Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from October 31, 1994, through November 10, 1994.

The last biweekly notice was published on November 9, 1994 (59 FR

55865).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the 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 December 23, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, 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 the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: August 11, 1994

Description of amendment request: The proposed amendment deletes

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

auxiliary coolant system critical headers from Technical Specification

Section 4.1.3, Table 4.1-3, Item 11.

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 amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The proposed change will delete the

requirement to perform a hydrostatic test on the component cooling

water [CCW] system at five year intervals to ensure the integrity of

the system. However, adequate testing of the system is provided as

required by the ASME Code Section XI. This testing includes a 10-

year system hydrostatic test as well as a 40-month interval system

inservice test and provides assurance of system integrity and the

ability to perform the intended function. Therefore, there is no

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated. The proposed change will delete the requirement to

perform a hydrostatic test on the component cooling water system at

five-year intervals to ensure the integrity of the associated system

headers. Operating characteristics of the system and its physical

configuration will remain unchanged, and the system will continue to

perform its intended function. There will be an overall decrease in

the frequency of testing the CCW system due to the elimination of

redundant testing and a decrease in operational activity associated

with testing the CCW system. Since there will be no functional or

hardware changes to the system, the proposed change will not create

the possibility of a new or different type of accident.

3. The proposed amendment does not involve a significant

reduction in the margin of safety. The proposed change will delete

the requirement to perform a hydrostatic test on the component

cooling water system at five-year intervals to ensure the integrity

of the system. However, adequate testing of the system is ensured by

the required ASME Code Section XI tests. This testing includes a 10-

year system hydrostatic test as well as a 40-month interval system

inservice test and provides assurance of system integrity and the

ability to perform the intended function. Therefore, there will be

no 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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of amendment request: October 24, 1994

Description of amendment request: The proposed amendment would

remove Technical Specifications (TS) 3.3.4, Turbine Overspeed

Protection; TS 3.7.12, Area Temperature Monitoring; and TS 3.11.2.6,

Gas Storage Tanks; and their associated bases; and relocate them to

licensee-controlled documents, such as the Final Safety Analysis

Report.

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 amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes will simplify the TS, and implement the

recommendations of the Commission's Final Policy Statement on TS

Improvements. Since the elements of these TS are being relocated to

licensee-controlled documents any future changes would be controlled

under 10 CFR 50.59. The proposed changes are administrative in

nature and do not involve any modifications to any plant equipment

or affect plant operation. Therefore, there would be no increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, do not

involve any physical alterations to plant equipment, and result in

no change in the method by which any safety-related system performs

its function. Therefore, the proposed changes do not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

These changes do not affect any Final Safety Analysis Report

(FSAR) Chapter 15 accident analyses or have any impact on margin as

defined in the Bases to the Technical Specifications. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

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

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

Date of amendment request: October 27, 1994

Description of amendment request: The proposed amendments will

improve consistency throughout the Technical Specifications and their

related Bases by removing outdated material and blank pages,

incorporating minor changes in text, making editorial corrections, and

resolving other inconsistencies identified by the plant operations

staff.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10 CFR 50.92, a determination may be made that a

proposed license amendment involves no significant hazards

consideration if operation of the facility in accordance with the

proposed amendment would not: (1) involve a significant increase in

the probability or consequences of an accident previously evaluated;

or (2) create the possibility of a new or different kind of accident

from any accident previously evaluated; or (3) involve a significant

reduction in a margin of safety. Each standard is discussed as

follows:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments consist of administrative changes to the

Technical Specifications (TS) for St. Lucie Units 1 and 2. The

amendments will update the index and remove blank pages; implement

minor changes in text to rectify reference, nomenclature, spelling,

and/or consistency-in-format errors; and otherwise improve

consistency within the TS for each unit. The proposed amendments do

not involve changes to the configuration or method of operation of

plant equipment that is used to mitigate the consequences of an

accident, nor do the changes otherwise affect the initial conditions

or conservatisms assumed in any of the plant accident analyses.

Therefore, operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed administrative revisions will not change the

physical plant or the modes of plant operation defined in the

Facility License for each unit. The changes do not involve the

addition or modification of equipment nor do they alter the design

or operation of plant systems. Therefore, operation of the facility

in accordance with the proposed amendments would not create the

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

previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendments are administrative in nature and do not

change the basis for any technical specification that is related to

the establishment of, or the preservation of, a nuclear safety

margin. Therefore, operation of the facility in accordance with the

proposed amendments would not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Mohan C. Thadani, Acting

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

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

Date of amendment request: October 20, 1994.

Description of amendment request: This supersedes the licensee's

original request dated July 19, 1994, and Noticed in the Federal

Register on August 3, 1994 (59 FR 39587). The licensee proposes to

change Turkey Point Units 3 and 4 Technical Specifications and its

associated BASES, which address the maximum allowed reactor thermal

power operation with inoperable main steam safety valves (MSSVs).

Westinghouse issued Nuclear Safety Advisory Letter 94-001 which

notified the licensee of a deficiency in the basis of the Turkey Point

Technical Specification 3/4.7.1, which allows the plant to operate at

reduced power levels with a specified number of MSSVs inoperable. This

amendment request corrects the allowable power level with inoperable

MSSVs and revises the TS to conform with the guidelines of the standard

technical specifications.

The licensee also proposed changes to TS 4.7.1.1 to indicate that

the provisions of TS 4.0.4 are not applicable for entry into mode 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed maximum allowable power level values will ensure that

the secondary side steam pressure will remain below 110 percent of

the design value following a Loss of Load/Turbine Trip event, when

one or more main steam safety valves (MSSVs) are declared

inoperable. The proposed change will not impact the classification

of the Loss of Load/Turbine Trip event as a Condition II probability

event (faults of moderate frequency) per ANSI--N18.2, 1973.

Accordingly, since the proposed maximum allowable power level will

maintain the capability of the MSSVs to perform their pressure

relief function associated with a Loss of Load/Turbine Trip event,

there will be no effect on the probability or consequences of an

accident previously evaluated.

The proposed addition of ACTION statement [a] to TS 3.7.1.1,

will not [a]ffect the probability or consequences of an accident

previously evaluated, since the proposed action is consistent with

the current Technical Specifications. Reducing the Power Range

Neutron Flux High Trip Setpoint to the maximum power level will

ensure the energy transfer to the most limiting steam generator is

not greater than the available relief capacity in that steam

generator. Entry into mode 3 does not require the availability of

the MSSV, since plant conditions (i.e., not operating at reactor

power) do not create the possibility of a secondary side

overpressurization event.

In addition, the proposed change to Surveillance Requirement

4.7.1.1, will not [a]ffect the probability or consequences of an

accident previously evaluated, since the proposed plant condition is

an analyzed shutdown condition. Entry into Mode 3 for surveillance

testing does not require the availability of the MSSV, since plant

conditions (i.e., not operating at reactor power) do not create the

possibility of a secondary side overpressurization event.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not involve any change to the configuration

of any plant equipment, and no new failure modes have been defined

for any plant system or component. The proposed maximum allowable

power level will maintain the capability of the MSSVs to perform

their pressure relief function to ensure the secondary side steam

design pressure is not exceeded following a Loss of Load/Turbine

Trip event. Therefore, since the function of the MSSVs is unaffected

by the proposed changes, the possibility of a new or different kind

of accident from any accident previously evaluated is not created.

The proposed addition of ACTION statement [a] to TS 3.7.1.1,

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

accident from any accident previously evaluated, since the proposed

action is consistent with the current Technical Specifications.

Reducing the Power Range Neutron Flux High Trip Setpoint to the

maximum power level will ensure the energy transfer to the most

limiting steam generator is not greater than the available relief

capacity in that steam generator. Entry into mode 3 does not require

the availability of the MSSV, since plant conditions (i.e., not

operating at reactor power) do not create the possibility of a

secondary side overpressurization event.

In addition, the proposed change to Surveillance Requirement

4.7.1.1, will not create the possibility of a new or different kind

of accident from any accident previously evaluated, since the

proposed plant condition is an analyzed shutdown condition. Entry

into Mode 3 for surveillance testing does not require the

availability of the MSSV, since plant conditions (i.e., not

operating at reactor power) do not create the possibility of a

secondary side overpressurization event.

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

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications do not

involve a significant reduction in a margin of safety. The algorithm

methodology used to calculate the maximum allowable power level is

conservative and bounding since it is based on a number of

inoperable MSSVs per loop; i.e., if only one MSSV in one loop is out

of service, the required action to reduce power to the maximum

allowable power level would be the same as if one MSSV in each loop

were out of service. Another conservatism with the algorithm

methodology is with the assumed minimum total steam flow rate

capability of the operable MSSVs. The assumption is that if one or

more MSSVs are inoperable per loop, the inoperable MSSVs are the

largest capacity MSSVs, regardless of which capacity MSSVs are

actually inoperable. Therefore, since the maximum allowable power

level calculated for the proposed changes using the algorithm

methodology are more conservative and ensure the secondary side

steam design pressure is not exceeded following a Loss of Load/

Turbine Trip event, this proposed license amendment will not involve

a significant reduction in a margin of safety.

The proposed addition of ACTION statement [a] to TS 3.7.1.1,

will not involve a significant reduction in a margin of safety,

since the proposed action is consistent with the current Technical

Specifications. Reducing the Power Range Neutron Flux High Trip

Setpoint to the maximum power level will ensure the energy transfer

to the most limiting steam generator is not greater than the

available relief capacity in that steam generator. Entry into mode 3

does not require the availability of the MSSV, since plant

conditions (i.e., not operating at reactor power) do not create the

possibility of a secondary side overpressurization event.

In addition, the proposed change to Surveillance Requirement

4.7.1.1, will not involve a significant reduction in the margin of

safety, since the proposed plant condition is an analyzed shutdown

condition. Entry into Mode 3 for surveillance testing does not

require the availability of the MSSV, since plant conditions (i.e.,

not operating at reactor power) do not create the possibility of a

secondary side overpressurization event.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Mohan C. Thadani, (Acting)

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

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

Date of amendment request: October 20, 1994

Description of amendment request: The licensee proposes to change

Turkey Point Units 3 and 4 Technical Specifications (TS) by removing

the schedule for the withdrawal of reactor vessel material surveillance

specimens. The control of changes to this schedule, by way of a license

amendment to modify the TS, duplicates the requirements of Section

II.B.3 of Appendix H to Part 50 of Title 10 of the Code of Federal

Regulations (10 CFR). These proposed license amendments are consistent

with the guidance provided to licensees by NRC Generic Letter (GL) 91-

01, ``Removal of the Schedule for the Withdrawal of Reactor Vessel

Material Specimens from Technical Specifications.'' Additionally, these

amendments propose to correct typographical errors in the TS BASES and

to revise the reference in the TS BASES to the American Society for

Testing and Materials (ASTM) standard by which the fracture toughness

properties of the ferritic materials in the reactor vessels are

determined.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed license amendments do not involve a change in the

probability or consequences of accidents previously evaluated since

no physical changes to the plant, their operation, nor their

procedures are involved. The proposed changes are administrative in

nature and involve the activity of relocating, from the Turkey Point

Units 3 and 4 Technical Specifications (TS) to the Updated Final

Safety Analysis Report (UFSAR), the schedule for the withdrawal of

reactor vessel material surveillance specimens. The control of

changes to this schedule, by way of a license amendment to modify

the TS, duplicates the requirements of Section II.B.3 of Appendix H

to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR).

These proposed license amendments are consistent with the guidance

provided to licensees by NRC Generic Letter (GL) 91-01, ``Removal of

the Schedule for the Withdrawal of Reactor Vessel Material Specimens

from Technical Specifications.'' The TS BASES are also revised to

remove references to the table being removed from the TS. In

accordance with GL 91-01, FPL commits to maintain, the NRC-approved

version of the specimen withdrawal schedule in the Turkey Point

Units 3 and 4 UFSAR.

The current Turkey Point Units 3 and 4 TS BASES provide

background information on the use of the data obtained from material

specimens. This background information clearly defines the purpose

and relationship of this information to the requirements included in

the regulations and the ASME Code. Therefore, the removal of the

schedule for specimen withdrawal from the TS will not result in any

relaxation of the regulatory requirements of Appendix H to 10 CFR

Part 50 and do not involve an increase in the probability or

consequences of an accident previously evaluated.

The typographical corrections in the TS BASES and the revision

to the reference to ASTM E-185 are consistent with the guidance for

implementing administrative corrections to the TS to ensure that

references in the TS BASES are proper and correct.

In summary, operation of the facility in accordance with the

proposed amendment would not involve an increase in the probability

or consequences of an accident previously evaluated.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed license amendments do not create the possibility of

a new or different kind of accident from any accident previously

evaluated since no physical changes to the plant, their operation,

nor procedures are involved. The proposed changes are administrative

in nature and involve the activity of relocating, from the Turkey

Point Units 3 and 4 Technical Specifications (TS) to the UFSAR, the

schedule for the withdrawal of reactor vessel material surveillance

specimens. The control of changes to this schedule, by way of a

license amendment to modify the TS, duplicates the requirements of

Section II.B.3 of Appendix H to Part 50 of Title 10 of the Code of

Federal Regulations (10 CFR). These proposed license amendments are

consistent with the guidance provided to licensees by NRC GL 91-01,

``Removal of the Schedule for the Withdrawal of Reactor Vessel

Material Specimens from Technical Specifications.'' The TS Bases are

also revised to remove references to the table being removed from

the TS.

The removal from the TS of the schedule for the withdrawal of

reactor vessel material surveillance specimens will not result in

any loss of regulatory control because changes to this schedule are

controlled by the requirements of Appendix H to 10 CFR Part 50. In

addition, to ensure that the surveillance specimens are withdrawn at

the proper time, Surveillance Requirement 4.4.9.1.2 indicates that

the specimens shall be removed and examined to determine changes in

their material properties, as required by Appendix H. In accordance

with GL 91-01, FPL commits to maintain, the NRC-approved version of

the specimen withdrawal schedule in the Turkey Point Units 3 and 4

UFSAR.

The typographical corrections in the TS BASES and the revision

to the reference to ASTM E-185 are consistent with the guidance for

implementing administrative corrections to the TS to ensure that

references in the TS BASES are proper and correct.

The current Turkey Point Units 3 and 4 TS BASES provide

background information on the use of the data obtained from material

specimens. This background information clearly defines the purpose

and relationship of this information to the requirements included in

the regulations and the ASME Code. Therefore, the removal of the

schedule for specimen withdrawal from the TS will not result in any

relaxation of the regulatory requirements of Appendix H to 10 CFR

Part 50 and would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed license amendments do not involve physical changes

to the plant, their operation, nor their procedures. The proposed

license amendments do not create the possibility of a new or

different kind of accident from any accident previously evaluated

since no physical changes to the plant, their operation, nor their

procedures are involved. The proposed changes are administrative in

nature and involve the activity of relocating, from the Turkey Point

Units 3 and 4 Technical Specifications (TS) to the UFSAR, the

schedule for the withdrawal of reactor vessel material surveillance

specimens. The control of changes to this schedule, by way of a

license amendment to modify the TS, duplicates the requirements of

Section II.B.3 of Appendix H to Part 50 of Title 10 of the Code of

Federal Regulations (10 CFR). These proposed license amendments are

consistent with the guidance provided to licensees by NRC GL 91-01,

``Removal of the Schedule for the Withdrawal of Reactor Vessel

Material Specimens from Technical Specifications.'' The TS Bases are

also revised to remove references to the table being removed from

the TS.

The removal from the TS of the schedule for the withdrawal of

reactor vessel material surveillance specimens will not result in

any loss of regulatory control because changes to this schedule are

controlled by the requirements of Appendix H to 10 CFR Part 50. In

addition, to ensure that the surveillance specimens are withdrawn at

the proper time, Surveillance Requirement 4.4.9.1.2 indicates that

the specimens shall be removed and examined to determine changes in

their material properties, as required by Appendix H. In accordance

with GL 91-01, FPL commits to maintain the NRC-approved version of

the specimen withdrawal schedule in the Turkey Point Units 3 and 4

UFSAR.

The typographical corrections in the TS BASES and the revision

to the reference to ASTM E-185 are consistent with the guidance for

implementing administrative corrections to the TS to ensure that

references in the TS BASES are proper and correct.

The current Turkey Point Units 3 and 4 TS BASES provide

background information on the use of the data obtained from material

specimens. This background information clearly defines the purpose

and relationship of this information to the requirements included in

the regulations and the ASME Code. Therefore, the removal of the

schedule for specimen withdrawal from the TS will not result in any

relaxation of the regulatory requirements of Appendix H to 10 CFR

Part 50 and would not involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,

P.C., 1615 L Street, NW., Washington, DC 20036

NRC Project Director: Mohan C. Thadani, (Acting)

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

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

Date of amendment request: October 28, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 1.7, ``CORE ALTERATION,'' to

indicate that movement or replacement of incore instrumentation is not

considered to be a CORE ALTERATION provided that there are no fuel

assemblies in the associated core cell. TS 3/4.9.3, ``Control Rod

Position,'' and associated Bases would be revised to be consistent with

the proposed revision of TS 1.7 by changing the requirement to verify

that all control rods be inserted only during loading of fuel

assemblies into the core rather than during CORE ALTERATIONS. The

licensee has stated that these proposed changes are consistent with the

NRC's ``Improved Standard Technical Specifications,'' (NUREG-1434) and

those to be incorporated in Revision 1.The proposed amendment would

also revise Item 1.i.3) of TS Tables 3.3.2-1 and 4.3.2.1-1 to delete

the requirement showing that the Standby Liquid Control System (SLCS)

initiates Reactor Water Cleanup (RWCU) isolation in OPERATIONAL

CONDITION 5. License Amendment No. 48 issued on September 30, 1993,

deleted the requirement for SLCS to be OPERABLE in OPERATIONAL

CONDITION 5 but due to an oversight, failed to delete item 1.i.3) and

associated notations from TS Tables 3.3.2-1 and 4.3.2.1-1. The proposed

amendment would correct this oversight.

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 purpose of the definition of CORE ALTERATION is to identify

operations which have the potential for adding reactivity to the

core while the vessel head is removed and fuel is in the vessel. The

proposed definition of CORE ALTERATION explicitly states that

movement of incore instruments and undervessel replacement is not

considered to be a CORE ALTERATION. The amount of fissile material

contained in any of these instruments is insignificant and thus

would not result in any change in reactivity of the core. Similarly,

control rod movement with no fuel assemblies in the associated core

cell has negligible impact on the reactivity of the remaining core.

Removal of a control rod by either the normal control rod drive

system or uncoupling and removing the blade from the top of the

vessel with no fuel in the associated cell is not considered a CORE

ALTERATION. It has negligible impact on the reactivity of the

remaining core and is not required to be covered by Specification 3/

4.9.3. In addition, the drop of a blade on irradiated fuel is

bounded by the fuel bundle drop.

The proposed change to Specification 3/4.9.3, ``Control Rod

Position,'' making it applicable only during loading of fuel

assemblies to reflect the remaining condition that results in the

addition of positive reactivity. Specification 3/4.9.1, ``Reactor

Mode Switch,'' requires the mode switch be locked in the refuel

position. This initiates the one-rod-out interlock which prevents

the selection of more than one control rod for movement.

Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin

be greater than or equal to 0.38% delta k/k analytically determined

or 0.28% delta k/k determined by test. These specifications ensure

that the reactor will not become critical when all control rods are

not inserted. Removal of the note referencing Special Test Exemption

3.10.3 is to be consistent with the revised definition.

The proposed change to eliminate RWCU isolation requirement upon

initiation of SLCS in OPERATIONAL CONDITION 5 is consistent with

Amendment 48, which eliminated the requirement for SLCS to be

OPERABLE in OPERATIONAL CONDITION 5.

Therefore, these changes will not involve a significant increase

in the probability or consequences of an accident from any

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 proposed changes to the definition of CORE ALTERATION and

Specification 3/4.9.3, ``Control Rod Position,'' and deletion of the

RWCU isolation requirement on SLCS initiation in OPERATIONAL

CONDITION 5 do not involve a physical change in any system's

configuration. Systems required to be OPERABLE for CORE ALTERATIONS

are still required to be OPERABLE, however, no new modes of

operation are introduced based on the proposed definition.

The purpose of the definition of CORE ALTERATION is to identify

operations which have the potential for adding reactivity to the

core while the vessel head is removed and fuel is in the vessel. The

proposed definition of CORE ALTERATION explicitly states that

movement of incore instruments and undervessel replacement is not

considered to be a CORE ALTERATION. The amount of fissile material

contained in any of these instruments is insignificant and thus

would not result in any change in reactivity of the core. Similarly,

control rod movement with no fuel assemblies in the associated core

cell has negligible impact on the reactivity of the remaining core.

Removal of a control rod by either the normal control rod drive

system or uncoupling and removing the blade from the top of the

vessel with no fuel in the associated cell is not considered a CORE

ALTERATION. It has negligible impact on the reactivity of the

remaining core and is not required to be covered by Specification 3/

4.9.3. In addition, the drop of a blade on irradiated fuel is

bounded by the fuel bundle drop.

The proposed change to Specification 3/4.9.3, ``Control Rod

Position,'' making it applicable only during loading of fuel

assemblies to reflect the remaining condition which results in the

addition of positive reactivity. Specification 3/4.9.1, ``Reactor

Mode Switch,'' requires the mode switch be locked in the refuel

position. This initiates the one-rod-out interlock which prevents

the selection of more than one control rod for movement.

Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin

be greater than or equal to 0.38% delta k/k analytically determined

or 0.28% delta k/k determined by test. These specifications ensure

that the reactor will not become critical when all control rods are

not inserted. Removal of the note referencing Special Test Exemption

3.10.3 is to be consistent with the revised definition.

The proposed change to eliminate RWCU isolation requirement upon

initiation of SLCS in OPERATIONAL CONDITION 5 is consistent with

Amendment 48, which eliminated the requirement for SLCS to be

OPERABLE in OPERATIONAL CONDITION 5.

Therefore, these changes will not create the possibility of a

new or different kind of accident from any previously evaluated.

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

proposed amendment, will not involve a significant reduction in the

margin of safety.

The proposed definition of CORE ALTERATION clearly details what

constitutes a CORE ALTERATION. The definition is consistent with

NUREG-1433, ``Improved Standard Technical Specifications.'' The

definition has no impact on safety limits, setpoints, or plant

design and thus does not affect a margin of safety.

The proposed change to Specification 3/4.9.3, ``Control Rod

Position,'' making it applicable only during loading of fuel

assemblies to reflect the remaining condition that results in the

addition of positive reactivity. Specification 3/4.9.1, ``Reactor

Mode Switch,'' requires the mode switch be locked in the refuel

position. This initiates the one-rod-out interlock which prevents

the selection of more than one control rod for movement.

Specification 3/4.1.1, ``Shutdown Margin,'' requires shutdown margin

be greater than or equal to 0.38% delta k/k analytically determined

or 0.28% delta k/k determined by test. These specifications ensure

that the reactor will not become critical when all control rods are

not inserted, thus does not affect a margin of safety. The removal

of the note referencing Special Test Exemption 3.10.3 is consistent

with the revised definition.

Elimination of the requirement to initiate RWCU isolation based

upon SLCS initiation in OPERATIONAL CONDITION 5 is consistent with

deletion of the requirement to have the SLCS OPERABLE during

OPERATIONAL CONDITION 5. Therefore, there is no impact on a margin

of safety.

Therefore, based upon the above, these proposed changes will not

involve a significant reduction [in] a margin of safety.

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

this review, it appears that the three standards of 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, Acting

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

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

Connecticut

Date of amendment request: October 4, 1994

Description of amendment request: The proposed amendment relocates

the primary containment isolation valve list from Technical

Specification (TS) Section 3.7.D to the Millstone Unit 1 technical

requirements manual (TRM). This change is in accordance with the

guidance of Generic Letter (GL) 91-08. The proposed amendment also

makes administrative and editorial changes to TS Section 3.7.D and

makes changes to the associated bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with 10 CFR

50.92 and concluded that the change does not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed change does not involve a significant hazards consideration

because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed change will not result in any hardware or operating

changes. The proposed change is based upon Generic Letter 91-08 and

merely removes the containment isolation valve table and all

references to the table. The removal of the isolation valve table

from the technical specifications does not affect the operability

requirements of any of the listed valves. The technical

specifications will continue to require the isolation valves to be

OPERABLE. LCO's [limiting condition for operation] and surveillance

requirements for the valves will also remain in the technical

specifications. The containment isolation valve table will be

relocated to the Millstone Unit No. 1 TRM which is controlled in

accordance with 10 CFR 50.59.

This change is administrative in nature and does not involve an

increase in the probability or consequence of an accident previously

evaluated. Further, the proposed change does not alter the design,

function, or operation of the valves involved, and therefore does

not affect the probability or consequence of any previously

evaluated accident.

The clarification of Surveillance Requirement 4.7.D.2 ensures

that the flow path affected by an inoperable primary containment

isolation valve is isolated and maintained in the isolated

condition. This change ensures that probability or consequence of a

previously analyzed accident is not increased.

The nonintent changes involved with this license amendment

request are administrative in nature and will not, in and of

themselves, increase the probability or consequences of any

transient or accident previously analyzed. This does not affect or

have any potential impact upon any of the design basis types of

accidents previously analyzed. There are no failure modes affected

by the changes. As such, there are no design basis accidents

affected by the changes.

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

from any previously analyzed.

The proposed change will not impose any different operational or

surveillance requirements, nor will the change remove any such

requirements. The change proposes to relocate the containment

isolation valve list from the technical specifications to the TRM.

Adequate control of information is maintained. Further, as stated

above, the proposed change does not alter the design, function, or

operation of the valves involved, and therefore no new accident

scenarios are created.

The clarification of Surveillance Requirement 4.7.D.2 ensures

that the flow path affected by an inoperable primary containment

isolation valve is isolated and maintained in the isolated

condition. Since this change only ensures that the position of a

valve in the isolated condition is recorded, this change cannot

create a new or different kind of accident.

The nonintent changes do not, by their nature, modify plant

response during operation or during any transient or accident.

Therefore, there are no failure modes that can represent a new

unanalyzed accident.

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

The proposed change will not reduce the margin of safety since

it has no impact on any safety analysis assumption. The proposed

change does not decrease the scope of equipment currently required

to be operable or subject to surveillance testing, nor does the

proposed change affect any instrument setpoints or equipment safety

functions.

The relocation of the valve list is consistent with the guidance

provided in GL 91-08. The intent of the technical specification will

be met since the change will not alter function or operability

requirements for any primary containment isolation valve.

The clarification of Surveillance Requirement 4.7.D.2 ensures

that the flow path affected by an inoperable primary containment

isolation valve is isolated and maintained in the isolated

condition. Therefore, this change ensures that the margin of safety

established by the safety analyses is maintained.

The nonintent changes involved with this license amendment

request are administrative in nature and will not, in and of

themselves, reduce any margin of safety. There is no impact on the

performance of any safety system. There is no increase in the

consequences of any accident and, as such, there is no 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: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee

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

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

Connecticut

Date of amendment request: September 28, 1994

Description of amendment request: The proposed change would revise

the Surveillance Requirement 4.6.1.2.a of the Technical Specifications

to permit a more flexible schedule for containment leakage Type A

testing. The information in the associated Bases Section would also be

changed. In conjunction with this amendment request, the licensee has

requested a partial and schedular exemption, dated September 28, 1994,

from the requirements of Section III.D.1.(a) of Appendix J to Title 10

of the Code of Federal Regulations, Part 50.

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 basis for this conclusion is that the three criteria of

10 CFR 50.92(c) are not compromised. The proposed change does not

involve a SHC [significant hazards consideration] because the change

would not:

Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Type A tests are performed to ensure that the total leakage from

containment does not exceed the maximum allowable primary

containment leakage rate at a calculated peak containment internal

pressure permitted by the Millstone Unit No. 3 Technical

Specifications and FSAR [Final Safety Analysis Report]. This assures

compliance with the dose limits of 10CFR100.

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 3 Technical Specifications will increase the

flexibility for scheduling the Type A tests. They do not modify the

maximum allowable leakage rate at the calculated peak containment

pressure, do not impact the design basis of the containment, and do

not change the post-accident containment response.

The first two Type A tests of the first 10-year service period

for Millstone Unit No. 3 have been conducted. The results of these

tests demonstrate that Millstone Unit No. 3 has maintained control

of containment integrity by maintaining margin between the

acceptance criterion and the ``As-Found'' and ``As-Left'' leakage

rates.

Historically, Type A tests have a relatively low failure rate,

where Type B and C testing (local leakage rate tests) could not

detect the leakage path. Most Type A test failures are attributed to

failures of Type B or C components (containment penetrations and

isolation valves). Type B and C components are tested per

Surveillance Requirement 4.6.1.2.d of the Millstone Unit No. 3

Technical Specifications. These tests are required to be conducted

at intervals no greater than 24 months, and the acceptance criterion

for the combined leakage rate for all penetrations and valves

subject to the Type B and C tests is 0.6 La. These local

leakage rate tests provide assurance that containment integrity is

maintained. The relatively low ``As-Left'' Type B and C total

leakage resulting from each successive outage indicates that the

leakage has been maintained within the technical specification

acceptance criterion, and demonstrates that improvements are

continually being made to the Type B and C program. The Type B and C

leakage results have decreased over the last three refueling

outages. This proposal does not request any changes to the

requirements for Type B and C testing. The Type B and C tests will

continue to be performed in accordance with the requirements of

Surveillance Requirement 4.6.1.2.d. These tests confirm that the

leak-tightness of the containment isolation valves and penetrations

has been maintained.

Based on the previous Type A, B, and C tests, the Millstone Unit

No. 3 containment's structural integrity is considered to be in

sound condition. No operations are known to have occurred which

would suggest any substantial degradation of these results.

Additionally, no structural modifications are planned for the next

refueling outage.

Based on the above, the proposed change to Surveillance

Requirement 4.6.1.2.a of the Millstone Unit No. 3 Technical

Specifications does not involve a significant increase in the

probability or consequences of an accident previously analyzed.

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

from any previously analyzed.

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 3 Technical Specifications will increase the

flexibility in scheduling the Type A tests. They do not make any

physical or operational changes to existing plant structures,

systems, or components. In addition, the proposed change does not

modify the acceptance criteria for the Type A tests. Maintaining the

leakage through the containment boundary to the atmosphere within a

specific value ensures that the plant complies with the requirements

of 10 CFR 100. The containment boundary serves as an accident

mitigator; it is not an accident initiator. Therefore, the proposed

change to Surveillance Requirement 4.6.1.2.a does not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

The proposed change to Surveillance Requirement 4.6.1.2.a of the

Millstone Unit No. 3 Technical Specifications will increase the

flexibility for scheduling the Type A tests. They do not modify the

maximum allowable leakage rate at the calculated peak containment

pressure, do not impact the design basis of the containment, and do

not change the post-accident containment response.

Based on the previous Type A, B, and C tests, the Millstone Unit

No. 3 containment's structural integrity is considered to be in

sound condition. No operations are known to have occurred which

would suggest any substantial degradation of these results.

Additionally, no structural modifications are planned for the next

refueling outage.

Based on the above, the proposed 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: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee

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

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

County, New Jersey

Date of amendment request: September 29, 1994

Description of amendment request: The proposed change would remove

the sections from the Techical Specifications that are entitled

``Seismic Instrumentation'' and ``Meteorological Instrumentation'' and

relocate the information and testing requirements to the Salem Updated

Final Safety Analysis Report. The proposed change conforms with the NRC

guidance presented in the ``Final Policy Statement on Technical

Specifications Improvements for Nuclear Power Reactors'' published in

the Federal Register (58 FR 39132).

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. Neither the relocation of the seismic/

meteorological specifications to the Salem UFSAR nor the elimination

of the Special Report requirements represent changes that affect

plant safety or 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

operability and surveillance of instrumentation that are not safety

related and will not impact the operation of any plant safety

related component or equipment. Therefore, these changes will not

create a new or unevaluated accident or operating condition.

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

safety.

In accordance with guidance provided by the NRC regarding the

improvement of Technical Specifications (58 FR 39132), the proposed

changes relocate the seismic and meteorological instrumentation

portion of the Technical Specification, with the exception of the

Special Report requirements, to the Salem UFSAR. These instruments

are not safety related and do not have any associated safety margins

which could be affected by this change.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

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

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

County, New Jersey

Date of amendment request: September 29, 1994

Description of amendment request: The proposed change to the

Technical Specifications revises the surveillance interval for

performing an air or smoke flow test through each containment spray

header from once every five years to once every ten years. The proposed

change implements a recommended line-item improvement from Generic

Letter 93-05, ``Line-Item Technical Specifications Improvements to

Reduce Surveillance Requirements for Testing During Power Operation.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not affect the assumptions, design

parameters or results of UFSAR accidents analyzed. The proposed

change does not involve a hardware change, a change to the operation

of any system or component, or a change to an existing structure.

The proposed change leads to a reduction in radiation exposure to

plant personnel and the reduction of an unnecessary burden on plant

staff. The Containment Spray System header and nozzles are

fabricated from corrosion resistant stainless steel and are

maintained dry. Operating experience demonstrates that the proposed

increase in the Containment Spray surveillance test interval would

not affect operability of the system. Testing the Containment Spray

System header and nozzles at the proposed increased surveillance

interval does not increase the probability or consequences of an

accident previously evaluated.

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

accident from any previously evaluated.

The proposed change does not modify equipment, affect the system

design basis or operability. This change does not alter parameters

utilized in the analyzed accident scenarios. The Containment Spray

System piping and nozzles are fabricated from corrosion resistant

stainless steel. The proposed change in surveillance frequency is

consistent with the guidance provided in GL 93-05. Testing the

Containment Spray System header and nozzles at the proposed

increased surveillance interval does not create the possibility of a

new of different kind of accident from those previously evaluated.

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

safety.

The proposed change only involves a decrease in the surveillance

frequency and does not alter the performance of the surveillance

itself. System equipment and operation remains unchanged.

Operability and reliability is still maintained by periodic testing.

Testing the Containment Spray System header and nozzles at the

proposed increased surveillance interval does not involve a

significant reduction in the margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

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

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

County, New Jersey

Date of amendment request: October 11, 1994

Description of amendment request: The proposed amendment would make

two Technical Specification changes concerning the pressurizer heaters.

The first change would add the phrase ``capable of being powered from

an emergency power supply'' to the Limiting Condition of Operation

(LCO) 3/4.4.4. The second change would alter the frequency of

surveillance requirement 4.4.4.2 from 92 days to every refueling

outage.

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:

0. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The request (both proposed changes) does not change any

assumption or parameter assumed to function in any of the design/

licensing basis analysis.

The proposed change as described in section IA merely relocates

the requirement to supply emergency power to the required heater

group from the action to the LCO statement.

The change as described in section IB does not eliminate the

surveillance requirement, but extends its frequency from 92 days to

once per refueling outage in accordance with NRC recommendation. The

design of the Salem Station Pressurizer heaters is identical to that

described in the NUREG 1366 (Improvements to Technical

Specifications Surveillance Requirements, published December 1992),

and Generic Letter 93-05 (Line-Item Technical Specifications

improvements to Reduce Surveillance Requirements for Testing During

Power Operation, issued on September 27, 1993), and the extension of

the surveillance requirement is a recognized enhancement and

assurance to the continued reliability of the pressurizer heaters.

Based upon the above, PSE&G concludes that the proposed changes

do not involve a significant increase in the probability or

consequences of an accident previously evaluated.2. Does not create

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

accident previously evaluated.

The proposed changes do not introduce any design or physical

configuration changes to the facility which could create new

accident scenarios.

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

safety.

As stated in response to question number 1 above, the request

does not change any assumption or parameter assumed to function in

any of the design/licensing basis analysis. One change merely

relocates a requirement from one section of the LCO to another, and

the second change incorporates the recommendations and enhancements

as stated in NUREG 1366 and GL 93-05.

Consequently, PSE&G concludes that the change does not involve a

significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

Saxton Nuclear Experimental Corporation, Docket No. 50-146, Saxton

Nuclear Facility, Bedford County, Pennsylvania

Date of amendment request: August 1, 1994. This supersedes the

request dated June 23, 1993.

Description of amendment request: The proposed amendment would

revise the technical specifications to allow characterization

activities related to the decommissioning of the Saxton Nuclear

Facility and add administrative activities associated with the

characterization activities.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

considerations because the changes would not:

1. Involve a significant increase in the probability or consequences of

an accident previously evaluated.

The activities associated with characterization of the facility

will have a minimum impact on the physical condition of the

containment vessel as it relates to the risk of fire and has no

effect on the risk of flooding.

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

from any previously analyzed.

In its present condition, the only accidents applicable to the

site are fire, flood, and radiological hazard. The possibility of a

new or different type of accident than that previously evaluated in

the FSAR will not be created by the implementation of activities

permitted by the approval of this amendment request.

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

No margins of safety relevant to the equipment at the facility

exist. Activities involved in characterization will not involve a

reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee 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: Saxton Community Library, 911

Church Street, Saxton, Pennsylvania 16678Attorney for the Licensee:

Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, D.C. 20037

NRC Project Director: Seymour H. Weiss

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

Description of amendments request: The proposed Technical

Specification changes will delete requirements for the chlorine

detections systems from Technical Specification 3/4.3.3.6 and its

associated bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Removal of the control room chlorine detection system does

not involve a significant increase in the probability or

consequences of an accident previously evaluated because on-site

gaseous chlorine will be limited to a maximum per container

inventory of 150 pounds located greater than 100 meters from the

control room, and manual isolation of the control room is provided.

This is in compliance with Regulatory Guide 1.95. Furthermore,

offsite chlorine storage and transportation meets the requirements

of Regulatory Guides 1.78 and 1.95. Therefore, the probability of

occurrence of an accident is not affected.

There are no radiological consequences associated with chlorine

release accidents. Therefore, the consequences of an accident

previously evaluated are not increased.

2. Removal of the control room chlorine detection system does

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

from any accident previously evaluated since the chlorine detectors

are utilized for detection of accidental chlorine release and are

not accident initiators. Gaseous chlorine has been removed from the

plant site, except for a permissible maximum per container inventory

of 150 pounds which will be located greater than 100 meters away

from the control room. In addition, there is a provision for the

manual isolation of the control room. Therefore, on-site chlorine

storage meets the requirements of Regulatory Guide 1.95.

Furthermore, offsite chlorine storage and transportation meet the

requirements of Regulatory Guides 1.78 and 1.95.

3. Removal of the control room chlorine detection system does

not involve a significant reduction in the margin of safety related

to the protection of control room operators from excessive levels of

chlorine since the onsite chlorine storage will be limited to a

maximum per container inventory of 150 pounds at the chlorination

house, which is located greater than 100 meters from the control

room. In addition, manual isolation of the control room is also

provided. This meets the requirements of Regulatory Guide 1.95.

Therefore, onsite and offsite chlorine storage and transportation

meets the requirements of Regulatory Guides 1.78 and 1.95.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: William H. Bateman

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: October 7, 1994 (TS 351)

Description of amendment request: The proposed amendment clarifies

the BFN diesel generator surveillance requirements which were thought

to be too ambiguous by both the NRC staff and TVA personnel. In

addition, the applicable Bases sections are being reviewed to provide

additional background information. TVA is revising Units 1 and 2 TS

Surveillance Requirements 4.9.B.3 and Unit 3 TS Surveillance

Requirement 4.9.B.2 to more closely reflect the requirements of

Improved Standard Technical Specifications (ISTS) for BWR/4s (NUREG-

1433), Section 3.8.1, AC Sources--Operating, Condition B for plant

operation with an inoperable diesel generator.

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 amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change revises the surveillance requirements for

plant operation with an inoperable diesel generator. Diesel

generator operation is not a precursor to any design basis accident

or transient analyzed in the Browns Ferry Updated Final Safety

Analysis Report. Therefore, this change does not increase the

probability of any previously evaluated accident.

The proposed change will eliminate the requirement for

unnecessary diesel generator starts and the incumbent diesel

generator wear when a diesel generator is made inoperable for

planned maintenance and testing. Thus, the proposed change will

result in an increase in the reliability and availability of the

diesel generators. Therefore, this change does not increase the

consequences of any previously evaluated accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed change to the surveillance requirements for plant

operation with an inoperable diesel generator does not involve a

modification to plant equipment. No new failure modes are

introduced. There is no effect on the function of any plant system

and no new system interactions are introduced by this change.

Therefore, the proposed amendment does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change will eliminate the requirement for

unnecessary diesel generator starts and the incumbent diesel

generator wear. Thus, the proposed change will result in an increase

in the reliability and availability of the diesel generators. Since

the ability of the diesel generators to perform their safety

function will not be degraded, the proposed amendment does not

involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: November 2, 1994 (TS 94-17)

Description of amendment request: The proposed change would add

Operating License Condition 2.C.(25) to provide temporary extension of

the intervals for the surveillance tests specified in the submittal on

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

be extended to October 1, 1995, which would result in extension of the

specified 18-month, 36-month and 54-month surveillances to 29.5, 48 and

71.5 months, respectively.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change is temporary and allows a one-time extension

of specific surveillance requirements (SRs) for Cycle 7 to allow

surveillance testing to coincide with the seventh refueling outage.

The proposed surveillance interval extension will not cause a

significant reduction in system reliability nor affect the ability

of the systems to perform their design function. Current monitoring

of plant conditions and continuation of the surveillance testing

required during normal plant operation will continue to be performed

to ensure conformance with TS operability requirements. Therefore,

this change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any previously analyzed.

Extending the surveillance interval for the performance of

specific testing will not create the posssibility of any new or

diffferent kind of accidents. No changes are required to any system

configurations, plant equipment, or analyses. Therefore, this change

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

accident from any accident previously evaluated.

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

Surveillance interval extension will not impact any plant safety

analyses since the assumptions used will remain unchanged. The

safety limits assumed in the accident analyses and the design

function of the equipment required to mitigate the consequences of

any postulated accidents will not be changed since only the

surveillance test interval is being extended. Historical performance

generally indicates a high degree of reliability, and surveillance

testing perforned during normal plant operation will continue to be

performed to verify proper performance. Therefore, the plant will be

maintained within the analyzed limits, and the proposed extension

will not significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: October 25, 1994

Description of amendment request: The proposed change would extend

the functional surveillance frequency for the hydrogen recombiners from

once per 6 months to once per 18 months. The proposed changes would

also delete the surveillance requirement to operate the containment

purge blower. Also, minor editorial changes would be made to improve

the clarity and consistency between the NA-&2 Technical Specifications

(TS).

The NRC has completed a comprehensive examination of surveillance

requirements in the TS that require testing at power. The evaluation is

documented in NUREG-1366, ``Improvements to Technical Specification

Surveillance Requirements,'' dated December 1992. The NRC staff found,

that while the majority of testing at power is important, safety can be

improved, equipment degradation decreased, and an unnecessary burden on

personnel resources eliminated by reducing the amount of testing at

power that is required by the TS. Based on the results of the

evaluations documented in NUREG-1366, the NRC issued Generic Letter

(GL) 93-05, ``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation,'' dated

September 27, 1993.

The Hydrogen Recombiner System (HRS) removes the hydrogen gasses

that accumulate in the containment atmosphere following a design-basis

loss-of-coolant accident. Using the guidelines provided by GL 93-05,

Item 8.5 and NUREG-1366, the licensee is requesting a change to the

functional surveillance testing frequency for the hydrogen recombiners

from once per 6 months to once per 18 months. These changes in the

surveillance requirements do not affect plant or HRS operations. In

addition, several other changes are being requested for clarity and

consistency between NA-1&2 TS.

TS Surveillance Requirement 4.6.4.2.a states in part that ''...

each purge blower operates for 15 minutes.'' NA-1&2 are equipped with

two different types of ``purge blowers.'' One type of purge blowers is

an integral part of the HRS. These hydrogen recombiner purge blowers

are capable of exhausting containment gasses directly to the atmosphere

even with the recombiner incapable of removing hydrogen gas. The second

type of purge blowers is the containment purge blowers which exhaust

directly from the containment to atmosphere and are not associated with

the hydrogen recombiners. Surveillance Requirement 4.6.4.2.a will be

modified to state that the purge blowers being referred to in this

surveillance requirement are the hydrogen recombiner purge blowers.

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:

Specifically, operation of North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Testing of the Hydrogen Recombiner System once per 18 months

will continue to assure that the Hydrogen Recombiner System will be

capable of performing its intended functions. The containment purge

blowers are not part of the Hydrogen Recombiner System and are not

assumed to function during accident conditions. Therefore, these

changes to the Hydrogen Recombiner System Technical Specifications

do not affect the probability or consequences of any previously

analyzed accident.

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

accident from any accident previously evaluated.

The proposed Technical Specification changes do not involve any

physical modification of the plant or result in a change in a method

of operation. Testing the Hydrogen Recombiner System once per 18

months will continue to assure that the Hydrogen Recombiner System

will be capable of performing its intended function. Therefore, a

new or different type of accident is not made possible.

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

The proposed Technical Specification changes do not affect any

safety limits or limiting safety system settings. System operating

parameters are unaffected. The availability of equipment required to

mitigate or assess the consequence of an accident is not reduced.

The containment purge blowers are not part of the Hydrogen

Recombiner System and are not assumed to function during accident

conditions. Testing of the Hydrogen Recombiner System once per 18

months will continue to assure that the Hydrogen Recombiner System

will be capable of performing its intended functions. Safety margins

are, therefore, not decreased.

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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Mohan C. Thadani, Acting

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: September 2, 1992

Description of amendment request: The proposed amendment would

revise the technical specifications to give the correct value for the

sodium pentaborate tank low-level alarm.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The change corrects the Technical Specifications to reflect the

correct and more conservative operating capability of the design. In

this instance there is no increase in the probability or

consequences of an accident previously evaluated because no changes

in concentration limits or volume are proposed by this change. The

Technical Specifications are being changed to recognize the more

prudent operating mode of the SLC [standby liquid control] storage

tank in that margin is available, and has always been available,

after a low level alarm. The margin allows corrective action to be

taken prior to exceeding Technical Specification limits. In summary,

a more prudent mode of operating is recognized by this change and

the design requirements of volume and concentration are not changed.

Hence, the accident analyses remains [sic] unaffected by this

change.

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

kind of accident from any accident previously evaluated?

The SLC function and reliability are not affected by this

change. No new modes of plant operation are introduced with this

change. Hence, no new or different kind of accident is credible.

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

of safety?

No change to the required volume and concentrations are being

proposed by this [modification]. Neither the original design or

accident analysis is affected by this change. A more prudent mode of

operation, that currently exists, is recognized by this proposal.

Therefore, there is no impact to a margin of safety with this

change.

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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,

1400 L Street, NW., Washington, D.C. 20005-3502

NRC Project Director: Theodore R. Quay Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey

County, Kansas

Date of amendment request: October 21, 1994 and supplement dated

October 27, 1994

Description of amendment request: This amendment request revises

Technical Specification Surveillance Requirements 4.7.1.2.1.c.2

(operability testing for the turbine-driven auxiliary feedwater (AFW)

pump automatic start feature) and 4.3.2.2 (engineered safety feature

actuation system instrumentation response time testing for the turbine-

driven AFW pump) to correct an inconsistency caused by system

limitations to supply steam to the turbine-driven AFW pump prior to

entry into Mode 3. These specifications are being revised to indicate

that the provisions of Technical Specification 4.0.4 are not applicable

for entry into Mode 3.

In addition, Technical Specification Surveillance Requirement

4.7.1.2.1.c is being revised to delete the requirement to be performed

during shutdown.

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.

WCNOC [Wolf Creek Nuclear Operating Corporation] is proposing to

modify Surveillance Requirements 4.3.2.2 and 4.7.1.2.1.c.2 by adding

an exemption for [from] the provisions of Technical Specification

4.0.4 and deleting the shutdown requirement. Entry into Mode 3 would

allow for appropriate test conditions (e.g., adequate steam pressure

available) to complete the operability testing of the turbine-driven

AFW pump. The acceptance criteria such as response time, or test

frequency, are not revised. Therefore, the surveillance will

continue to verify the operability of the turbine-driven AFW pump.

Additionally, the proposed changes are consistent with the new

improved Standard Technical Specifications for Westinghouse plants

(NUREG-1431)

Considering the above, the proposed changes to Surveillance

Requirements 4.3.2.2 and 4.7.1.2.1.c.2, of the WCGS [Wolf Creek

Generating Station] Technical Specifications, do not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

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

or different kind of accident from any previously evaluated.

The proposed changes do not make any physical or operational

changes to existing plant structures, systems, or components. The

proposed changes do not introduce any new failure modes. They simply

allow tests to be performed at appropriate conditions rather than

during shutdown.

Additionally, the proposed changes do not modify the acceptance

criteria for the tests. The purpose of the tests is to ensure that

the turbine-driven AFW pump can perform its intended function.

Thus, the proposed changes do not create the possibility of a

new or different kind of accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed changes do not have any adverse impact on the

Updated Safety Analysis Report accident analyses. The applicable

acceptance criteria for the turbine-driven AFW pump will not be

modified by these proposed changes. The proposed changes will permit

the tests to be conducted under the proper conditions, so that the

ability of the turbine-driven AFW pump to perform its intended

safety function can be confirmed.

Based on the above discussions it has been determined that the

requested technical specification revision does not involve a

significant increase in the probability or consequences of an

accident or other adverse condition; or 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 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, NW., Washington, DC 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.

Philadelphia Electric Company, Public Service Electric and Gas

Company, Delmarva Power and Light Company, and Atlantic City

Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of amendment request: October 25, 1994

Brief description of amendment request: The proposed amendment

would clarify the minimum reactor steam pressure required for

Surveillance Requirement (SR) 4.5.C.1(e). The revised SR will require

the licensee to verify that the High Pressure Coolant Injection Pump,

with reactor pressure less than or equal to 175 psig, develop a flow

rate of greater than or equal to 5000 gpm against a system head

corresponding to reactor pressure. The current SR specifies that the

test be performed at 150 psig but does not provide a range of

acceptable pressures.

Date of publication of individual notice in Federal Register: Nov.

7, 1994 (59 FR 55498)

Expiration date of individual notice: December 7, 1994

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.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: June 17, 1994, supplemented by

letter dated September 21, 1994,

Brief description of amendments: The amendments allow removal of

five tables of component lists from the Palo Verde Technical

Specifications (TS) in accordance with NRC Generic Letter (GL) 91-08,

``Removal of Component Lists from Technical Specifications.'' The

affected tables are Table 3.3-9B, Table 3.3-9C, Table 3.6-1, Table 3.8-

2, and Table 3.8-3. These five removed tables will be incorporated into

a new document, which will be administratively controlled according to

the change control provisions of the TS.

Date of issuance: October 31, 1994

Effective date: October 31, 1994, to be implemented no later than

45 days from the date of issuance.

Amendment Nos.: 85, 73, and 57

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 20, 1994 (59 FR

37061) The supplemental letter provided certain revised TS pages for

clarification purposes and did not change the original no significant

hazards determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated October 31, 1994.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: January 4, 1994

Brief description of amendments: These amendments revise Technical

Specification 3.2.3, ``Azimuthal Power Tilt,'' to change the azimuthal

power tilt limit from less than or equal to 10 percent to less than or

equal to 3 percent when the core operating limit supervisory system is

out of service. The associated TS Bases are similarly changed.

Date of issuance: November 3, 1994

Effective date: November 3, 1994, to be fully implemented no later

than 45 days from the date of issuance

Amendment Nos.: 86, 74, and 58

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

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

22001) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 3, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert

Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: May 27, 1993

Brief description of amendment: The amendment revises the heatup

and cooldown curves and the low-temperature overpressure protection

(LTOP) controls. The changes to the LTOP controls support proposed

modifications to allow a variable-setpoint (VLTOP) protection system.

The VLTOP system will increase the allowable operating pressure band in

the LTOP region and increase the flexibility in the use of the reactor

coolant pumps.

Date of issuance: November 1, 1994

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

30 days.

Amendment No.: 178

Facility Operating License No. DPR-69: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (59

FR 37064) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 1, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Date of application for amendment: November 4, 1993, as

supplemented April 27, 1994, and October 10, 1994.

Brief description of amendment: The proposed amendment revises

Technical Specification 6.13.1 to provide use of alarming dosimeters in

high radiation areas. This change includes newly revised 10 CFR Part 20

requirement references and is consistent with NUREG-1413, Standard

Technical Specifications--Westinghouse Plants, Specification 5.11.1.

Date of issuance: November 4, 1994

Effective date: November 4, 1994

Amendment No.: 152

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

Technical Specifications.

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

4935) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 4, 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, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of application for amendment: September 28, 1993, as amended

April 5, 1994.

Brief description of amendment: The amendment revises Technical

Specification 3/4.8.1, ``AC Sources--Operating'', and associated Bases

to be consistent with the new ``Standard Technical Specifications for

Westinghouse Plants'', NUREG-1431, Revision 0.

Date of issuance: November 4, 1994

Effective date: November 4, 1994

Amendment No. 51

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

Technical Specifications.

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

57845) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 4, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Iowa-Illinois Gas and Electric

Company, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power

Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-254

and 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock

Island County, Illinois; Docket Nos. 50-295 and 50-304, Zion

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

Date of application for amendments: July 8, 1994

Brief description of amendments: The amendment revises the

operating licenses by adding a license condition that would allow the

commitments made in response to NUREG-0737, ``Clarification of TMI

Action Plan Requirements,'' to be controlled pursuant to the

requirements of 10 CFR 50.59.

Date of issuance: November 3, 1994

Effective date: November 3, 1994

Amendment Nos.: for Dresden, Amendment Nos. 129 and 123; for Quad

Cities, Amendment Nos. 150 and 146; and for Zion, Amendment Nos. 158

and 146.

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

DPR-39, and DPR-48. The amendments revised the operating licenses.

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

45021) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 3, 1994. No significant

hazards consideration comments received: No

Local Public Document Room locations: for Dresden, the Morris

Public Library, 604 Liberty Street, Morris, Illinois 60450; for Quad

Cities, the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois

61021; and for Zion, the Waukegan Public Library, 128 N. County Street,

Waukegan, Illinois 60085.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Power Plant,

Unit 1, Monroe County, Michigan

Date of application for amendment: December 9, 1993 (Reference NRC-

93-0143).

Brief description of amendment: This amendment modified the

Technical Specifications (TS) incorporated in Possession-Only License

No. DPR-9 as Appendix A by modifying the Protected Area definition and

Waste Disposal Surveillances to provide the appropriate 10 CFR Part 20

references in conformance with a revision of 10 CFR Part 20 (56 FR

23360).

Date of issuance: November 3, 1994.

Effective date: This license amendment is effective as of the date

of its issuance and must be fully implemented no later than 30 days

from the date of issuance.

Amendment No.: 10.Possession-Only License No. DPR-9: The amendment

revised the TS.

Date of initial notice in Federal Register: July 20, 1994 (59 FR

37070) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 3, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

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

Date of application for amendments: November 21, 1991.

Brief description of amendments: The amendments were submitted as a

result of NRC recommendations pertaining to Generic Letter 90-06 for

the power-operated relief valves and block valves and low-temperature

overpressure protection systems.

Date of issuance: October 27, 1994

Effective date: October 27, 1994

Amendment Nos.: 150 and 132

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 10, 1993 (58

FR 59748) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated Ocotber 27, 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: December 8, 1993, as

supplemented April 20, September 8, 1994, and October 25, 1994.

Brief description of amendments: The amendments revise Technical

Specification 3.4 to address the need to bypass automatic initiation of

the Emergency Feedwater system with the main feedwater pump discharge

pressure is below actuation setpoint during startup and shutdown in

order to prevent inadvertent actuation. The amendments also deleted

operability requirements for the Emergency Condenser Cooling Water

(ECCW) system.

Date of Issuance: October 31, 1994

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

issuance.

Amendment Nos.: 207, 207, and 204

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

amendments revised the Technical Specifications.

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

39584) The April 20, September 8, and October 25, 1994 supplements

provided additional information that did not change the scope of the

December 8, 1994, application and the initial proposed no significant

hazards consideration determination.The Commission's related evaluation

of the amendments is contained in a Safety Evaluation dated October 31,

1994. No significant hazards consideration comments received: No

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

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

Date of amendment request: February 5, 1993, as supplemented by

letter dated August 1, 1994.

Brief description of amendment: The amendment revised the Technical

Specifications to incorporate a technical review and control process to

supplement the onsite technical review and approval of new procedures

and changes thereto affecting nuclear safety.

Date of issuance: November 4, 1994Effective date: November 4, 1994

Amendment No.: 100

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

Technical Specifications.

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

45022) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 4, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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 amendment: May 23, 1994

Brief description of amendment: This amendment revises Technical

Specifications Section 3/4.7.1.1, Turbine Cycle, Safety Valves, to

delete a specific reference to the 1994 edition of the ASME Code and

refer to testing in accordance with Technical Specification 4.0.5, the

In-Service Inspection and In-Service Testing Specification.

Date of Issuance: November 1, 1994

Effective Date: November 1, 1994

Amendment No.: 68

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

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

34664) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 1, 1994No significant

hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

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

Linn County, Iowa

Date of application for amendment: June 30, 1994

Brief description of amendment: The proposed amendment would

clarify the requirement for the audit of conformance to Technical

Specifications, delete the requirement for Safety Committee oversight

of the Emergency Plan and Security Plan and allow designation by the

Plant Superintendent signature authority for procedure approval.

Date of issuance: November 2, 1994

Effective date: Date of issuance and to be implemented within 60

days

Amendment No.: 202

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

Technical Specifications.

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

39591) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 2, 1994No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

Illinois

Date of application for amendment: August 12, 1994

Brief description of amendment: The amendment modifies Clinton

Power Station Technical Specification 3/4.6.2.2, ``Drywell Bypass

Leakage,'' to allow drywell bypass leakage rate tests to be performed

at intervals as long as five years based on the demonstrated

performance of the drywell structure.

Date of issuance: November 3, 1994

Effective date: November 3, 1994

Amendment No.: 94

Facility Operating License No. NPF-62. The amendment revised the

Technical Specifications.

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

FR 49428) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 3, 1994. No significant

hazards consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727.

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 12, 1993

Brief description of amendments: The amendments revise the

Technical Specifications for the accumulators to allow extended action

time for improper boron concentration, to provide a consistent action

statement for both units, and to modify the surveillances on the boron

concentration and the isolation valve.

Date of issuance: November 8, 1994

Effective date: November 8, 1994

Amendment Nos.: 184 and 169

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 67848). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 8, 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: December 22, 1993

Brief description of amendments: The amendments revise the action

statement in the Technical Specifications for Steam Generator Stop

Valves to be more consistent with NUREG-1431, Standard Technical

Specifications Westinghouse Plants. The proposed changes allow both

greater time for compensatory action as well as operation in Modes 2

and 3 with valves inoperable but closed. A Unit 2 action requirement is

also revised.

Date of issuance: November 8, 1994

Effective date: November 8, 1994

Amendment Nos.: 185 and 170

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

revised the Technical Specifications.

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

4939) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 8, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Niagara Mohawk Power Corporation, Docket Nos. 50-220, and 50-410,

Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County,

New York

Date of application for amendments: June 9, 1994

Brief description of amendments: The amendments modify paragraph

2.D(4) of

Facility Operating License No. DPR-63 and paragraph 2.E of

Facility Operating License No. NPF-69 to require compliance with

the amended Physical Security Plan. The changes involve the number of

armed security force members that comprise the response force for each

shift at the site.

Date of issuance: October 31, 1994

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

30 days.

Amendment Nos.: Unit 1--150--Unit 2--58

Facility Operating License Nos. DPR-63 and NPF-69: Amendments

revise the Facility Operating Licenses.

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

FR 49432) The Commission's related evaluation of the amendments is

contained in a Safeguards Evaluation Report dated October 31, 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: September 26, 1994

Brief description of amendment: The amendment revises the Technical

Specifications (TS) by adding a footnote to Surveillance Requirement

4.6.1.2.d that defers the performance of Type B and C Containment leak

rate tests to the end of the twelfth refueling outage.

Date of issuance: October 31, 1994

Effective date: October 31, 1994

Amendment No.: 181

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

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (59 FR 52005, October 13, 1994)

That notice provided an opportunity to submit comments on the

Commission's proposed no significant hazards consideration

determination. No comments have been received. The notice also provided

for an opportunity to request a hearing by November 14, 1994, but

indicated that if the Commission makes a final no significant hazards

consideration determination any such hearing would take place after

issuance of the amendment. The Commission's related evaluation of the

amendment, finding of exigent circumstances, and final determination of

no significant hazards consideration are contained in a Safety

Evaluation dated October 31, 1994.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

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: September 1, 1994

Brief description of amendment: The amendment revises the Technical

Specifications concerning the Reactor Coolant System Volume.Date of

issuance: November 8, 1994

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

30 days.

Amendment No.: 182

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

Technical Specifications.

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

FR 49432). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 8, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of application for amendments: February 16, 1994 (Reference

LAR 94-04)

Description of amendment request: The proposed amendments revise

the combined Technical Specifications (TS) for the Diablo Canyon Power

Plant, Unit Nos. 1 and 2. Specifically, TS 4.2.2, ``Heat Flux Hot

Channel Factor--FQ(z),'' and 6.9.1.8, ``Core Operating Limits

Report,'' are revised as follows: (1) The 2-percent FQ(z) penalty

listed in TS 4.2.2.2.e.1) would be deleted and the statement revised to

indicate the use of an appropriate factor to be specified in the Core

Operating Limits Report (COLR). (2) TS 6.9.1.8.b.1 would be changed to

reference Revision 1 of WCAP 10216-P-A, ``Relaxation of Constant Axial

Offset Control FQ(z) Surveillance Technical Specification,'' dated

February 1994.

Date of issuance: October 31, 1994

Effective date: 60 days from date of issuance.

Amendment Nos.: 96 and 95

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

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

17603) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 31, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407No significant hazards

consideration comments received: No.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

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

17603) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 31, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

PECO Energy Company, Docket No. 50-171, Peach Bottom Atomic Power

Station, Unit 1, Peach Bottom, Pennsylvania.

Date of application for amendment: May 9, 1994.

Brief description of amendment: This amendment modified Possession-

Only License No. DPR-12 and the Technical Specifications (TS)

incorporated as Appendix A by changing the name of Philadelphia

Electric Company to PECO Energy Company, by providing the appropriate

10 CFR Part 20 references, and by reducing the required frequency for

performing periodic inspections in the containment vessel below ground

level for water accumulation.

Date of issuance: November 3, 1994.

Effective date: This license amendment is effective as of the date

of its issuance and must be fully implemented no later than 30 days

from the date of issuance.

Amendment No.: 8.Possession-Only License No. DPR-12: The amendment

revised Possession-Only License No. DPR-12 and the TS.

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

45030) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 3, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania

Date of application for amendments: September 16, 1994

Brief description of amendments: These amendments extend the

snubber functional testing interval from 18 months (+/- 25%) to 24

months (+/- 25%) (plus or minus was published as [greater than or equal

to] in the initial Federal Register notice), and increase the sample

plan size from 10 percent to 13.3 percent. The combination of these two

changes will ensure that the entire population of snubbers will be

tested in a 15-year period.

Date of issuance: November 2, 1994

Effective date: November 2, 1994Amendment Nos. 81 and 42

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

FR 50019) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 2, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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, 1992, supplemented

June 27, 1994, and September 26, 1994

Brief description of amendment: The amendment revised Technical

Specifications Section 4.6.B (Emergency Power System Periodic Tests--

Station Batteries) to incorporate changes which allow battery testing

surveillance interval extensions to accommodate operation on a 24-month

fuel cycle. These changes followed the guidance provided in Generic

Letter 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.

Date of issuance: November 2, 1994

Effective date: November 2, 1994

Amendment No.: 155

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

Technical Specifications.

Date of initial notice in Federal Register: October 28, 1992 (57 FR

48825) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 2, 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.

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: July 21, 1994, as supplemented

September 26, 1994.

Brief description of amendment: The amendment relocates fire

protection requirements from the Technical Specifications to the plant

fire protection program in accordance with the guidance provided in

Generic Letter (GL) 86-10, ``Implementation of Fire Protection

Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements

from the Technical Specifications.'' The amendment also modifies the

Facility Operating License to incorporate the standard fire protection

license condition provided in GL 86-10.

Date of issuance: November 3, 1994

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

30 days.

Amendment No.: 218

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

Technical Specifications and License.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42345) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 3, 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.

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: August 19,

1994, as supplemented October 4, 1994

Brief description of amendments: The amendments reduce the minimum

setpoints and allowable values for the steam generator low and low-low

level reactor protection system signals.

Date of issuance: November 4, 1994

Effective date: November 4, 1994

Amendment Nos. 159 and 140

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

revised the Technical Specifications.

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

FR 47180) The supplemental letter provides additional information but

does not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated November 4, 1994.

No significant hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

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: June 13, 1994

Brief description of amendments: These amendments add a new section

3.0.6 to the technical specifications and the associated Bases, that

permits an out-of-service component to be returned to service under

administrative controls for the purpose of determining operability, and

make an editorial correction.

Date of issuance: November 8, 1994

Effective date: November 8, 1994

Amendment Nos. 160 and 141

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

revised the Technical Specifications.

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

39590) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 8, 1994.No significant

hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

South Carolina Electric & Gas Company, South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: December 17, 1993

Brief description of amendment: The change revises TS 3/4.3.3.6,

Accident Monitoring Instrumentation, and its associated bases;

relocates TS 3/4.6.5.1, Hydrogen Monitors, and TS 3/4.3.3.1, Tables

3.3-6 and 4.3-3, Item 1.c, Reactor Building Area High Range Radiation

Monitors, into the Accident Monitoring TS.

Date of issuance: November 7, 1994

Effective date: November 7, 1994

Amendment No.: 118

Facility Operating License No. NPF-12. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 16, 1994 (59

FR 7699) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 7, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Fairfield County Library,

Garden and Washington Streets, Winnsboro, South Carolina 29180.

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: May 16, 1994 (TS 94-03)

Brief description of amendments: The amendments remove the response

time limits for the reactor trip and engineered safety feature

functions from the technical specifications in accordance with Generic

Letter 93-08.

Date of issuance: November 9, 1994

Effective date: November 9, 1994

Amendment Nos.: 190 and 182

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

revise the technical specifications.

Date of initial notice in Federal Register: June 22, 1994 (59 FR

32236) The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated November 9, 1994No significant

hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

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

Date of amendment request: September 19, 1994

Brief description of amendments: The amendments allow a one-time

six-month extension for certain emergency diesel generator technical

specification surveillance requirements and other related surveillance

requirements. The one-time extension from 18 to 24 months for the

affected surveillance requirements is applicable only to Unit 2, Train

A, until completion of the second refueling outage for Unit 2.

Date of issuance: November 2, 1994

Effective date: Effective as of its date of issuance, to be

implemented within 30 days.

Amendment Nos.: Unit 1--Amendment No. 31; Unit 2--Amendment No. 17

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

revised the Technical Specifications.

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

FR 50024) The October 20, 1994, submittal provided additional

clarifying information and did not change the initial no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

November 2, 1994. No significant hazards consideration comments

received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

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

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

Date of amendment request: September 19, 1994

Brief description of amendments: The amendments revise the

technical specifications by eliminating the requirement that the 18-

month surveillance requirements (SRs) for the emergency core cooling,

containment spray, spray additive, containment isolation valves,

auxiliary feedwater and component cooling water systems be performed

``during shutdown'' or ``during REFUELING MODE or COLD SHUTDOWN.'' The

SRs are still required to be performed on an 18-month surveillance

interval, but may be performed in any mode in which it is technically

and operationally acceptable to perform the testing.

Date of issuance: November 2, 1994

Effective date: Within 30 days of its date of issuance

Amendment Nos.: Unit 1--Amendment No. 32; Unit 2--Amendment No. 18

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

revised the Technical Specifications.

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

FR 50022) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 2, 1994.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

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

NuclearPower Plant, Kewaunee County, Wisconsin

Date of application for amendment: March 31, 1994

Brief description of amendment: The amendment revises the Kewaunee

Nuclear Power Plant Technical Specifications (TS) by incorporating

operability and surveillance requirements for the recently installed

Auxiliary Feedwater Pump Low Discharge Pressure Trip instrumentation.

Surveillance requirements were added to Table TS 4.1-1, ``Minimum

Frequencies for Checks, Calibrations and Test of Instrument Channels.''

TS 3.4, ``Steam and Power Conversions System,'' has been revised to

explicitly link operability of the associated Auxiliary Feedwater Pump

Low Discharge Pressure Trip channel to operability of the associated

auxiliary feedwater pump. In addition, minor format inconsistencies in

TS 3.4.b.1.A and 3.4.b.1.B were corrected.

Date of issuance: November 1, 1994

Effective date: Date of issuance, to be implemented within 30 days

Amendment No.: 112

Facility Operating License No. DPR-43. Amendment revised the

Technical Specifications.

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

34671) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 1, 1994. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin

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

54301.

Dated at Rockville, Maryland, this 16th day of November 1994.

For the Nuclear Regulatory Commission.

Steven A. Varga,

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

Regulation

[Doc. 94-28758 Filed 11-22-94; 8:45 am]

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