Rates

Through Valves in Potential Bypass Leakage Paths,'' to increase the

maximum allowable leakage rate of each of the eight main steamline

isolation valves from 6.0 scfh to 24.0 scfh.

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 operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes to Technical Specification Table 3.6.1.2-1

would allow a maximum leakage of 24.0 scfh for each of the eight

MSIVs [main steamline isolation valves]. The current Technical

Specifications allow a maximum leakage for an MSIV of 6.0 scfh.

Closure of one or more of the MSIVs at rated power is a pressure

transient for the reactor coolant pressure boundary. This pressure

transient is evaluated in Section 15.2.4 of the USAR [Updated Safety

Analysis Report]. Closure of MSIV(s), as analyzed in the USAR, could

occur due to manual or automatic actions. A change to the leakage

limit for the MSIVs does not affect either the manual or automatic

actions that would close the MSIVs. Therefore, the proposed change

to the table cannot affect the probability of the closure of one or

more MSIVs at rated power.

The radiological evaluation of the DBA-LOCA [Design Basis

Accident--Loss-of-Coolant Accident] incorporates a maximum leakage

of 24.0 scfh for each of the four main steam lines. In addition, the

revised radiological evaluation includes the impact of the proposed

license amendment currently under review by the Staff which would

increase the rated operation of NMP2 from 3323 to 3467 megawatts

thermal (see NMPC letter dated July 22, 1993 to the NRC). The

revised radiological evaluation also includes the impact of License

Amendment No. 56 (see NMPC letter dated July 1, 1994 to the NRC and

License Amendment No. 56, dated August 30, 1994).

The new doses from the revised radiological analysis for a DBA-

LOCA, as shown in Table 1 [of December 13, 1994, amendment request],

continue to remain below 10 CFR [Part] 100 guideline values and GDC

[General Design Criterion] 19 limits. The impact of the increased

MSIV leakage on vital area access and equipment qualification is

minimal and acceptable. Therefore, operation with the proposed

change to the Technical Specifications will not significantly

increase the consequences of an accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The safety function of the MSIVs is to isolate the main steam

lines in a timely manner to preclude the uncontrolled leakage of

radioactive steam. This is accomplished by providing the MSIVs with

the capability of rapidly closing automatically in response to

various plant conditions. The increase in the leakage limit for the

MSIVs from 6.0 scfh to 24.0 scfh will not inhibit the MSIVs'

isolation function. Therefore, operation with the proposed increase

in the MSIV leakage will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The revised radiological analysis follows the very conservative

fuel failure and instantaneous release assumptions of RG [Regulatory

Guide] 1.3, with the exception of regulatory position C.1.f as

permitted by SRP [Standard Review Plan] Section 6.5.5, ``Pressure

Suppression Pool as a Fission Product Cleanup.'' The Staff approved

the use of SRP Section 6.5.5. as part of the licensing basis of NMP2

in License Amendment No. 56.

The revised radiological analysis incorporates the maximum

allowable leakage limit of 24.0 scfh for each of the four main steam

lines. The revised radiological analysis also includes the impacts

of the proposed power uprate of NMP2 and License Amendment No. 56.

The new doses from the revised radiological analysis remain below

the Staff acceptance criteria of 10 CFR [Part] 100 guideline values

and GDC 19 (see Table 1 [of December 13, 1994, amendment request]).

Therefore, operation with the proposed changes to the Technical

Specifications will not significantly reduce a margin of safety.

Accordingly, as determined by the analysis above, this proposed

amendment involves no significant hazards consideration.

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: Reference and Documents

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

York 13126.

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

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

NRC Project Director: Michael J. Case. [[Page 3676]]

Philadelphia Electric Company, Public Service Electric and Gas Company,

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

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

Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: August 3, 1994.

Description of amendment request: The proposed changes would delete

a footnote in the Technical Specifications (TS) regarding snubber

functional testing frequency and make permanent the current one-time

snubber functional test frequency of 24 months.

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, because the probability of a seismic or other dynamic

event is independent of the surveillance period for snubber tests.

The change does not introduce any failure mechanisms to the

previously considered events. The consequences of an accident

previously evaluated in the SAR [Safety Analysis Report] is not

increased by the proposed revision to [t]he snubber TS. No physical

changes are being made to the plant. The snubbers' role in

mitigating the consequences of an accident is to provide restraint

during seismic or other dynamic events while permitting the slow

movement of piping and components during heatup and cooldown. The

proposed TS change will not affect the snubbers ability to continue

to perform this role for the following reasons: (1) Changing the

inspection cycle to 24 months will not reduce the ability of the

functional testing to confirm the operability of the snubber

population. The original interval of 18 months was selected to

accommodate the need to test snubbers that were inaccessible during

normal operation. Since snubbers do not require preventative

maintenance during the operating cycle, the additional time added by

a 24 month operating cycle has minimal impact, if any, on snubber

operability. (2) The requirement to monitor service life remains

part of TS. The review of snubber service life records is a

documentation review of the snubbers service life. If a snubber's

service life would expire prior to the next scheduled review then

the snubber is reconditioned, replaced or reevaluated to extend its

service life. (3) Snubber functional testing has shown no failure

mechanism which would be aggravated by an extension of the test

interval to 24 months. A historical search of completed snubber

functional STs was completed. The historical search indicated that

even though the snubbers did not always meet the initial screening

functional test criteria of the ST, the piping system was operable

based on an engineering evaluation and there was no evidence of a

time dependent failure mechanism. To ensure the snubber remains

operational during the next operating cycle, snubbers not meeting

the screening ST acceptance criteria are either replaced or

reconditioned.

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

or different kind of accident from any previously evaluated because

the proposed change does not involve operational procedure or

physical changes to the plant. Since snubbers will continue to meet

their design basis of protecting the piping and equipment during

dynamic events, the possibility of a different type of accident will

not be created.

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

in a margin of safety. There may be a slight increase, if any, in

the possibility of undetected snubber failures because of the

increase in the interval of functional testing for snubbers;

however, the historical data of previous snubber functional

surveillance testing and the supporting engineering evaluations

indicate that on those occasions where snubbers did not meet initial

surveillance testing requirements, the piping systems were all

operable. Therefore, the probability of occurrence of a malfunction

of equipment is minimal and equipment important to safety (ITS) that

use snubbers will continue to meet design requirements and the

margin of safety will be unaffected.

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: Government Publications

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

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

Pennsylvania 17105.

Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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

Generating Station, Salem County, New Jersey

Date of amendment request: September 29, 1994.

Description of amendment request: This amendment requests revision

of Table 4.3.6-1 ``Control Rod Block Instrumentation Surveillance

Requirements.'' The channel calibration frequencies for the Source

Range Monitor (SRM) and the Intermediate Range Monitor (IRM) would be

changed as follows: the up-scale and the down-scale trip functions on

each instrument would be changed from Note ``SA'', once-per-184 days to

note ``R'', once-per-refuel interval.

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes involve no hardware changes, no changes to

the operation of any systems or components, and no changes to

existing structures. The revision of channel calibration frequencies

for the SRM and IRM trip function portion of the control rod block

instrumentation represent changes that do not affect plant safety

and do not alter existing accident analyses.

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

accident from any previously evaluated.

The proposed changes are procedural in nature concerning the

calibration frequency of instrumentation that have historically

shown little set point drift. The channel calibration methodology

for the SRM and IRM control rod block trip functions remain

unchanged. The proposed changes while slightly increasing the

possibility of an undetected instrument error will not create a new

or unevaluated accident or operating condition.

3. Will not involve a significant reduction in a margin of

safety.

The proposed changes are in accordance with recommendations

provided by the NRC regarding the improvement of Technical

Specifications. These changes will result in the perpetuation of

current safety margins while reducing regulatory burden and

decreasing equipment degradation.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070.

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

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

NRC Project Director: John F. Stolz.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: December 16, 1994. [[Page 3677]]

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.4 by removing the Limiting Conditions for Operation (LCOs) for the

Turbine Overspeed Protection System (TOPS). Tables TS 4.1-1 and TS 4.1-

3 would also be revised to remove the surveillance requirements for the

TOPS instrumentation and turbine valves. The TOPS and related

requirements would be relocated to the Updated Safety Analysis Report

(USAR).

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:

Significant Hazards Determination for Proposed Changes to TS

3.4.c and Table TS 4.1-1 and Associated Bases Changes

In accordance with 10 CFR Part 50, Section 50.91 and using the

standards provided in Section 50.92, the proposed change has been

reviewed to determine that no significant hazards exist as a result

of this change. The analysis showed:

(1) The proposed amendment will not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

The purpose of the Turbine Overspeed Protection System (TOPS) is

to prevent an overspeed event, which is a precursor to a potential

turbine-generated missile. Neither Transient Analyses nor Design

Basis Accidents (DBAs) evaluated in the accident analyses contained

in Chapter 14 of the Kewaunee Nuclear Power Plant (KNPP) Updated

Safety Analysis Report (USAR) assume operation of the TOPS. The

calculations and probabilities associated with USAR section 14.2.7,

``Turbine Missile Damage to the Spent Fuel Pool,'' are not affected

by this amendment. This amendment does not implement physical

changes to the plant and does not change the KNPP's existing

requirements. As a result, this change will not increase the

probability of a previously evaluated accident.

The purpose of the TOPS is preventative and it serves no

function to mitigate the consequences of any accident previously

evaluated. Therefore, removing the requirements associated with the

TOPS from the TSs will not affect the consequences of an accident

previously evaluated.

(2) The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

This amendment does not involve any changes in the operational

characteristics of the surveillance tests and will impose no new

requirements. This change will simply relocate the same testing

requirements from the KNPP Technical Specifications to the KNPP

USAR. Since this change is administrative in nature, it will not

create a new or different kind of accident from any accident

previously evaluated.

(3) The proposed amendment will not involve a significant

reduction in the margin of safety.

KNPP's USAR section 14.2.7, ``Turbine Missile Damage to the

Spent Fuel Pool,'' will not be affected by this amendment.

Relocating the TOPS and related requirements is a change that is

administrative in nature and does not alter the intent of any

requirements. Therefore it can be concluded that this change will

not involve a significant reduction in the margin of safety.

Significant Hazards Determination for Proposed Change to Table

TS 4.1-3 and associated Basis Change

In accordance with 10 CFR Part 50, Section 50.91 and using the

standards provided in Section 50.92, the proposed change has been

reviewed to determine that no significant hazards exist as a result

of this change. The analysis showed:

(1) The proposed amendment will not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

This amendment does not involve any changes in the operation or

frequency of the turbine valve tests. This amendment will simply

relocate the turbine valve testing requirements from the Kewaunee

Nuclear Power Plant's (KNPP's) Technical Specifications (TSs) to the

Updated Safety Analysis Report (USAR). This change is administrative

in nature and therefore will not involve a significant increase in

the probability or consequence of an accident previously evaluated.

(2) The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

This amendment is administrative in nature and will not change

any requirements. This change will simply relocate the requirements

from the KNPP TSs to the USAR. The purpose of the turbine stop and

governor valves is to control steam flow to the turbine. This

amendment will not adversely affect the steam flow control

capability of the turbine valves. Therefore, this change will not

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

any accident previously evaluated.

(3) The proposed amendment will not involve a significant

reduction in the margin of safety.

This amendment will simply relocate the existing turbine valve

testing requirements and will not result in any changes to the

requirements. The KNPP will continue to follow the recommendations

of WCAP 11525, ``Probabilistic Evaluation of Reduction in Turbine

Valve Test Frequency.'' As a result, KNPP will continue to maintain

acceptably low probabilities of turbine valve failure. Since the

same requirements still exist and turbine valve testing will

continue to be consistent with the recommendations of WCAP 11525,

this amendment will not involve a significant decrease 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 Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: Leif J. Norrholm.

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

Generating Station, Coffey County, Kansas

Date of amendment request: December 12, 1994.

Description of amendment request: This amendment request proposes

revising Technical Specifications 4.7.1.2.1.b.1 and 4.7.1.2.1.b.2 to

clarify the surveillance requirements for verifying the correct

required position for the valves in the auxiliary feedwater system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes do not affect the ability of the auxiliary

feedwater system to perform its intended safety function. The

changes are administrative in nature since they merely clarify the

demonstration of operability required in the surveillance

requirements.

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

or different kind of accident from any previously evaluated.

There are no new failure modes or mechanisms associated with the

proposed changes. The changes are administrative changes to remove

confusion when performing surveillance requirements to demonstrate

operability.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

These proposed changes do not effect [sic] any technical

specification margin of safety. The changes only provide

clarification for performance of surveillance requirements.

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.

[[Page 3678]]

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: Theodore R. Quay

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.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: November 25, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification 5.3.1.E to allow 2645 fuel assemblies to

be stored in the fuel pool. This is an increase of 45 fuel assemblies

from the current limit of 2600. The 45 additional storage locations

currently exist in the racks in the fuel pool. They were included in

the re-racking project allowed by License Amendment No. 76 but were not

incorporated in the Technical Specifications since, at the time, it was

believed they would not be needed.

Date of publication of individual notice in Federal Register:

December 20, 1994 (59 FR 65542).

Expiration date of individual notice: January 19, 1995.

Local Public Document Room location: Ocean County Library,

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

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: December 22, 1994.

Description of amendment request: The proposed amendment is a Line

Item Technical Specifications Improvement and would revise the Cooper

Nuclear Station Technical Specifications, definition 1.0.J, concerning

entering an operational condition consistent with the wording proposed

in NRC Generic Letter 87-09, ``Sections 3.0 and 4.0 of the Standard

Technical Specifications on the Applicability of Limiting Conditions

for Operation and Surveillance Requirements,'' dated June 4, 1987.

Date of individual notice in the Federal Register: January 3, 1995

(60 FR 153).

Expiration date of individual notice: February 2, 1995.

Local Public Document Room location: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305.

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.

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

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

Date of application for amendment: August 11, 1994.

Brief description of amendment: The amendment deletes the

requirement to perform a 5-year interval hydrostatic test on the

auxiliary coolant system critical headers from TS Section 4.1.3, Table

4.1-3, Item 11.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment No.: 155.

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60379).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1994. No significant hazards

consideration comments received: No.

Local Public Document Room location: Hartsville Memorial Library,

147 West College, Hartsville, South Carolina 29550

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

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

Date of application for amendment: July 28, 1994.

Brief description of amendment: The amendment allows an increased

limit for fuel enrichment. The changes allow for the storage of fuel

with an enrichment not to exceed 4.95 + 0.05 w/o U-235 in the new and

spent fuel storage racks.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No. 156.

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45018).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

[[Page 3679]]

Local Public Document Room location: Hartsville Memorial Library,

147 West College, Hartsville, South Carolina 29550.

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: October 4, 1994.

Brief description of amendments: The amendments revised the

Technical Specification to eliminate a compliance conflict when

swapping the Centrifugal Changing (NV) pumps in Modes 4, 5, and 6. In

eliminating the conflict, this amendment permits flexibility in the

operation of the NV pumps during unit startup without a safety concern.

Date of issuance: November 17, 1994.

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment Nos.: 152 and 134.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 13, 1994 (59 FR

52003).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 17, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

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

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

Date of application of amendments: November 11, 1993, as

supplemented February 23, April 12, and July 29, 1994.

Brief description of amendments: The amendments reflect the

consolidation of the Quality Verification Department with the Nuclear

Generation Department that realigned the Nuclear Safety Review Board to

report to the Senior Nuclear Officer, change an organizational unit

term from ``group'' to ``division,'' modify titles of positions

designated to approve modifications, clarify the responsibilities of

the Safety Assurance Manager, and delete the requirement to perform an

annual fire protection audit.

Date of Issuance: January 4, 1995.

Effective date: To be implemented within 30 days from the date of

issuance.

Amendment Nos.: 208, 208, and 205.

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

619). The February 23, April 12 and July 29, 1994, letters provided

clarifying information that did not change the scope of the November

11, 1993, application or the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 4, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: February 16, 1994.

Brief description of amendment: This amendment deletes the Appendix

B Section 4.2.2 requirement to perform infrared aerial photography

every other year.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No: 65.

Facility Operating License No. NPF-73. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 6, 1994 (59 FR

34663). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: November 29, 1994, as

supplemented by letters dated December 20 and 21, 1994.

Brief description of amendment: The amendment deleted the

requirement to perform the full complement of steam generator

surveillances as outlined in the technical specifications (TSs) when

the steam generators are subjected to special inspections that are in

addition to the periodic inspections required by the TSs. This

amendment is applicable only to the special steam generator inspection

scheduled for January 1995.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment No.: 158.

Facility Operating License No. NPF-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 5, 1994 (59 FR

62416). The additional information contained in the supplemental

letters dated December 20 and 21, 1994, was clarifying in nature and

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

staff's proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

Mississippi Electric Power Association, and Mississippi Power & Light

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

Claiborne County, Mississippi

Date of application for amendment: June 17, 1994, as supplemented

by letter dated August 17, 1994.

Brief description of amendment: The amendment removed License

Condition 2.C.(25)(b) and Attachment 2 to Facility Operating License

No. NPF-29, ``Transamerica Delaval Inc. (TDI) Diesel Generator

Maintenance and Surveillance Requirements (NUREG-1216, August 1985).''

Date of issuance: January 4, 1995.

Effective date: January 4, 1995.

Amendment No: 114.

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

License.

Date of initial notice in Federal Register: September 14, 1994 (59

FR 47167).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 4, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

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

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendments: September 23, 1993 and

clarified July 25, 1994. The July 25, 1994 submission did not change

the amendment described in the initial Federal Register notice.

[[Page 3680]]

Brief description of amendments: This amendment makes changes to

Technical Specification 6.2.3, Independent Safety Engineering group.

The change maintains the requirement to perform independent technical

reviews while providing increased flexibility to accomplish this

function.

Date of Issuance: December 22, 1994.

Effective Date: December 22, 1994.

Amendment Nos.: 69.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1993 (58 FR

57851). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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: October 20, 1994.

Brief description of amendments: These amendments relocate the

diesel fuel oil testing program requirements to Technical

Specifications (TS) Section 6 and to the Bases section of the TS. Also

added were actions statements to address diesel fuel oil which does not

meet the program limits.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos. 169 and 163.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55870).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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: October 20, 1994.

Brief description of amendments: The amendments remove the schedule

for withdrawal of reactor vessel material specimens from the Technical

Specifications as discussed in Generic Letter 91-01.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos. 170 and 164.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60381).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 1994.

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. 50-424 and

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

Georgia

Date of application for amendments: September 13, 1994, as

supplemented by letter dated December 6, 1994.

Brief description of amendments: The amendments replace Containment

Systems Technical Specification (TS) 3.6.2.2 for the Spray Additive

System with a new Emergency Core Cooling System (ECCS) TS 3.5.5 for the

ECCS Recirculation Fluid pH Control System.

Date of issuance: January 5, 1995.

Effective date: Phase I to be implemented following Unit 2 Cycle 4

refueling outage; Phase II to be implemented following Unit 1 Cycle 6

refueling outage.

Amendment Nos.: 77 and 56 Phase 1; 78 and 57 Phase II.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 1994 (59 FR

53840). The December 6, 1994, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: October 19, 1991, as

supplemented March 9, April 27, and December 15, 1994.

Brief description of amendment: The amendment establishes

additional requirements for the availability of Local Power Range

Monitors (LPRMs) associated with the Average Power Range Monitoring

(APRM) system. These additional requirements further restrict the

allowable number of out of service LPRM/APRM detectors in order to

ensure a sufficient response to regional thermal hydraulic oscillations

in the reactor core to prevent violation of the Minimum Critical Power

Ratio (MCPR) safety limit. The amendment also identifies a lower bound

MCPR operating limit for each cycle as identified in the Core Operating

Limits Report. This limit shall be greater than or equal to 1.47.

Date of Issuance: December 29, 1994.

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

30 days.

Amendment No.: 176.

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

Technical Specifications.

Date of initial notice in Federal Register: November 13, 1991 (56

FR 57697). The March 9, April 27, and December 15, 1994, letters

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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: July 18, 1994.

Brief description of amendments: The amendments revised TS Table

4.3-1, Reactor Trip System Instrumentation Surveillance Requirements;

TS 3.3.4, Turbine Governor Valves; and TS [[Page 3681]] 3.7.1.2,

Turbine Driven Auxiliary Feedwater Pump, to remove one-time amendments

that are no longer necessary. In addition, six minor editorial changes

were made.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994, to be implemented within 31 days

of issuance.

Amendment Nos.: Unit 1--Amendment No. 67; Unit 2--Amendment No. 56.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45024).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 27, 1994.

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, Texas 77488.

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

County, Iowa

Date of application for amendment: June 30, 1994, as supplemented

November 10, 1994.

Brief description of amendment: The proposed amendment would add

Operability Requirements, Limiting Conditions for Operations (LCO) and

Surveillance Requirements for the Control Building Chillers.

Date of issuance: December 29, 1994.

Effective date: Date of issuance, to be implemented within 120

days.

Amendment No.: 205.

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

Technical Specifications.

Date of initial notice in Federal Register: August 3, 1994 (59 FR

39592). The additional information contained in the supplemental letter

dated November 10, 1994, was clarifying in nature and did not change

the NRC staff's initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, Iowa 52401.

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: November 15, 1993.

Brief description of amendments: The amendments make various

administrative and editorial changes to the Technical Specifications.

Date of issuance: December 30, 1994.

Effective date: December 30, 1994.

Amendment Nos.: 186 and 172.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67849).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 30, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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: July 19, 1994.

Brief description of amendments: The amendments remove the specific

requirements for Types A, B, and C containment leakage rate tests from

the Technical Specifications and replace these requirements with a

requirement to perform Types A, B, and C testing in accordance with

Appendix J to 10 CFR Part 50.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment Nos.: 187/173.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: September 28, 1994 (59

FR 49430).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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: November 15, 1993, and

supplemented October 7, 1994.

Brief description of amendments: The amendments replace the current

Technical Specification testing requirements for the Event V reactor

coolant system pressure isolation valves with the requirements from

ASME Boiler and Pressure Vessel Code, Section XI.

Date of issuance: January 5, 1995.

Effective date: January 5, 1995.

Amendment Nos.: 188/174.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: January 5, 1994 (59 FR

623).

At the request of the NRC, the licensee submitted the October 7,

1994, supplement to clarify the new requirements. This supplement did

not change the NRC's initial proposed no significant hazards

considerations finding; therefore, renoticing was not warranted.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station, Unit 1, Oswego County, New York

Date of application for amendment: August 26, 1994.

Brief description of amendment: The amendment revises Technical

Specification 4.3.3.c.(1) to permit a one-time extension of the second

10-year service period for the primary containment integrated leakage

rate (Type A) test. The one-time extension permits delaying the third

Type A test of the second 10-year service period from the 1995

refueling outage until the 1997 refueling outage. This delay will

result in an interval of approximately 46 months between the second and

third Type A tests of the second 10-year service period.

Date of issuance: December 29, 1994.

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

30 days.

Amendment No.: 151. [[Page 3682]]

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

Technical Specifications.

Date of initial notice in Federal Register: September 28, 1994 (59

FR 49431). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 29, 1994.

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: April 22, 1994.

Brief description of amendment: The amendment changes Table 3.3-9

of the Technical Specifications by modifying the indicated measurement

range for the neutron flux monitor on the remote shutdown panel. The

amendment also includes some corrections of typographical errors in the

Technical Specifications.

Date of issuance: December 20, 1994.

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

30 days.

Amendment No.: 183.

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

Technical Specifications.

Date of initial notice in Federal Register: May 25, 1994 (59 FR

27059).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 20, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Thames Valley State Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360.

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: July 22, 1994.

Brief description of amendment: The amendment (1) changes the title

of Figure 3.1-5 to be consistent with the applicable Limiting Condition

For Operation (LCO), (2) relocates the Chemical and Volume Control

System (CVCS) valve position requirements to the Reactivity Control

Systems--Shutdown Margin specifications, and (3) consolidates action

statements to be expressed in the LCOs rather than in Surveillance

Requirements. The amendment also clarifies the requirements for

calculating the heat flux hot channel factor FQ(z) when using the

base load option.

Date of issuance: December 29, 1994.

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

30 days.

Amendment No.: 99.

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

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45029).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

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 2, 1994, as supplemented

August 25, 1994.

Brief description of amendment: The amendment revises the Technical

Specifications (TS) to remove expired one-time extensions of

surveillance, removes an obsolete definition of charging pump

operability, and incorporates 11 line item improvements in accordance

with the guidance provided in Generic Letter 93-05. Several editorial

changes have been made to renumber TS pages and delete the blank pages

from the TS.

Date of issuance: January 3, 1995.

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

30 days.

Amendment No.: 100.

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

Technical Specifications.

Date of initial notice in Federal Register: September 14, 1994, (59

FR 47170).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 3, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: October 3, 1994, as

supplemented November 30, 1994.

Brief description of amendments: The amendments revise Prairie

Island Nuclear Generating Plant Technical Specification 4.6, ``Periodic

Testing of Emergency Power Systems.'' Specifically, the amendments

modify the emergency diesel generator (EDG) 24-hour load test

requirements to provide an indicated load range of 103-110 percent of

the continuous rating. These amendments also rephrase various EDG test

requirements to provide clarity and delete the requirements to verify

that the auto-connected loads do not exceed 3000 kilowatts (Unit 2 5100

kilowatts).

Date of issuance: January 5, 1995.

Effective date: January 5, 1995, with full implementation within 30

days.

Amendment Nos.: 113 and 106.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55877).

The November 30, 1994, request provided additional clarification

that was within the scope of the initial notice and did not affect the

staff's proposed no significant hazards consideration findings.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated January 5, 1995.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: October 7, 1994.

Brief description of amendment: The amendment (1) deletes the

surveillance requirements contained in Technical Specification (TS)

3.6(3)a for the raw water backup valves to the containment cooling

coils, (2) deletes the surveillance requirements in TS 3.2, Table 3-5,

item 6, for raw water valves, and (3) revises the basis of TS 2.4 to

reflect these changes. [[Page 3683]]

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 166.

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55879).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

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: April 5, 1994.

Brief description of amendments: These amendments delete the

frequency requirements for a number of audits listed under Technical

Specification 6.5.2.8 and also remove the audit requirements for the

Emergency Plan and the Security Plan since these requirements have been

added to the respective plan documents. The TS changes included in the

April 5, 1994, application were approved with the exception of those

related to the fire protection and loss prevention programs. These

proposed changes are still under evaluation by the staff and will be

addressed in a future safety evaluation.

Date of issuance: December 22, 1994.

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

within 30 days.

Amendment Nos.: 137 and 107.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 25, 1994 (59 FR

27061).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 22, 1994.

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. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

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

supplemented November 29, 1994.

Brief description of amendment: The amendment revised Technical

Specifications Section 6.0 (Administrative Controls) to reflect, in

part, licensee management changes in the corporate organization.

Specifically, the title of Executive Vice President--Nuclear Generation

was changed to Executive Vice President and Chief Nuclear Officer and a

new position, Vice President Regulatory Affairs and Special Projects,

which reports to the Executive Vice President and Chief Nuclear

Officer, was established. In addition, the list of Safety Review

Committee (SRC) members, which was previously by job title, was deleted

and replaced with a description of SRC membership requirements,

including individual qualifications and the minimum number of SRC

members was reduced from 8 to 6.

Date of issuance: December 22, 1994.

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

30 days.

Amendment No.: 220.

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

Technical Specifications.

Date of initial notice in Federal Register: September 30, 1994 (59

FR 50021).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1994.

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.

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: August 4, 1994, as supplemented

November 10, 1994.

Brief description of amendment: The amendment revises requirements

in the Technical Specifications (TSs) related to primary containment

atmosphere monitoring and drywell to torus differential pressure.

Specifically, TS 3.7.A.6. has been revised to adopt primary containment

inerting/deinerting requirements that are consistent with NUREG-1433,

``Standard Technical Specifications--General Electric Plants, BWR/4.''

TSs 4.7.A.6.a. and 4.7.A.7.a. have been revised to provide frequencies

for the verification of primary containment oxygen concentration and

pressure differential between the drywell and torus. TSs 3.7.A.7.a.(1),

3.7.A.7.a.(3), and 3.7.A.8. have been revised to provide requirements

for establishing and maintaining differential pressure between the

drywell and torus that are consistent with NUREG-1433. TS 3.7.A.9. has

been deleted and related requirements have been incorporated into Notes

for Table 3.2-8. Several administrative changes to Tables 3.2-8 and

4.2-8 have also been made to improve the overall quality of the TSs.

Date of issuance: December 28, 1994.

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

30 days.

Amendment No.: 221.

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

Technical Specifications.

Date of initial notice in Federal Register: August 31, 1994 (59 FR

45032).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 28, 1994.

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.

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: September 16, 1994, as

supplemented November 29, 1994.

Brief description of amendment: The amendment revised Technical

Specifications Section 6.0 (Administrative Controls) to reflect, in

part, licensee management changes in the corporate organization.

Specifically, the title of Executive Vice President--Nuclear Generation

was changed to Executive Vice President and Chief Nuclear Officer and a

new position, Vice President Regulatory Affairs and Special Projects,

which reports to the Executive Vice President and Chief Nuclear

Officer, was established. In addition, the list of Safety Review

Committee (SRC) members, which was previously by job title, was deleted

and replaced with a description of SRC membership requirements,

including [[Page 3684]] individual qualifications and the minimum

number of SRC members was reduced from 8 to 6.

Date of issuance: December 22, 1994.

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

30 days.

Amendment No.: 156.

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

Technical Specifications.

Date of initial notice in Federal Register: September 30, 1994 (54

FR 50021).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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: September 29, 1994.

Brief description of amendments: The amendments change the

Technical Specification surveillance interval for performing an air or

smoke flow test through each containment spray header from 5 to 10

years.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994.

Amendment Nos. 163, 144.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60385).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 27, 1994.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, New Jersey 08079.

Southern California Edison Company, et al, Docket No. 50-206, San

Onofre Nuclear Generating Station, Unit No. 1, San Diego County,

California

Date of application for amendment: April 18, 1994, as supplemented

October 26, 1994.

Brief description of amendment: The amendment revises Sections 2.C

and 2.D of the San Onofre Nuclear Generating Station, Unit 1 (SONGS 1)

Operating License. Section 2.C will be revised to modify or delete

several licensing conditions which either no longer apply or require

revision to apply to SONGS 1 in its permanently shutdown and defueled

condition. Section 2.D will be revised to exempt Fire Protection

reporting from the reporting requirements of Section 2.D.

Date of issuance: December 22, 1994.

Effective date: January 21, 1995.

Amendment No.: 156.

Facility Operating License No. DPR-13: The amendment revised the

license conditions.

Date of initial notice in Federal Register: May 25, 1994 (59 FR

27066).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 22, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713.

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

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

Alabama.

Date of amendments request: October 20, 1994.

Brief Description of amendments: The amendments delete the

requirements for the control room chlorine detection system from the TS

and the associated Bases Sections. This request is based on the fact

that all stored gaseous chlorine has been removed from the plant site

except for containers having an inventory of 150 pounds or less.

Date of issuance: December 28, 1994.

Effective date: December 28, 1994.

Amendment Nos.: 111 and 102.

Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60386).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 28, 1994.

No significant hazards consideration comments received: No.

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

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

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

Date of application for amendments: September 9, 1994 (TS 94-04).

Brief description of amendments: The amendments revise the

technical specifications related to the cold leg injection

accumulators.

Date of issuance: December 27, 1994.

Effective date: December 27, 1994.

Amendment Nos.: 192 and 184.

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

revise the technical specifications.

Date of initial notice in Federal Register: October 12, 1994 (59 FR

51629).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 27, 1994.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327, Sequoyah Nuclear Plant,

Units 1, Hamilton County, Tennessee

Date of application for amendment: November 2, 1994 (TS 94-17).

Brief description of amendment: The amendment adds Operating

License Condition 2.C.(25) to provide a limited extension of the

surveillance test intervals for certain specified instrumentation on

Unit 1 to coincide with the Cycle 7 refueling outage. The surveillance

intervals that are affected are specified in the attached safety

evaluation and are for tests that would be extended to October 1, 1995,

and would result in extension of the specified 18-, 36- and 54-month

surveillances to 29.5, 48 and 71.5 months, respectively.

Date of issuance: January 3, 1995.

Effective date: January 3, 1995.

Amendment No.: 193.

Facility Operating License Nos. DPR-77: Amendment revises the

operating license.

Date of initial notice in Federal Register: November 23, 1994 (59

FR 60387).

The Commission's related evaluation of the change to the operating

license is contained in a Safety Evaluation dated January 3, 1995.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of application for amendments: October 11, 1994.

[[Page 3685]]

Brief description of amendments: The amendments revise the

surveillance frequencies of the hydrogen analyzer channel functional

test and channel calibration.

Date of issuance: December 23, 1994.

Effective date: December 23, 1994.

Amendment Nos. 195 and 195.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55893).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 23, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

Generating Station, Coffey County, Kansas

Date of amendment request: July 22, 1994.

Brief description of amendment: This amendment revises Section 6 of

the Technical Specifications to reflect title changes in the Wolf Creek

Nuclear Operating Corporation organization.

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 81.

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

Technical Specifications.

Date of initial notice in Federal Register: October 26, 1994 (59 FR

53845).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

Generating Station, Coffey County, Kansas

Date of amendment request: February 23, 1994.

Brief description of amendment: This amendment revised Technical

Specifications 3.8.1.1, ``AC Sources Operating,'' and 3.8.1.2, ``AC

Sources Shutdown,'' to increase the minimum volume of fuel oil required

for the emergency diesel generator fuel oil day tanks. Several other

revisions are included that make editorial corrections and incorporate

requirements that were inadvertently omitted from previous amendment

requests that have been approved.

Date of issuance: December 29, 1994.

Effective date: December 29, 1994.

Amendment No.: 82.

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

Technical Specifications.

Date of initial notice in Federal Register: April 13, 1994 (59 FR

17609).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 29, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Dated at Rockville, Maryland, this 10th day of January 1995.

For the Nuclear Regulatory Commission.

Jack W. Roe,

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

Reactor Regulation

[FR Doc. 95-1026 Filed 1-17-95; 8:45 am]

BILLING CODE 7590-01-P