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

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice Applications and Amendments to Facility Operating

Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from March 7, 1994, through March 18, 1994. The

last biweekly notice was published on March 16, 1994.

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 P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, 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 20555.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By April 29, 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 20555 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 20555, 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 20555, and at the local public document

room for the particular facility involved.

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

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

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: February 18, 1994

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Figure 3.2-1, ``REACTOR COOLANT

COLD LEG vs CORE POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD

LEG TEMPERATURE,'' for Units 1 and 3 to include the cold leg

temperature between 552 deg.F and 562 deg.F at core power levels

between 90 percent and 100 percent within the AREA OF ACCEPTABLE

OPERATION. Also, the proposed amendment would modify TS 3/4.1.1.4,

``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES 3/4.1.1.4, ``MINIMUM

TEMPERATURE FOR CRITICALITY,'' for all units to allow the minimum

temperature for criticality to be established at 545 deg.F, rather than

the current value of 552 deg.F, to establish the surveillance

temperature at 552 deg.F, rather than the current 557 deg.F, and to

clarify the BASES for this TS.

Basis for proposed no significant hazards consideration

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

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

analyses performed confirmed that the existing safety analysis for

cycle 5 of all three PVNGS [Palo Verde Nuclear Generating Station]

units remains valid for a 10 deg.F reduction in RCS [reactor coolant

system] temperature.

Standard 2 -- Create the possibility of a new or different kind of

accident from any accident previously evaluated.

This amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated. The

analyses performed demonstrated that the current licensing basis

analyses results remain valid with a 10 deg.F reduction in RCS [reactor

coolant system] temperature, and that the safety system settings remain

unchanged.

Standard 3 -- Involve a significant reduction in a margin of

safety.

This amendment request will not involve a significant reduction in

a margin of safety. There is no reduction in the margin of safety since

the changes apply only to the reactor coolant cold leg temperature and

the minimum temperature for criticality, the safety analyses have been

reevaluated (and reperformed where necessary) using the new

temperature, and the results remain valid. All other safety limits and

safety system settings remain unchanged. Therefore, there is no

reduction in any margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

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

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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: February 4, 1994

Description of amendment request: The proposed amendment revises

the Action Statement of Technical Specification 3.6.5, Vacuum Relief

System, to require that in Modes 1-4 with one vacuum relief system

inoperable the system be restored to operable status within seventy-two

hours or be in at least hot standby within the next six hours.

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 amendment does not physically alter the plant in

any manner. The proposed amendment does not introduce any new

equipment nor does it require any existing equipment or systems to

perform a different type of function than they are currently

designed to perform. The proposed amendment to Technical

Specification 3.6.5 allows additional time to restore an inoperable

containment vacuum relief system to operable status. Changing the

completion time to seventy-two hours remainsquite conservative for

this non-ESF system since a seventy-two hour restoration time is

specified for two-train ESF systems which mitigate Final Safety

Analysis Report (FSAR) Chapter 15 accidents. The CVRS [containment

vacuum relief system] is designed to protect the structural

integrity of containment during an inadvertent actuation of the

containment spray system, which is not an FSAR Chapter 15 accident.

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 amendment does not introduce any new equipment nor

does it require any existing equipment or systems to perform a

different type of function than they are currently designed to

perform. 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.

The proposed amendment to Technical Specification 3.6.5 allows

additional time to restore an inoperable containment vacuum relief

system to operable status. Changing the completion time to seventy-

two hours remains conservative since a seventy-two hour restoration

time is specified for two-train ESF systems which mitigate FSAR

Chapter 15 accidents. The CVRS is designed to protect the structural

integrity of containment during an inadvertent actuation of the

containment spray system, which is not an FSAR Chapter 15 accident.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety as defined in the Technical

Specifications of FSAR.

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: S. Singh Bajwa

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 amendment request: February 24, 1994

Description of amendment request: The proposed amendments would

provide surveillance requirements for a planned modification to the

Keowee emergency power generators' underground power path breaker

closing logic. The planned modification would provide an automatic

close feature for the underground path breakers under certain specified

conditions. The modification is needed to correct a design deficiency

which resulted in a single failure vulnerability when both Keowee units

are in their normal alignment. The single failure vulnerability is

being prevented by means of administrative controls pending

implementation of a permanent corrective action. The proposed

amendments would add an annual operability test to Technical

Specification 4.6, Emergency Power Periodic Testing, of the automatic

close feature.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

The Keowee Hydro units provide the main source of emergency

power for the Oconee Nuclear units, but they are not accident

initiators. The FSAR [Final Safety Analysis Report] Loss of Electric

Power Accident assumes two types of events: (1) Loss of load (unit

trip) and (2) Loss of all system and station power. The changes

performed by the modification that added the automatic closure

circuitry do not increase the likelihood of either. Also, the

modifications to the Keowee operating logic will not adversely

affect the ability to mitigate LOOP [Loss of Offsite Power], LOCA

[Loss of Coolant Accident], and LOCA/LOOP accidents as described in

the FSAR. The loss of all station power accident analysis

assumptions are still valid. This modification has no adverse impact

on the ability of the Keowee Units to satisfy their design

requirements to achieving rated speed and voltage within 23 seconds

of receipt of an emergency start signal.

The surveillance change that is included in [the] amendment

request is provided to assure the availability of the electrical

power systems for mitigation of Design Basis Accidents (DBAs). As

described within the technical justification [from the licensee's

application], the Keowee breaker circuitry was modified to allow the

Keowee Unit that is aligned to the overhead power path to

automatically close to the underground power path if the postulated

fault occurs. The surveillance change is an additional restriction

not presently included in the Technical Specifications. [The]

amendment will ensure the operability of the Keowee Unit ACB [Air

Circuit Breaker] automatic close feature and will assure that proper

testing requirements are maintained.

Based on the above and the technical justification provided in

[the amendment application], there is no significant increase in the

probability of a DBA as a result of this change, nor is there a

significant increase in the consequences of a DBA as a result of

this change since the proposed amendment assures the availability of

the electrical power system.

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

accident from any kind of accident previously evaluated:

The proposed change makes physical changes to the plant

configuration. However, the modification simply changes the Keowee

control logic to remove the possibility of a certain postulated

failure from causing a loss of emergency power to the Oconee nuclear

units. The Keowee emergency power systems will remain operable and

available to mitigate accidents. Operation of ONS [Oconee Nuclear

Station] in accordance with [the] Technical Specifications will not

create any failure modes not bounded by previously evaluated

accidents. Consequently, this change will not create the possibility

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

previously evaluated.

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

Margins of safety associated with [the] Technical Specifications

have been evaluated. No safety or design limits are adversely

affected, so margins of safety as defined in the bases to any

Technical Specifications are not reduced as a result of the Keowee

modification. The design basis of the auxiliary electrical system is

to supply the required ES [Engineered Safeguards] loads of one Unit

and safe shutdown loads of the other two units. The Technical

Specification amendment includes an additional surveillance

restriction not presently included in the Technical Specifications.

The proposed amendment assures the continued availability of the

electrical power systems; thus preserving the existing margin of

safety. Therefore, there will be no 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

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

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

NRC Project Director: David B. Matthews, Director

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

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

Date of amendment request: September 7, 1993, as supplemented

February 8, 1994

Description of amendment request: The proposed amendment would

revise the Physical Security Plan (PSP).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below.

The accident mitigation features of the plant are not affected by

the proposed compensatory measures for protecting the site during

periods when security systems are degraded and therefore no decrease

occurs in the effectiveness of the security program to protect against

radiological sabotage or increased risk to the public health and

safety. This is due to continued compliance with existing regulatory

requirements and other commitments within the security plan. These

changes have no impact on the design basis security threat and

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

accident. New systems, modes of equipment operation, failure modes or

other plan situations are not introduced by these changes. The proposed

changes allow flexibility for the use of compensatory measures and do

not change any safety limits, LCOs, or surveillance requirements on

equipment to operate the plant.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

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

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

Date of amendment request: February 14, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to reflect changes that have

been made to 10 CFR Part 20 AND 10 CFR 50.36a.

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 revisions to the liquid and gaseous concentration

release rate limits will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because there will be no change in the types and amounts of

effluents that will be released, nor will there be an increase in

individual or cumulative occupational radiation exposures.

The administrative changes for definitions, terminology,

paragraph references, and record keeping requirements are necessary

so that the Waterford 3 Technical Specifications will remain

consistent with the revised federal regulations (i.e., 10CFR20 and

10CFR50.36a). Record retention and reporting requirements will

continue to meet NRC regulations. These changes are administrative

in nature and do not affect plant hardware or operation.

Restricting access to high radiation areas via guards rather

than locked doors provides operational flexibility while continuing

to meet the underlying intent of precluding unauthorized access.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

Changes to the liquid and gaseous concentration limits are

necessary to provide adequate operational flexibility. Operational

history at Waterford 3 has demonstrated that the use of

concentration values associated with the old 10CFR20.106

requirements has resulted in calculated maximum individual doses to

a member of the public that are small percentages of the limits of

10CFR50, Appendix I. The proposed revisions will not create the

possibility of a new or different kind of accident from any

previously evaluated because the revisions will not change the types

and amounts of effluent that will be released.

The administrative changes for definitions, terminology,

paragraph references, and record keeping are necessary so that the

Technical Specifications will remain consistent with the revised

federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record

retention and reporting requirements will continue to meet NRC

regulations. These changes are administrative in nature and do not

affect plant hardware or operation.

Restricting access for ALARA [as low as reasonably achievable]

with guards rather than locked doors will continue to meet the

underlying intent of the TS. These changes do not involve plant

hardware or operation.

Therefore, the proposed changes will not create the possibility

of a new or different kind of accident from any previously

evaluated.

The proposed revisions do not involve any changes in the types

or increases in the amounts of effluents released off site. The

methodology used to control radioactive effluents and calculate

effluent monitor setpoints will result in the same effluent release

rate as the current methodology. The basic requirements for TS

concerning effluent releases (10CFR50.36a) indicate that compliance

with TS will keep average annual release to small percentages of

10CFR20 limits. For liquid effluent releases, the annual dose of 500

mrem, that is the bases for the concentrations in the new 10CFR20.

The 50.36a requirements further indicate that operational

flexibility is allowed, compatible with considerations of health and

safety, which may temporarily result in release higher than such

small percentages, but still within the limits specified in the old

10CFR20.106 that references Appendix B maximum permissible

concentrations (MPCs). For gaseous effluent releases, the limits

associated with the gaseous release rate TS will be maintained at

the current instantaneous dose rate limits. Compliance with the

limits of the new 10CFR20.1301 will be demonstrated by operating

within the limits of 10CFR50, Appendix I, and 40CFR190. The revision

will not change the types and amounts of effluent that will be

released.

The administrative changes for definitions, terminology,

paragraph references, and record keeping are necessary so that the

Technical Specifications will remain consistent with the revised

federal regulations (i.e., 10CFR20 and 10CFR50.36a). Record

retention and reporting requirements will continue to meet NRC

regulations. These changes are administrative in nature and do not

affect plant hardware or operation.

Controlling access to high radiation areas for ALARA can be

performed effectively by guards in place of locked doors. These

changes do not involve plant hardware or operation.

Therefore, the proposed changes will not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Florida Power and Light Company, et al., Docket No. 50-335, St.

Lucie Plant, Unit No. 1, St. Lucie County, Florida

Date of amendment request: February 22, 1994

Description of amendment request: The proposed amendment modifies

the minimum stored borated water inventory requirements for Operational

Modes 1 through 4 by revising Figure 3.1-1 and Limiting Condition for

Operation (LCO) 3.1.2.8 of the unit Technical Specifications (TS). The

associated bases for TS 3/4.1.2 are also revised to reflect the

bounding borated water makeup volumes, as a function of boric acid

concentration, which define the proposed inventory requirements. The

proposed amendment will significantly improve operational flexibility

with no risk to plant safety and will provide for consistency of

operation between the two St. Lucie units.

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 amendment will reduce the minimum borated water

inventory required to be stored in the Boric Acid Makeup Tanks

(BAMT) during unit operation in Modes 1 through 4. The reduction in

BAMT inventory will not affect any equipment postulated to

malfunction in the Updated Final Safety Analysis Report (UFSAR) to

initiate an accident nor will it impact the operation of any other

equipment whose malfunction could adversely affect safety-related

structures, systems, or components. Credit is not taken for boron

addition to the Reactor Coolant System from the BAMTs for purposes

of reactivity control in accidents analyzed in the UFSAR. The

minimum required capability to achieve and maintain safe shutdown

for such events has not been altered. Therefore, operation of the

facility in accordance with the proposed amendment will 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 reduction in minimum required BAMT inventory does not change

the boration system function, configuration, operation, or design

basis as described in the UFSAR. The proposed change does not alter

the modes of plant operation and does not affect the operation of

safety-related structures, systems, or components. Therefore,

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.

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

amendment would not involve a significant reduction in a margin of

safety.

The reduced BAMT minimum inventory requirements are defined by

analyses that utilize an approved plant cooldown scenario and

conservative physics parameters representative of the present and

future planned reactor core designs for St. Lucie Unit 1. The

analytical methodology employed to determine the revised inventory

requirements is the same as that used to establish the existing

inventory requirements. The existing reactivity control Limiting

Conditions for Operation (LCO) related to safe shutdown margins and

redundant boron flow paths have not been altered. Sufficient

quantities of borated water will continue to be stored in the BAMTs

to assure compliance with these LCOs during the prescribed plant

operating modes. Therefore, operation of the facility in accordance

with the proposed amendment would not involve a significant

reduction in a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this

proposed license amendment involves no significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: Herbert N. Berkow

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: February 18, 1994

Description of amendment request: The licensee proposes to change

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

the frequencies specified for audits performed under the cognizance of

the Company Nuclear Review Board (CNRB). The periodicity of the audits

for these activities will be controlled as described in the licensee's

Topical Quality Assurance Report (FPLTQAR), wherein the minimum audit

frequency for any activity is established as biennial unless the audit

is otherwise required to be performed more frequently by the TS, Code

of Federal Regulations, or other licensing commitments. Periodic audits

of selected aspects of operational phase activities are performed with

a frequency commensurate with safety significance. During the interval

between the periodic audits, continuing performance evaluations are

conducted of activities important to plant safety.

In addition, the licensee proposes to revise the TS in accordance

with Generic Letter 93-07. Generic Letter 93-07, ``Modifications of the

Technical Specifications Administrative Control Requirements for

Emergency and Security Plans,'' issued December 28, 1993, provided

guidance for changes to the TS to remove the audit of the emergency and

security plans and implementing procedures from the list of

responsibilities of the company nuclear audit and review group. The

basis of this change is that Parts 50 and 73 of Title 10 of the Code of

Federal Regulations (10 CFR) include provisions that are sufficient to

address these requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments relocate the administrative control

criteria for minimum audit frequencies from the facility TS to the

FPL Quality Assurance (QA) Program. The QA Program is described in

the FPL Topical Quality Assurance Report pursuant to 10 CFR 50,

Appendix B. In addition, the proposed amendments in accordance with

Generic Letter 93-07, changes the TS to remove the audit of the

emergency and security plans and implementing procedures from the

list of responsibilities of the Company Nuclear Review Board. The

changes being proposed are administrative in nature and do not

affect assumptions contained in plant safety analyses, the physical

design and/or operation of the plant, nor do they affect the TS that

preserve safety analysis assumptions. Therefore, operation of the

facility in accordance with the proposed amendments would not affect

the probability or consequences of an accident previously analyzed.

(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 changes being proposed are administrative in nature and will

not change the physical plant or the modes of operation defined in

the Facility License. The change does not involve the addition or

modification of equipment nor does it 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

amendments would not involve a significant reduction in a margin of

safety.

The changes being proposed are administrative in nature and do

not alter the bases for assurance that safety-related activities are

performed correctly or the basis for any TS that is related to the

establishment of or maintenance of a 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: 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: Herbert N. Berkow

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

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

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

Matagorda County, Texas

Date of amendment request: March 14, 1994

Description of amendment request: The licensee proposes to make a

change to the technical specifications to add a new Limiting Condition

For Operation (LCO), 3.0.6. LCO 3.0.6 will allow equipment removed from

service or declared inoperable to comply with actions to be returned to

service, under administrative controls, solely to perform testing. The

new LCO will permit non-compliance with the applicable Action statement

to perform the post-maintenance and surveillance testing required to

demonstrate the operability of the equipment being returned to service

or the operability of other equipment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The implementation of LCO 3.0.6 will allow the orderly and

judicious return to service of inoperable equipment. This LCO will

permit equipment removed from service to comply with required actions

to be returned to service under administrative controls to verify the

component or system will perform its safety function. The

administrative controls will ensure the time involved will be limited

to only the time required to demonstrate the component or system's

operability. The implementation of this new LCO will provide an

acceptable method of testing technical specification equipment prior to

its return to operable service following required maintenance. These

actions will ensure that the equipment being returned to service is

capable of performing its designed safety function prior to being

declared operable. Therefore, this action will ensure the probability

or consequences of an accident previously evaluated are not

significantly increased.

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

different kind of accident from any previously evaluated.

The equipment is only being tested in its designed configuration or

being returned to service to allow testing of another component or

system. Therefore, the use of this new LCO will not result in a new or

different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction in

the margin of safety.

The use of the new LCO will only allow the return to service of

equipment that is expected to operate as designed. The use of the LCO

will be limited to the performance of testing on the equipment being

returned to service or on other equipment that is dependent on the

equipment being returned to service. This testing is limited to post-

maintenance testing and the testing necessary to prove operability.

Since the equipment will be controlled by administrative requirements

that will ensure all necessary actions are taken, this change does not

involve a significant reduction in a margin of safety.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Wharton County Junior College,

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

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

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

NRC Project Director: Suzanne C. Black

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 amendment requests: February 22, 1994

Description of amendment requests: The proposed amendments would

modify the technical specifications to reduce surveillance requirements

for testing during power operation. This modification was proposed to

licensees in NRC 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:

Criterion 1

Although the surveillance requirements are lessened by these

proposed changes, the changes are consistent with those found

acceptable by the NRC in Generic Letter 93-05. The proposed changes

have been determined to be compatible with our plant operating

experience. Based on these considerations, it is concluded that the

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Criterion 2

The proposed changes do not involve physical changes to the

plant or changes in plant operating configuration. The changes only

involve frequency of testing required to be performed. The changes

are consistent with those found to be acceptable by the NRC in

Generic Letter 93-05. Thus, it is concluded that the proposed

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

accident from any accident previously evaluated.

Criterion 3

Although the surveillance requirements are lessened by these

proposed changes, the changes are consistent with those found

acceptable by the NRC in Generic Letter 93-05. The proposed changes

have been determined to be compatible with our plant operating

experience. Based on these considerations, it is concluded that the

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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Ledyard B. Marsh

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of amendment request: February 15, 1994

Description of amendment request: The proposed amendment would

delete from the Technical Specifications the operational and

surveillance requirements for the turbine overspeed protection system.

The licensee intends to continue testing of the overspeed protection

system as part of plant procedures.

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:

(a) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences for an accident previously

evaluated. The proposed deletion of the turbine overspeed protection

T/S [technical specification] will not significantly change the

surveillance tests on the Unit 2 turbine. The surveillance schedule

and tests will be under administrative procedures outside of the TSs

similar to that of Unit 1 and will be in line with operating

experience at Cook Nuclear Plant and applicable industry experience.

The Unit 2 turbine is now operating in its ninth operating cycle

with over 90,000 hours of operation. Turbine overspeed protection

surveillance results have been very good since unit startup in 1978.

In 1983, a wear problem was found with the overspeed plungers.

Replacement plungers were installed. Then in 1988, these plungers

were replaced with parts having stellited (hardened) surfaces. There

have been no subsequent problems. Our expectation is that the

turbine overspeed protection system will remain available to perform

its function of preventing excessive turbine overspeed. Lastly, the

STS [Standard Technical Specifications] developed by the MERITS

program in NUREG-1431 do not include a T/S for turbine overspeed

protection. The omission of an overspeed protection T/S in NUREG-

1431 indicates that a T/S is not needed to ensure an adequate level

of safety for a nuclear facility. This view is supported by WCAP

11618 which uses the NRC's ``Interim Policy Statement Criteria'' to

evaluate the need for a turbine overspeed protection T/S and

concludes that it is not needed. For these reasons, we believe that

deleting the turbine overspeed protection T/S will not significantly

increase the probability or consequences of an accident previously

evaluated.

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

accident from any previously analyzed.

The proposed amendment does not create the possibility of a new

or different kind of accident from any previously evaluated. This

request to delete the turbine overspeed protection T/S eliminates a

control on the surveillance testing of the Unit 2 turbine. The

design function of the turbine overspeed protection and the

operation of the turbine/generator remain the same. The operating

history of the Unit 2 surveillance results to date and our continued

testing support the view that the turbine overspeed protection will

remain available. For these reasons, we believe that the proposed

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

of accident from any previously analyzed.

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

The proposed amendment does not involve a significant reduction

in the margin of safety. Turbine overspeed protection surveillance

results have been excellent since 1983. The years of operating data

well within acceptance criteria on Unit 2 turbine overspeed

protection provide ample evidence that there is no significant

degradation of the system to perform its function. The reliability

of the overspeed protection was improved by the replacement of the

plungers with parts having stellited surfaces. The surveillance

schedule and tests will be based on operating experience at Cook

Nuclear Plant and applicable industry experience. Surveillance

testing will continue under an administrative program outside of

TSs. Thus the turbine overspeed protection is expected to remain

available. Also by eliminating this T/S we will be reducing the

potential for shutting down the unit because of difficulties

performing this T/S surveillance unrelated to the functionality of

the valves and overspeed trip protection. Lastly, the STS developed

by the MERITS program in NUREG-1431 do not include a T/S for turbine

overspeed protection. The omission of an overspeed protection T/S in

NUREG-1431 indicates that a T/S is not needed to ensure an adequate

level of safety for a nuclear facility.

This view is supported by WCAP 11618 which uses the NRC's

``Interim Policy Statement Criteria'' to evaluate the need for a

turbine overspeed protection T/S and concludes that it is not

needed. For these reasons, we believe that the turbine overspeed

protection system will remain operable and so this proposed

amendment 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: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Ledyard B. Marsh

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of amendment request: February 22, 1994

Description of amendment request: The proposed amendment would

revise the reactor coolant system heatup and cooldown curves in the

Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to the P-T [pressure-temperature] curves

are being updated as a result of the Unit 2 Capsule U analysis,

WCAP-13515. The analysis was required per the removal schedule

established in Table 4.4-5 of the Cook Nuclear Plant Technical

Specifications. The analysis was performed based on guidance from R/

G 1.99 [Regulatory Guide 1.99, ``Radiation Embrittlement of Reactor

Vessel Materials''], Revision 2. The change only involves a revised

time frame for material qualification from 12 EFPY [effective full-

power years] to 15 EFPY as supported by the aforementioned

Westinghouse analysis. Therefore, we conclude that the changes will

not involve a significant increase in the probability or

consequences of a previously evaluated accident, nor will the

changes involve a significant reduction in a margin of safety.

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

accident from any previously analyzed.

The proposed changes do not involve any physical modifications

to the plant. Therefore, the changes should not create the

possibility of a new or different kind of accident from any

previously analyzed or evaluated.

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

See the response to (1) above.

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: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: Ledyard B. Marsh

Long Island Power Authority, Docket No. 50-322, Shoreham Nuclear

Power Station, Unit 1 (SNPS), Wading River, New York

Date of application for amendment: Amendment No. 11, November 4,

1993 (Reference LSNRC-2115)

Brief description of amendment: This license amendment request

(LSNRC-2115) proposes to delete from the Possession-Only License (POL)

the requirements associated with the safe storage and handling of

irradiated fuel, the accompanying Appendix A of SNPS Technical

Specifications, and Appendix B of SNPS Environmental Protection Plan

(non-radiological). This proposed amendment will update the SNPS POL to

reflect the status of the facility after irradiated fuel removal from

the site. SNPS License Condition No. 3 prohibits this amendment from

being implemented until all the fuel has been removed from SNPS, and

the licensee has certified to the NRC that all the fuel has been

removed.

Basis for the proposed no significant hazards consideration

determination: In accordance with the requirements and standards in 10

CFR 50.92(c), the licensee has provided an analysis of the issues

related to the no significant hazards consideration.

The licensee's analysis of the issues related to no significant

hazards consideration are presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes will become effective after the fuel and

its related hazards are removed from the site.

Therefore, the proposed changes will update the SNPS license to

reflect the facility status after the removal of irradiated fuel.

This action will not increase the probability or consequences of any

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes will update the license by deleting

requirements which will no longer apply to SNPS and will not have an

adverse impact on the operation of the remaining plant systems and

components.

Therefore, the proposed change does not create the possibility

for an accident or malfunction different from any previously

analyzed.

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

of safety?

The proposed change will update the license to reflect the

status of the facility after the removal of irradiated fuel from the

site.

Therefore, the proposed changes will not reduce the margins of

safety for the remaining plant systems and components.

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

this review the three standards of 50.92(c) are satisfied. The NRC

staff agrees with the licensee's analysis and has determined that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Shoreham Wading River Public

Library, Shoreham Wading River High School, Route 25A, Shoreham, NY

11792

Attorney for licensee: Mr. W. Taylor Reveley, III, Hunton and

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

VA 23219-4074

NRC Branch Chief: John H. Austin

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

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

County, Minnesota

Date of amendment requests: February 14, 1994

Description of amendment requests: The proposed amendments would

revise Technical Specifications to reflect the new configuration for

the Unit 1 480V safeguards bus arrangement (two 480V safeguards buses

fed by each 4160V safeguards bus). This would make the specifications

the same for both units since the configuration for the two units will

become the same during the 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:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

SBO/ESU [Station Blackout/Electrical Safeguards Upgrade] Project

modifications as reflected in the proposed Technical Specifications

changes were evaluated to determine their impact, if any, on

potential transients and accidents as described in the Prairie

Island USAR [Updated Safety Analysis Report]. Each transient and

accident was evaluated in terms of the mitigating actions described

or assumed in the USAR analysis. The role of the modified systems in

mitigating the event was analyzed in order to evaluate whether the

modification:

(1) changed, degraded or prevented actions described or assumed

in the USAR analysis;

(2) altered any assumptions made in evaluating the radiological

consequences of the accident;

(3) played a direct part in mitigating the radiological

consequences of the accident; or

(4) affected any fission product barrier.

The evaluation demonstrated that the USAR transient and accident

analyses remain valid and bounding.

As part of the evaluation, the revised emergency diesel

generator load sequence was analyzed and found to be bounded by the

existing analyses.

In particular, the USAR analyses of the loss of offsite power

(LOOP) event and the large break loss of coolant accident (LBLOCA)

remain valid and bounding. In addition, the current USAR analysis

for the radiological consequences of a LBLOCA remains valid.

Further, the plant response to a loss of AC power event is not

degraded as a result of these changes but, in fact, is significantly

improved.

In order to determine the effect of the modifications upon the

probability and consequences of an accident, the following items

were specifically evaluated:

(1) the applicable design, material and construction standards;

(2) instrumentation accuracies and response times;

(3) the equipment operating and design limits, including

electrical bus loading, emergency diesel generator loading and

battery loading;

(4) the system interfaces;

(5) voltage margins; and

(6) coordination of protective devices.

Structures, systems and components involved in the modifications

were evaluated as follows:

(1) The design specifications for the new structures, systems

and components were considered for the following requirements:

- seismic;

- separation including control/power circuit interaction,

redundancy/separation of systems, and isolation between safety and

non-safety circuits;

- environmental parameters;

- severe meteorological events;

- missiles; and

- fire protection.

All structures, systems and components meet the appropriate

design requirements for their respective classifications.

(2) Structures, systems and components were additionally

evaluated for the following:

- Structural loads were determined for new cable runs in the

existing plant and for new cable penetrations in the existing

structures.

- New electrical loads requirements were determined.

- System/equipment protection features have been maintained in

the modification.

- Support system performance was specified to maintain the

safety function of the equipment.

- System/equipment redundancy and independence is maintained.

- The frequency of operation of existing equipment was evaluated

and determined not to be affected.

- The testing requirements imposed on new structures, systems

and components are in accordance with their safety classification.

Failures of systems and components involved in the modifications

were analyzed, and it was determined that all safety functions were

maintained.

Required engineered safeguards features loads are accommodated

with the improved auxiliary electrical systems configuration; and,

as demonstrated by the performance of a failure modes and affects

analysis, no single failure will prevent the modified plant from

performing its required safety function in the event of an accident

on either unit.

For the reasons discussed above, the proposed amendment does not

significantly increase the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The SBO/ESU Project modifications as reflected in the proposed

Technical Specifications changes were evaluated to determine if they

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

from any accident previously evaluated.

The modifications were evaluated to determine the types of

accidents which could result from malfunction of the new/modified

structures, systems and components. It was determined that no new or

different kinds of accidents from those previously evaluated are

created. USAR analyses remain bounding.

For these reasons, 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 will not involve a significant

reduction in the margin of safety.

The new Unit 1 480V safeguards configuration provides additional

circuit breakers for improved motor control center (MCC) feeder

circuit coordination by eliminating subfed 480V MCCs from safeguards

480V buses. The proposed Technical Specification changes identify

the new 480V buses and require the operability of both of the buses

per train rather than the one bus per train of the current

configuration and current Technical Specification requirements.

Since the operability requirements are not decreased nor are the

allowed out-of-service times increased by the proposed changes, the

margin of safety is maintained.

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 requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: Ledyard B. Marsh

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 amendment requests: February 16, 1994 (Reference LAR 94-03)

Description of amendment request: The proposed amendments would

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

Power Plant Unit Nos. 1 and 2 to revise TS 4.6.1.2, ``Containment

Integrity.'' The specific TS changes proposed are as follows:

(1) The requirement to conduct three Type A tests specifically at

40 plus or minus 10 month intervals during each 10-year service period

would be replaced with a requirement to conduct three Type A tests at

approximately equal intervals during each 10-year service period.

(2) The requirement to conduct the third Type A test of each set

during the shutdown for the 10-year plant inservice inspection would be

deleted.

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:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes do not affect the initiation of any

accident, nor do the proposed changes involve modifications to any

plant equipment.

The proposed change to the schedule provides flexibility in

meeting the current requirement for 3 tests in 10 years and is

consistent with the intent of the 10 CFR 50, Appendix J requirement

to perform Type A tests at approximately equal intervals. The test

type and test method used for Type A testing would not be changed.

The Type A test acceptance criteria would not be changed, and

containment leakage will continue to be maintained within the

required limits.

Elimination of the requirement to perform the third Type A test

during the shutdown for the 10-year plant ISI does not involve any

modification to plant equipment or affect the operation or design

basis of the containment. These surveillances are independent of

each other and provide assurance of different plant characteristics.

The Type A tests assure the required leak-tightness of the

containment to demonstrate compliance with the guidelines of 10 CFR

100. The 10-year ISI program provides assurance of the integrity of

plant structures, systems, and components and verifies the

operational readiness of pumps and valves in accordance with 10 CFR

50.55a.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes do not involve modifications to any

existing equipment or affect the operation or design basis of the

containment. The proposed changes do not affect the response of the

containment during a design basis accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed changes to the schedule provide flexibility in

meeting the Type A testing schedule requirements. These proposed

changes do not affect or change any limiting conditions for

operation (LCO) or any other surveillance requirements in the TS and

the Bases for the surveillance requirement remains unchanged. The

testing method, acceptance criteria, and bases are not changed and

still provide assurance that the containment will perform its

intended function.

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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120

NRC Project Director: Theodore R. Quay

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania

Date of amendment request: January 10, 1994

Description of amendment request: The proposed amendment would

relocate the seismic monitoring instrumentation Limiting Condition for

Operation, Surveillance Requirements, and associated tables and Bases

contained in TS sections 3.3.7.2 and 4.3.7.2 to the Updated Final

Safety Analysis Report (UFSAR).

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 Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The function of the seismic monitoring instrumentation system is

to monitor the magnitude and effect of a seismic event only, and

cannot initiate or mitigate an accident previously evaluated.

Furthermore, the proposed TS changes to relocate the seismic

monitoring instrumentation requirements from TS to the UFSAR are in

accordance with the criteria for determining those requirements that

should remain in the TS as defined by the NRC in its final policy

statement, ``Final Policy Statement on Technical Specifications

Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The

seismic monitoring instrumentation LCO, SRs, and associated tables

and Bases proposed for relocation from TS to the LGS UFSAR will

continue to be implemented by administrative controls that will

satisfy the applicable requirements of TS section 6 ``Administrative

Controls.'' Those requirements include a review of changes to plant

systems and equipment and to the applicable administrative controls

in accordance with the provisions of 10CFR50.59.

Criterion 2 of the July 22, 1993 NRC final policy statement

states, ``A process variable, design feature, or operating

restriction that is an initial condition of a Design Basis Accident

or Transient analysis that either assumes the failure of or presents

a challenge to the integrity of a fission product barrier.'' The

seismic monitoring instrumentation system is not a system that

monitors a process variable that is an initial condition for

accident or transient analyses. The seismic monitoring

instrumentation is also not a design feature or an operating

restriction that is an initial condition of Design Basis Accident or

transient analyses since it only provides information regarding the

magnitude of and the plant equipment response to a Design Basis

earthquake. Therefore, the current LGS seismic monitoring

instrumentation TS requirements do not meet Criterion 2 of the July

22, 1993 NRC final policy statement.

Criterion 3 of the July 22, 1993 NRC final policy statement

states, ``A structure, system, or component that is part of the

primary success path and which functions or actuates to mitigate a

Design Basis Accident or Transient that either assumes the failure

of or presents a challenge to the integrity of a fission product

barrier.'' The LGS seismic monitoring instrumentation system does

not provide a function or actuate in order to mitigate the

consequences of a Design Basis Accident or transient. Therefore, the

current LGS seismic monitoring instrumentation TS requirements do

not meet Criterion 3 of the July 22, 1993 NRC final policy

statement.

Criterion 4 of the July 22, 1993 NRC final policy statement

states, ``A structure, system or component which operating

experience or probabilistic safety assessment has shown to be

significant to public health and safety.'' Operating experience has

shown that the LGS seismic monitoring instrumentation system has no

impact on public health and safety as defined by the NRC final

policy statement. Furthermore, LGS specific probabilistic risk

assessment (PRA) does not credit the seismic monitoring

instrumentation system as a significant factor in the plant response

to an accident. Therefore, the current LGS seismic monitoring

instrumentation TS requirements do not meet Criterion 4 of the July

22, 1993 NRC final policy statement for determining those

requirements that should remain in TS. This conclusion is consistent

with the function of the seismic monitoring instrumentation system

stated above.

These proposed TS changes will maintain the current operation,

maintenance, testing, and system operability controls of the seismic

monitoring instrumentation system. Furthermore, any future changes

to the seismic monitoring instrumentation system will be evaluated

for the effect of those changes on system reliability as required by

10CFR50.59. The seismic monitoring instrumentation system

performance will not decrease due to these proposed TS changes and

the system will continue to be administratively controlled in

accordance with TS Section 6, including the requirements of

10CFR50.59, thereby precluding a future decrease in its performance.

In accordance with the current TS Section 3.3.7.2, with the

seismic monitoring instrumentation inoperable, the plant would not

be required to shutdown and the provisions of TS Section 3.0.3

(i.e., plant shutdown) would not be applicable. Therefore, the

inoperability of this system and therefore the consequences of an

accident while this system is inoperable, was previously evaluated

as not significant enough to require a change to the plant operating

condition.

Since the seismic monitoring instrumentation system does not

monitor a process variable that is an initial condition for an

accident or transient analyses, or actuates any accident mitigation

feature, and since the operation, maintenance, testing, and

modification of the seismic monitoring instrumentation system will

continue to be administratively controlled, including the

requirements of 10CFR50.59; therefore, maintaining the reliability

of the system, the proposed TS changes will not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The function of the seismic monitoring instrumentation system is

to monitor the magnitude and effect of a seismic event only. The

proposed TS changes to relocate the seismic monitoring instruments

requirements from TS to the UFSAR are in accordance with the

criteria for determining those requirements that should remain in

the TS as defined by the NRC in its final policy statement, dated

July 22, 1993. The seismic monitoring instrumentation system does

not monitor a process variable that is an initial condition for an

accident or transient analyses. The seismic monitoring

instrumentation is also not a design feature or an operating

restriction that is an initial condition of a Design Basis Accident

or transient analyses since it only provides information regarding

the magnitude of and the plant equipment response to a Design Basis

earthquake.

These proposed TS changes to relocate the TS requirements to the

UFSAR will not alter the operation of the plant, or the manner in

which the seismic monitoring instrumentation system will perform its

function, and any future changes will continue to be

administratively controlled in accordance with TS Section 6,

including the requirements of 10CFR50.59.

These proposed TS changes will not impose new conditions nor

result in new types of equipment which will result in different

types of malfunctions of equipment important to safety than any type

previously evaluated.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

These proposed TS changes to relocate the seismic monitoring

instrumentation requirements from TS to the UFSAR are in accordance

with the criteria for determining those requirements that should

remain in the TS as defined by the NRC in final policy statement,

dated July 22, 1993.

Criterion 1 of the NRC final policy statement states,

``Installed instrumentation that is used to detect, and indicate in

the control room, a significant abnormal degradation of the reactor

coolant pressure boundary.'' The NRC final policy statement explains

that ''...This criterion is intended to ensure that Technical

Specifications control those instruments specifically installed to

detect excessive reactor coolant leakage. This criterion should not,

however, be interpreted to include instrumentation to detect

precursors to reactor coolant pressure boundary leakage or

instrumentation to identify the source of actual leakage (e.g.,

loose parts monitor, seismic instrumentation, valve position

indicators).'' Based on the above NRC guidance, the LGS UFSAR, and

TS Bases 3.3.7.2, the seismic monitoring instrumentation does not

detect, and indicate in the control room, a significant abnormal

degradation of the reactor coolant pressure boundary. Therefore, the

current LGS seismic monitoring instrumentation TS requirements do

not meet Criterion 1. Furthermore, operating experience has shown

that the LGS seismic instrumentation system has no impact on public

health and safety as defined by the NRC final policy statement. In

addition, the LGS specific PRA does not credit the seismic

monitoring instrumentation system as a significant factor in the

plant response to accidents.

The seismic monitoring instrumentation LCO, SRs, and associated

tables and Bases proposed for relocation to the LGS UFSAR will

continue to be implemented by administrative controls that will

satisfy the applicable requirements of TS section 6 ``Administrative

Controls.'' Those requirements include a review of future changes to

the system and applicable administrative controls in accordance with

the provisions of 10CFR50.59.

Accordingly, based on the above discussion of NRC specific

guidance, operating experience, and continued imposition of

administrative controls, the proposed TS changes do 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: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

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

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

Date of amendment request: February 3, 1994

Description of amendment request: The licensee commenced operating

on a 24-month fuel cycle, instead of the previous 18-month fuel cycle,

with fuel cycle 9. Fuel cycle 9 started in August 1992; however, the

facility shut down in February 1993 for a ``Performance Improvement

Plan'' outage and a restart date has not yet been established. In order

to accommodate operation on a 24-month cycle after the facility

restarts, the licensee requested an amendment to the Technical

Specifications (TSs) to incorporate the changes listed in items 1-7

below:

(1) The licensee proposed changing the calibration frequency for

the reactor coolant temperature instrument channels (specified in TS

Table 4.1-1) to accommodate operation on a 24-month cycle.

(2) The licensee proposed changing the calibration frequency for

the steam generator level instrument channels (specified in TS Table

4.1-1) to accommodate operation on a 24-month cycle.

(3) The licensee proposed changing the calibration frequency for

the containment pressure instrument channels (specified in TS Table

4.1-1) to accommodate operation on a 24-month cycle.

(4) The licensee proposed changing the calibration frequency for

the steam line pressure instrument channels (specified in TS Table 4.1-

1) to accommodate operation on a 24-month cycle.

(5) The licensee proposed changing the calibration frequency for

the turbine first stage pressure instrument channels (specified in TS

Table 4.1-1) to accommodate operation on a 24-month cycle.

(6) The licensee proposed changing the calibration frequency for

the turbine trip low auto stop oil pressure instrument channels

(specified in TS Table 4.1-1) to accommodate operation on a 24-month

cycle.

(7) The licensee proposed changing the calibration frequency for

the 480V bus undervoltage and alarm relays (specified in TS Table 4.1-

1) to accommodate operation on a 24-month cycle.

These proposed changes follow the guidance provided in Generic

Letter 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.

The licensee also requested the following additional changes:

(1) The addition to TS Table 3.5-5 of limiting conditions for

operation (LCO) requirements for a wide range containment pressure

variable.

(2) The addition of a quarterly functional test surveillance

requirement to Item 4 of TS Table 4.1-1 for the low average temperature

actuation circuits of the reactor coolant temperature channels.

(3) The addition of a second line to Item 14 of TS Table 4.1-1 to

specify surveillance requirements for the wide range containment

pressure instrumentation.

(4) The revision of Item 20 to TS Table 4.1-1 to clarify that both

the reactor trip and the engineered safety features (ESF) actuation

relay logic channels are functionally tested.

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:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no signicant hazards based on the

following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of any accident

previously evaluated?

Response:

The proposed changes do not involve a significant increase in

the probability or consequences of any accident previously

evaluated. The proposed changes extend the calibration intervals

(given in [TS] Table 4.1-1) for the reactor coolant loop temperature

instrumentation used for Engineered Safety Features Actuation

Systems (ESFAS) and Post Accident Monitoring (PAM) functions, the

steam generator (SG) level instrumentation used for ESFAS and PAM

functions, the containment pressure instrumentation used for ESFAS

and PAM functions, the steam line pressure instrumentation used for

ESFAS and PAM functions, the turbine first stage pressure

instrumentation used for ESFAS functions, the 480V bus undervoltage

and alarm relays used for ESFAS functions, and the turbine trip low

auto stop oil pressure instrumentation. These changes are being made

to accommodate a 24 month operating cycle. Other changes include: 1)

the addition to [TS] Table 3.5-5 of limiting conditions for

operation (LCO) requirements for the wide range containment pressure

channels; 2) the addition of a quarterly functional test

surveillance requirement to Item 4 of [TS] Table 4.1-1 for the low

Tavg [average temperature] actuation circuits of the reactor

coolant temperature channels; 3) the addition of a second line to

Item 14 of [TS] Table 4.1-1 to specify the surveillance requirements

for the wide range containment pressure channels; and 4) the

revision of Item 20 to [TS] Table 4.1-1.

Extension of the calibration intervals in question were

evaluated and the results documented in the ESFAS and Indicating

Instrument Surveillance Test Extension reports (References 7 and 8

[Engineered Safety Features Actuation Systems Surveillance Test

Extensions, NYPA document IP3-RPT-ESS-00400, dated May 10, 1993 and

Indicating Instruments Surveillance Test Extensions, NYPA document

IP3-RPT-MULTI-00424, dated May 5, 1993]). ESFAS and indicating

instrument drift analyses were performed to evaluate actual past and

projected future instrument drift. Revised safety system loop

accuracy/setpoint calculations, which include any additional

instrument uncertainties resulting from the proposed calibration

interval extensions, show that sufficient margin exists between the

analytical and field trip settings for the low Tavg, the SG

low-low level, the high and high-high containment pressure, the high

differential steam line pressure, the low steam line pressure, the

high steam flow (dependent upon turbine first stage pressure), the

turbine trip low auto stop oil pressure, and the 480V bus

undervoltage trip functions. Safety analyses are not affected.

Additionally, postulated uncertainties associated with the extended

calibration intervals for the wide range reactor coolant loop

temperature, the narrow and wide range SG level, and the steam line

pressure instrumentation will be accomodated by changes to the

Emergency Operating Procedure (EOP) settings. Extension of the

calibration interval for the narrow range containment pressure

instrumentation channels does not affect EOP settings. Safety

analyses are not affected by the EOP setting changes.

The results of the changes to: 1) add to [TS] Table 3.5-5 LCO

requirements for the wide range containment pressure channels, 2)

add a quarterly functional test surveillance requirement to Item 4

of [TS] Table 4.1-1 for the low Tavg actuation circuits of the

reactor coolant temperature channels, and 3) add a monthly channel

check surveillance requirement to [TS] Table 4.1-1 for the wide

range containment pressure channels are consistent with Westinghouse

Standard Technical Specifications (W STS - Reference 12 [NUREG-1431,

Revision O, ``Standard Technical Specifications - Westinghouse

Plant,'' dated September 28, 1992]). The addition of LCO

requirements to [TS] Table 3.5-5 for the wide range containment

pressure instrumentation, the addition of a quarterly functional

test requirement to Item 4 of [TS] Table 4.1-1 for the low Tavg

actuation circuits, and the separation of surveillance requirements

for the narrow and wide range containment pressure instrumentation

into two lines on [TS] Table 4.1-1 consitute additional technical

specification controls. Changes which consitute additional technical

specification limitations and controls are classified by Federal

Register dated April 6, 1983 (48 FR 14870, April 6, 1983) as not

likely to involve significant hazards considerations. The change to

[TS] Table 3.5-5 ensures conistency with the Authority's commitment

to Regulatory Guide (RG) 1.97 [``Instrumentation for Light-Water-

Cooled Nuclear Power Plants to Assess Plant and Environs Conditions

During and Following an Accident''] for the containment pressure

variable.

The current surveillance requirement specified by Item 20 has

been interpreted by Indian Point 3 as including on-line testing of

both the reactor trip and engineered safety features (ESF) actuation

logic channels, but since the wording may be confusing, this

application proposes to change the wording to clarify that both the

reactor trip and the ESF actuation logic channels are functionally

tested at least every two months on a staggered basis (i.e., one

train per month). The change is consistent with W STS and only

involves a wording change which strengthens the Technical

Specification requirement. The change does not involve hardware,

procedural, or operational changes, and, therefore, does not affect

safety analyses.

(2) Does the proposed license amendment create the possibility

of a new or different kind of accident from any previously

evaluated?

Response:

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

different kind of accident from any previously evaluated. Extension

of the calibration intervals in question were evaluated and the

results documented in the ESFAS and Indication Instrument

Surveillance Test Extension reports. ESFAS and indicating instrument

drift analyses were performed to evaluate actual past and projected

future instrument drift. Revised safety system loop accuracy/

setpoint calculations and EOP setting calculations show that,

although some EOP setting changes will be made to accommodate

postulated drift associated with the extended calibration intervals,

safety analyses are not affected.

The changes to 1) specify LCO and surveillance requirements for

the wide range containment pressure instrumentation channels, 2) add

a quarterly functional test surveillance requirement to Item 4 of

[TS] Table 4.1-1 for the low Tavg actuation circuits of the

reactor coolant temperature channels, and 3) clarify that both

reactor trip and ESF actuation logic channels are functionally

tested constitute additional technical specification limitation and

controls. Additionally, these changes are consistent with W STS.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

The proposed changes do not involve significant reductions in

margins of safety. Loop accuracy/setpoint calculations show that

sufficient margin exists between the analytical and field trip

settings for the low Tavg, the SG low-low level, the high and

high-high containment pressure, the high differential steam line

pressure, the low steam line pressure, the high steam flow

(dependent upon turbine first stage pressure), the turbine trip low

auto stop oil pressure, and the 480V bus undervoltage trip functions

to accommodate postulated uncertainties associated with the extended

calibration intervals. And, although changes to EOP settings will be

made to accommodate the postulated uncertainties associated with the

extended calibration intervals for the wide range reactor coolant

loop temperature, the narrow and wide range SG level, and the steam

line pressure instrumentation, the EOP setting changes do not in any

way adversely affect the analytical limits established by safety

analyses.

Extension of the calibration intervals in question do not affect

safety analyses. The other changes being made in this application

involve additional technical specification limitations and controls

and are consistent with W STS. None of the changes involve

significant reductions in margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

New York, New York 10019.

NRC Project Director: Robert A. Capra

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: March 4, 1994

Description of amendment request: The proposed change would modify

Sections 5.3 and 5.6 of the Technical Specifications (TSs) to allow the

use of Westinghouse Vantage+ fuel with ZIRLO cladding. The present TSs

require the fuel rod cladding to be Zircaloy-4, which is used in the

Westinghouse Standard and Vantage 5H fuel designs.

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 to Technical Specifications 5.3.1 and 5.6.1

for Salem Generating Station (SGS) Unit Nos. 1 and 2:

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

consequences of an accident previously evaluated.

The fuel cladding design criteria for SGS would remain the same

for ZIRLO clad fuel as it is for Zircaloy-4 clad fuel. All fuel

design and performance criteria will continue to be met using NRC-

approved methods and no new single failure mechanisms will be

introduced. The use of ZIRLO clad fuel does not introduce any

changes to plant equipment or operation that would adversely affect

accident initiators or precursors. The proposed changes would not

result in any changes to compliance with licensing basis safety

limits.

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

accident from any accident previously evaluated.

The proposed changes would require that NRC[-]approved methods

be used in fuel assembly design. No new operating configurations

potentially resulting in the occurrence of a previously unanalyzed

event would be allowed by the proposed change.

3. do not involve a significant reduction in a margin of safety.

The proposed change would continue to require that NRC[-

]approved methods are used to ensure compliance with the fuel design

and safety limits which ensure that an acceptable margin of safety

is maintained relative to fuel assembly design.

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: Charles L. Miller

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: March 1, 1993

Description of amendment request: The proposed amendment would

clarify Technical Specification 3.6.1.2, Primary Containment Leakage,

and revise the ``as-found'' value of the overall integrated primary

containment leakage rate which is used when determining the test

schedule for future Type A tests within Surveillance Requirement

4.6.1.2.b. This amendment also requests an exemption from the

requirements of 10 CFR 50 Appendix J, Primary Reactor Containment

Leakage Testing for Water-Cooled Power Reactors.

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.

These proposed changes clarify Technical Specification 3.6.1.2

by providing a more definitive action to take if the leakage rate

limit(s) specified in the LCO are not being met. The current Action

is not clear on what actions are necessary if the leakage rate

limits (e.g., Type B and C limits) are known to be exceeded while

the reactor coolant system (RCS) temperature is above 200 deg. F,

which has caused compliance difficulties. The revised Action is

modeled after the one in the Primary Containment Integrity

Specification, which (through the definition of Primary Containment

Integrity) includes a provision that the containment leakage rates

be in compliance with the requirements of Specification 3.6.1.2.

Surveillance Requirement 4.6.1.2.b has been revised to reflect

the actual plant design basis leakage rate of La as the value

against which the ``as-found'' Type A test results are compared when

determining the test schedule for future Type A tests. The

probability of exceeding the maximum allowable leakage rate,

La, is not significantly increased since the ``as-left''

leakage rate requirement of 0.75 La (which must be met during

startup from any outage in which a Type A test has been performed)

is still imposed through LCO 3.6.1.2.a, Action 3.6.1.2.a and

Surveillance Requirement 4.6.1.2.a. The Applicability of

Specification 3.6.1.2 has been modified to resolve an existing

conflict with the current Action, which requires that a reactor

coolant system temperature of 200 deg. F not be exceeded with a

leakage rate greater than 0.75 La (during startups from outages

in which a Type A ILRT has been performed). With the modified

Applicability and the retained LCO requirement for the ``as-left''

leakage rate to be less than or equal to 0.75 La, the

requirement of the current Action (not to exceed to 200 deg. F) is

implicitly maintained, due to the provisions contained within

specification 3.0.4. This maintains the same margin for degradation

between performances of the periodic Type A tests as is provided in

the current specification. Since the analysis leakage limit of

La has not changed, the offsite radiological consequences of an

accident assumed in the safety analyses have not been affected.

The deletion of the current link between Specifications 3.6.1.2

and 3.10.1 is an administrative change only, made because the two

Specifications no longer overlap and the link is therefore

unnecessary.

In summary, there is no change in the probability or

consequences of any accident since the clarifications of the

existing LCO, Applicability, Actions, Surveillance Requirements and

the revised ``as-found'' acceptance criterion do not change the

design of the plant, nor the operational characteristics of any

plant system, nor the procedures by which the Operators run the

plant.

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

or different kind of accident from any accident previously

evaluated.

The proposed Action to address situations when the leakage rate

limit(s) cannot be met in Operational Conditions 1, 2 and 3, with

the reactor coolant system temperature greater than 200 deg. F, does

not create the possibility of a new or different kind of event - it

only provides the measures to be taken following determination of

increased containment leakage. The clarification to the existing

Applicability simply resolves an existing conflict between the

Applicability and the Action, and ensures that the same requirements

that were contained within the former Action are maintained

following implementation of the change, by preventing plant startup

above a RCS temperature of 200 deg. F (following an outage in which

a Type A test has been performed), unless the leakage rate is below

the 0.75 La test acceptance criterion. Additional changes are

being made to clarify the application of Appendix J requirements.

Revising the ``as-found'' value of La does not create the

possibility of a new or different kind of event - since the analysis

limit value, La, has not been increased and no new mode of

operation has been introduced.

In summary, the proposed changes do not create the possibility

of a new or different kind of accident, since no design changes are

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

and the method and manner of plant operation remains unchanged.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed Action simply imposes a more definitive action to

take when a leakage rate limit(s) is exceeded, consistent with the

Primary Containment Integrity Specification. The changes to the

Surveillance Requirements to reflect the ``as-found'' value of

La are consistent with the intent of the requirements specified

in Appendix J, and similar requirements have been provided for other

plants. The current requirement for ``as-left'' leakage rates to be

less than or equal to 0.75 La before increasing the reactor

coolant system temperature above 200 deg. F from outages in which a

Type A ILRT has been performed has been retained since the proposed

Action now includes a shutdown requirement, and in accordance with

Technical Specification 3.0.4, ``Entry into an OPERATIONAL CONDITION

or other specified condition shall not be made when the conditions

for the LCO are not met and the associated ACTION requires a

shutdown if they are not met within a specified time interval.''

Since the new Action includes a shutdown provision and the LCO

retains the current limit of 0.75 La, a change into the new

Applicability of Specification 3.6.1.2 cannot occur if 0.75 La

is exceeded. This ensures that the same margin as currently exists

today is maintained for possible degradation between performance of

the periodic Type A tests. The other changes are clarifications and

are administrative in nature. Therefore, the proposed changes do not

involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: John N. Hannon

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

Callaway County, Missouri

Date of amendment request: February 10, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification Table 2.2-1 and Bases Section 2.2.1. The

Functional Unit 14 of Table 2.2-1 would be revised to correct the Total

Allowance, reflecting the undervoltage relay span and to correct the

Allowable Value, reflecting the rack measurement and test equipment

(M&TE) uncertainty. The Bases would be revised to clarify the

relationship between the Trip Setpoint and Allowable Value, expressed

in voltage, and the Total Allowance, Z and S values, expressed in

percent of the undervoltage relay span (calibrated span of 70-100

volts).

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

consideration because operation of the Callaway Plant with these

changes would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Overall protection system performance will remain within the

bounds of the accident analyses documented in FSAR Chapter 15, WCAP-

10961-P, and WCAP-11883 since no hardware changes are proposed.

The RCP undervoltage reactor trip function is a primary trip

function and is credited in FSAR Section 15.3.2, Complete Loss of

Forced Reactor Coolant Flow. The trip setpoint is designed to ensure

plant operation within the DNB design basis. There will be no effect

on this analysis, or any other accident since the safety analysis

limit and trip response time are unaffected and remain the same as

discussed in FSAR Section 15.0.6 and FSAR Table 15.0.4.

The RCP undervoltage reactor trip will continue to function in a

manner consistent with the above analysis assumptions and the plant

design basis. As such, there will be no degradation in the

performance of nor an increase in the number of challenges to

equipment assumed to function during an accident situation.

These Technical Specification revisions do not involve any

hardware changes nor do they affect the probability of any event

initiators. There will be no change to normal plant operating

parameters, ESF actuation setpoints, accident mitigation

capabilities, accident analysis assumptions or inputs. Therefore,

these changes will not increase the probability of an accident

previously evaluated.

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

accident from any previously evaluated.

As discussed above, there are no hardware changes associated

with these Technical Specification revisions nor are there any

changes in the method by which any safety-related plant system

performs its safety function.

Changes to the Total Allowance and Allowable Value terms in

Technical Specification Table 2.2-1 will require only minor changes

to the acceptance criteria sections of a few surveillance

procedures. The normal manner of plant operation is unaffected. If

an undervoltage relay setpoint is found to be below the nominal trip

setpoint in Table 2.2-1, entry into Action Statements a or b of

Specification 2.2.1 will be affected insofar as the Allowable Value

is being lowered and the Total Allowance value contained in Equation

2.2-1 is being raised. However, the nominal trip setpoint is

unchanged and the required plant condition for exiting the Action

Statements, i.e. adjusting the trip setpoint consistent with the

Table 2.2-1 value, is likewise unchanged. The revisions to the Total

Allowance and Allowable Value correct errors in their derivation and

were calculated using the previously approved Westinghouse setpoint

methodology. The setpoint equations cited in that methodology are

unchanged; however, inputs to those equations have been revised to

reflect the undervoltage relay span and the rack M&TE uncertainty.

No new accident scenarios, transient precursors, failure

mechanism, or limiting single failures are introduced as a result of

these changes. There will be no adverse effect or challenges imposed

on any safety-related system as a result of these changes.

Therefore, the possibility of a new or different type of accident is

not created.

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

There will be no change to the DNBR Correlation Limit, the

design DNBR limits, or the safety analysis DNBR limits discussed in

Bases Section 2.1.1.

As discussed previously, the response time of the RCP

undervoltage reactor trip function will remain within the

assumptions used in the accident analyses. The analysis of the

complete loss of flow accident will remain as presented in FSAR

Section 15.3.2.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment

of protection functions. There will be no impact on DNBR limits,

FQ, F-delta-H, LOCA PCT, peak local power density, or any other

margin of safety. The safety analysis limit, 9384 Vac at the RCP

motor, and the nominal trip setpoint, 10,584 Vac, remain the same as

before.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: John N. Hannon

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

Callaway County, Missouri

Date of amendment request: February 17, 1994

Description of amendment request: The proposed amendment would

revise Technical Specifications 3/4.5.1 and Bases Section 3/4.5.1. A

new Action Statement a. would be added to Specification 3.5.1 to

provide a 72 hour allowed outage time (AOT) for one accumulator

inoperable due to its boron concentration not meeting the 2300-2500 ppm

band. The AOT for Action Statement b. would be changed to 24 hours in

lieu of the current AOT of 1 hour. Surveillances 4.5.1.1.a.1) and

4.5.1.1.b would be revised and Surveillance 4.5.1.2 would be deleted

per the guidance of NRC Generic Letter 93-05. Bases Section 3/4.5.1

would be revised to discuss the 72 hour and 24 hour AOTs for Action

Statements a. and b. above.

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 to the Technical Specifications do not

involve a significant hazards consideration because operation of

Callaway Plant in accordance with these changes would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Overall protection system performance will remain within the

bounds of the accident analyses documented in FSAR Chapter 15, WCAP-

10961-P, and WCAP-11883 since no hardware changes are proposed.

The safety injection (SI) accumulators are credited in FSAR

Section 15.6.5 for large and small break LOCA. There will be no

effect on these analyses, or any other accident analysis, since the

analysis assumptions are unaffected and remain the same as discussed

in FSAR Section 15.6.5. Design basis accidents are not assumed to

occur during allowed outage times covered by the Technical

Specifications. As such, the ECCS Evaluation Model equipment

availability assumptions made in FSAR Section 15.6.5 remain valid.

The SI accumulators will continue to function in a manner

consistent with the above analysis assumptions and the plant design

basis. As such, there will be no degradation in the performance of

nor an increase in the number of challenges to equipment assumed to

function during an accident situation.

These Technical Specification revisions do not involve any

hardware changes nor do they affect the probability of any event

initiators. There will be no change to normal plant operating

parameters, ESF actuation setpoints, accident mitigation

capabilities, accident analysis assumptions or inputs. The effect on

the Callaway core damage frequency has been quantified as

insignificant. Therefore, these changes will 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 evaluated.

As discussed above, there are no hardware changes associated

with these Technical Specification revisions nor are there any

changes in the method by which any safety-related plant system

performs its safety function. The normal manner of plant operation

is unaffected.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of these changes. There will be no adverse effect or challenges

imposed on any safety-related system as a result of these changes.

Therefore, the possibility of a new or different type of accident is

not created.

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

There will be no change to the DNBR Correlation Limit, the

design DNBR limits, or the safety analysis DNBR limits discussed in

Bases Section 2.1.1.

As discussed previously, the performance of the SI accumulators

will remain within the assumptions used in the large and small break

LOCA analyses, as presented in FSAR Section 15.6.5.

There will be no effect on the manner in which safety limits or

limiting safety system settings are determined nor will there be any

effect on those plant systems necessary to assure the accomplishment

of protection functions. There will be no impact on DNBR limits,

FQ, F-delta-H, LOCA PCT, peak local power density, or any other

margin of safety.

Based upon the preceding information, it has been determined

that the proposed changes to the Technical Specifications do not

involve a significant increase in the probability or consequences of

an accident previously evaluated, create the possibility of a new or

different kind of accident from any accident previously evaluated,

or involve a significant reduction in a margin of safety. Therefore,

it is concluded that the proposed changes meet the requirements of

10CFR50.92(c) and do not involve a significant hazards

consideration.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: John N. Hannon

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: March 1, 1994

Description of amendment request: The proposed change would revise

the Technical Specifications (TS) for the North Anna Power Station,

Units No. 1 and No. 2 (NA-1&2). Specifically, the change would

eliminate certain surveillance requirements for the emergency diesel

generators which have been determined to be unnecessary.

The NRC has completed a comprehensive examination of surveillance

requirements in 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 TS. Based on the results of the evaluations

documented in NUREG-1366, the NRC issued Generic Letter 93-05, ``Line-

Item Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' dated September 27,

1993.

The safety function of the Emergency Diesel Generators (EDGs) is to

supply AC electrical power to plant safety systems whenever the

preferred AC power supply is unavailable. Consistent with Generic

Letter 93-05, Item 10.1 and NUREG-1366, the licensee is requesting a

change to the testing requirements of an operable EDG when the

alternate safety buses' EDG is inoperable or an offsite circuit is

inoperable, the separation of the hot restart test of an EDG from the

24 hour loaded run, and the elimination of fast loading of EDGs except

for the 18 month surveillance test of the Loss of Offsite Power (LOOP)

capability.

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 of

occurrence or consequences of an accident previously evaluated.

Modifying the operability testing requirements for an inoperable

EDG or inoperable offsite AC source(s), gradual loading of EDGs

during surveillance testing, and separating the hot restart test of

an EDG from the 24 hour load run test of EDGs does not affect the

probability of occurrence or consequences of any previously

evaluated accidents. Surveillance testing of the EDG in accordance

with Revision 2 of Regulatory Guide 1.9 (December 1979) will

continue to ensure that the EDGs will be capable of performing their

intended safety functions. Therefore, modifying the operability

testing requirements for an inoperable EDG or inoperable offsite AC

source(s), gradual loading of EDGs during surveillance testing, and

separating the hot restart test of an EDG from the 24 hour load run

test of EDGs does 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.

Modifying the operability testing requirements for an inoperable

EDG or inoperable offsite AC source(s), gradual loading of EDGs

during surveillance testing, and separating the hot restart test of

an EDG from the 24 hours load run test of EDGs does not involve any

physical modifications of the plant or result in a change in a

method of operation. Surveillance testing of the EDG in accordance

with Revision 2 of Regulatory Guide 1.9 (December 1979) will

continue to ensure that the EDGs will be capable of performing their

intended safety functions. Therefore, a new or different type of

accident is not made possible.

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

Modifying the operability testing requirements for an inoperable

EDG or inoperable offsite AC source(s), gradual loading of EDGs

during surveillance testing, and separating the hot restart test of

an EDG from the 24 hour load run test of EDGs does not affect any

safety limits or limiting safety systems settings. System operating

parameters are unaffected. The availability of equipment required to

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

Surveillance testing of the EDG in accordance with Revision 2 of

Regulatory Guide 1.9 (December 1979) will continue to assure that

the EDGs will be capable of performing their intended safety

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: Herbert N. Berkow

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

Project No. 2, Benton County, Washington

Date of amendment request: May 10, 1993

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to incorporate new power to flow

limits based on core power stability calculations performed for Cycle

9. In addition, the proposed amendment would clarify the maximum

measured decay ration permitted during 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. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's evaluation of

the licensee's analysis is presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change related to the instability regions on the power

to flow map is based on new calculations using a new code (STAIF) while

maintaining a decay ration of 0.9 or less as required in IEB 88-07. The

closer the operators come to a decay ration of 1.0, the closer the core

comes to potential core power instabilities. By ensuring that the decay

ratio is maintained below 0.9, the operators reduce the likelihood of

core power instabilities. The result of the revised calculations using

this new code is that the restricted regions are expanded over the

regions contained in the current TS. This increase in restricted

regions results in plant operation further from potential core power

instabilities compared to the restricted regions in the current TS,

resulting in a decreased probability of core power oscillations. The

power to flow map regions are operating restrictions that, for the core

power oscillation restricted regions, are intended to reduce the

likelihood of the onset of oscillations. The core power oscillation

restricted regions on the power to flow map do not contribute to any

mitigative actions or plant response after a power oscillation occurs,

thus the proposed change does not change the consequences of any

accidents previously evaluated.

The proposed amendment would also change the wording of the

technical specifications to clarify that action must be taken to reduce

the measured decay ration if any two neutron signals of ``greater than

or equal to 0.75,'' as opposed to the current wording ``greater than

.75,'' are measured. This would not have any measurable effect on the

implementation of the affected TS, and would, if anything, result in

action being taken at a maximum lower value than the current TS. This

proposed amendment would not, therefore, involve a change in the

probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change; modifies existing restrictions on the power to

flow map, and does not involve any modifications to plant systems or

components or the manner in which they are operated.

Changing the wording of the TS to require that action be taken to

reduce the measured decay ration if any two neutron signals of

``greater than or equal to 0.75,'' as opposed to the current wording

``greater than .75,'' are measured, does not involve any modifications

to plant systems or components or the manner in which they are

operated.

Based on these considerations, this does not create or increase the

possibility of a new or different kind of accident.

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

of safety?

The margin of safety related to the proposed TS change is the core

power vs. core flow restrictions on the power to flow map. These

restrictions are currently based on maintaining a decay ratio less than

0.9, which provides a margin of at least a decay ration of 0.1 from

what is defined as a decay ration (1.0) that would result in an

unstable core. Since the revised curves are based on ensuring decay

rations of less than 0.9 are maintained, the existing margin of safety

is maintained.

Changing the wording of the TS to require that action be taken to

reduce the measured decay ratio if any two neutron signals of ``greater

than or equal to 0.75,'' as opposed to the current wording ``greater

than .75,'' are measured, would not have any measurable effect on the

implementation of the affected TS, and would, if anything, result in

action being taken at a maximum lower value than the current TS. This

would not have any significant impact on how close the plant was

allowed to operate to potential core power instability, and would not,

therefore, have a significant effect ont he margin of safety related to

the proposed TS.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &

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

NRC Project Director: Theodore R. Quay

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

Project No. 2, Benton County, Washington

Date of amendment request: July 29, 1993, with supplemental

information provided March 11, 1994 and March 17, 1994

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to reflect a new refueling platform.

Specifically, the amendment would add new values for protective

features in the TS to reflect the new refueling platform. Values for

the old refueling platform are retained in the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The staff's evaluation of the licensee's analysis is

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The only accident evaluation affected by the proposed changes are

those associated with the Fuel Handling Accident (FHA) analyses

presented in WNP-2 Final Safety Analysis Report (FSAR) section 15.7.4.

As discussed therein, the fuel handling accident event that produces

the largest number of failed spent fuel rods is the drop of a spent

fuel bundle into the reactor core when the reactor vessel head is off.

The probability of dropping a spent fuel assembly onto other fuel

assemblies in the reactor vessel does not increase with the new design.

The NF500 mast functions identically to the old mast when grappling,

lifting, or moving a fuel assembly. It does not degrade platform design

features such as grapple fail-safe on loss of air, dual lifting cables,

backup cable reel brake, and the grapple engaged loaded interlock, all

of which serve to protect against a fuel drop event. The new mast is

more rigid than the previous mast design and, therefore, is less prone

to mast bowing. The consequences of dropping a fuel assembly are also

unaffected because the weight of the mast is not considered in existing

FHA analysis. The number of postulated fuel pins which fail as a result

of the FHA is unaffected since the energy imparted by the dropped

assembly is independent of the mast design, and mitigating systems will

function as previously analyzed. Further, analysis by GE of a

postulated accident in which the exposed portion of the NF500 mast is

struck by a missile and severed while lifting a fuel bundle with both

falling onto the top of the core has been conducted, showing that the

consequences of the increased weight of the mast and bundle are bounded

by the current WNP-2 FSAR analysis for the fuel bundle only FHA.

Retaining the ability to use the old mast does not introduce any

changes to the current TS that reflect the analysis of the old mast.

The proposed change would not, therefore, significantly increase the

probability or consequences of a previously analyzed accident.

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

kind of accident from any accident previously evaluated?

No new failure modes are introduced as a result of the proposed

changes. The NF500 mast in intended as an exact replacement for the

currently installed mast, and is designed to match or exceed the

strength and performance of the NF400 mast in all areas. No new fuel

handling methods or surveillance procedures will be necessary as a

result of installation of the new mast. The proposed change does not

affect the manner in which protective interlocks operate. Limits on

fuel travel in all directions are unchanged. Retaining the ability to

use the NF400 mast presents no new accident possibilities since no

changes in fuel mast operation would result from use of the existing

mast. The proposed change would not, therefore, create the possibility

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

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

of safety?

The changed refueling mast cutoff and interlock values account for

the increase weight of the mast, or a portion thereof, and do not

affect the margins related to the fuel bundle drop analyses. The new

mast has the same single failure protection as the old mast. The

proposed change would not, therefore, involve a significant reduction

in a margin of safety.

The NRC staff has determined that 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: Nicholas S. Reynolds, Esq., Winston &

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

NRC Project Director: Theodore R. Quay

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

Project No. 2, Benton County, Washington

Date of amendment request: December 20, 1993

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to address new containment purge and

vent valves to be installed in the 1994 refueling outage. The TS are

being modified to remove the requirement to ensure the remaining

existing-design valves' position remains at less than or equal to

70 deg. because the valves have a permanently installed mechanical stop

to limit the open position to ensure adequate closure times. In

addition, this modification is being requested because the current TS

are too limiting for the new valves, which are designed to close from a

90 deg. open position. The TS are also being modified to change the

containment leak testing requirements for the new valves from 6 months

to 2 years, to reflect the improved seat design of the replacement

valves. Additional administrative changes are proposed to delete an

out-of-date note, and to relocate an action statement requirement from

surveillance section of the TS to the action statements section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

Regarding the removal of the requirement to ensure the remaining

existing-design valves' position remains at less than or equal to

70 deg.: The maximum open position of the containment purge and vent

valves is not one of the initiating events for any previously evaluated

accident in the WNP-2 FSAR. Thus the proposed change will not affect

the probability of an accident previously evaluated. The containment

purge and vent valves' position is considered in the accident analyses,

and could affect the analyzed consequences of events. The current

limiting condition for operation (LCO) and Action Statement requires,

and the surveillance verifies, that the permanently installed 70 deg.

block is in place and effective. If the existing valves were open

further than 70 deg., the valves may not close in time. The valves have

a welded mechanical stop installed that limits the position to no more

than 70 deg. open, which is a fixed condition that can only be changed

by plant modification requiring evaluation against the requirements of

10 CFR 50.59. The licensee considers the mechanical stop as sufficient

to ensure the existing valves will remain within existing analysis

bounds for a design basis loss of coolant accident (LOCA). In addition,

the new valves are qualified to close within the 5 seconds assumed in

the design basis LOCA. The licensee considers, therefore, that the

existing and new valves will operate as required for accident

mitigation with the proposed change, and that the proposed change will

not affect the consequences of accidents previously evaluated.

Regarding the modification of the containment leak testing

requirements to reflect the new design valves: The containment purge

and vent valves are not one of the initiating events for any previously

evaluated accident in the WNP-2 FSAR. Thus the proposed change will not

affect the probability of an accident previously evaluated. The metal

to metal seat valves meet the Appendix J criteria necessary to be

tested as type C valves. Type C valves can be tested every 2 years,

compared to every 6 months for the current valves. The testing

frequency is based on the performance of the valve types to ensure that

they are capable of maintaining the necessary leak tightness over the

test interval. The new valves' design has been certified to provide the

same leak tightness over 2 years that the current valves provide over 6

months, thus the consequences of analyzed events remains unaffected by

the proposed change.

Regarding the administrative changes: The proposed change would (1)

delete a note that was applicable only through April 10, 1988, and (2)

move an action that is currently stated in the SURVEILLANCE

REQUIREMENTS section of a TS to the ACTIONS section of the same TS.

These changes do not affect the design or operation of the plant or the

implementation of the affected TS, and as such would not affect the

probability or consequences of previously analyzed events.

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

kind of accident from any accident previously evaluated?

Regarding the removal of the requirement to ensure the remaining

existing-design valves' position remains at less than or equal to

70 deg.: No aspect of the design or plant operation is affected by

deletion of the surveillance or removal of the reference to the block

from the LCO and Action Statement, no new modes of plant operation are

introduced, and the proposed change does not require physical

modification of the plant. The valves not being replaced will continue

to be limited from opening greater than 70 deg. by the welded and non-

adjustable blocking feature. The capability of these valves to close

within 5 seconds to meet the limiting design basis accident (LOCA) will

remain unchanged. The replacement valves will be capable of closing

within the same 5 seconds from a full-open position of 90 deg.. Since

the proposed change does not introduce any new component, system, or

plant operating conditions, the change does not create the possibility

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

Regarding the modification of the containment leak testing

requirements to reflect the new design valves: The proposed change in

surveillance frequency for the replacement valves does not introduce

any new mode of plant operation, nor does it involve plant

modifications. The new valves operate in the same manner as the old

valves, only the seating surfaces are different. This does not affect

the way the valves operate to perform their function. The proposed

change would not, therefore, involve any new or different kinds of

accidents from any previously evaluated.

Regarding the administrative changes: The proposed changes do not

introduce any new modes of plant or equipment operation, nor do they

involve physical modification of the plant. The proposed change would

not, therefore, create the possibility of a new or different kind of

accident from any accident previously evaluated.

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

of safety?

Regarding the removal of the requirement to ensure the remaining

existing-design valves' position remains at less than or equal to

70 deg.: The margin of safety of concern with the proposed change is

the need for the containment purge and vent valves to close in 5

seconds, which will ensure that part 100 limits for design basis events

are not exceeded. The proposed change does not affect the maximum open

position of the existing valves, thus the valves will still close

within 5 seconds. In addition, the new valves, with their maximum full

open position of 90 deg., are a new design that will still close within

the five seconds from the full open position, thereby preserving the

existing margin of safety.

Regarding the modification of the containment leak testing

requirements to reflect the new design valves: The margin of safety

involved in the proposed TS change is the amount of leakage that may

occur due to plant degradation that may affect the design basis

accident assumptions for leakage. The new design valves have been

certified to provide the same leak tightness over 2 years that the

current valves provide over 6 months, thus the leakage assumptions for

design basis events is unaffected. The proposed change would not,

therefore, affect the margin of safety provided by the TS.

Regarding the administrative changes: There are no margins of

safety affected by the administrative changes.

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, N.W., Washington, D.C. 20005-3502

NRC Project Director: Theodore R. Quay

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

Project No. 2, Benton County, Washington

Date of amendment request: January 6, 1994

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to remove the requirements for the

Seismic Monitoring Instrumentation from the TS and relocate them to the

FSAR and plant procedures. The requirements described in the

specifications will be maintained in the FSAR and plant procedures.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The staff's evaluation of the licensee's analysis is

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The seismic monitors only provide monitoring and recording of

seismic events that might occur in the vicinity of WNP-2. The

instrumentation are not relied upon in current accident analyses for

any automatic or manual initiation of safety systems in response to a

seismic event. The proposed change would not, therefore, significantly

increase the probability or consequences of a previously analyzed

accident.

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

kind of accident from any accident previously evaluated?

The proposed change does not affect the manner in which the plant

is operated, maintained, or tested. The proposed change would not,

therefore, create the possibility of a new or different kind of

accident from any previously analyzed.

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

of safety?

The seismic monitors provide monitoring and recording functions

only, and are not relied upon in accident analyses for automatic or

manual initiation of any safety system. Thus the results of analyzed

events, and the associated margins of safety, are unaffected by the

administrative removal of the seismic monitors from the TS. The

proposed change would not, therefore, involve a significant reduction

in a margin of safety.

The NRC staff has determined that 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: Nicholas S. Reynolds, Esq., Winston &

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

NRC Project Director: Theodore R. Quay

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

Project No. 2, Benton County, Washington

Date of amendment request: February 17, 1994

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to support hydrostatic testing of the

reactor coolant system. Specifically, the proposed amendment would: (1)

add a Special Test Exception that would allow Mode 4 (Cold Shutdown)

operation up to 212 deg.F, compared to the current limit of 200 deg.F,

without shutdown cooling in operation, to conduct hydrostatic testing,

and (2) add a new reactor metal temperature vs reactor vessel pressure

(P/T) limit curve that is applicable up to 8 effective full power years

(EFPY), for use during hydrostatic testing and non-nuclear plant

heatup.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

Regarding the proposed Special Test Exception: The proposed change

would allow performance of hydrostatic testing in OPERATIONAL CONDITION

4 at temperatures greater than 200 deg.F but less than or equal to

212 deg.F. Operating in this condition is only allowed if specified

OPERATIONAL CONDITION 3 secondary containment requirements are met. The

operating condition is not considered as an initiator for any event

analyzed in the FSAR, therefore the proposed change would not affect

the probability of an accident previously evaluated.

The specified OPERATIONAL CONDITION 3 requirements compensate for

the allowed temperature increase and assure that the consequences of a

potential leak will be conservatively bounded by the existing FSAR

accident analyses, as discussed below.

The hydrostatic test is conducted near water solid, all rods in,

and temperature less than or equal to 212 deg.F. The stored energy in

the core will be very low (approximately 43 days of shutdown conditions

and partial core replacement during refueling) and the potential for

failed fuel and a subsequent increase in coolant activity above

Technical Specification limits is minimal. In addition, secondary

containment will be OPERABLE and capable of handling airbone

radioactivity from leaks that could occur during the performance of the

testing. Maintaining the temperature less than or equal to 212 deg.F

will ensure that any leak will not flash to steam, thereby ensuring the

potential for airborne activity remains low. Requiring the standby gas

treatment system (SGTS) to be OPERABLE will conservatively ensure that

any airborne radiation from leaks will be processed by the SGTS thereby

limiting releases to the environment. Existing pipe breaks analyzed in

Chapter 15 of the FSAR are bounding for the proposed condition. In the

event of a large break, the reactor would rapidly depressurize,

allowing the low pressure ECCS subsystems to operate. The capability of

the subsystems required for OPERATIONAL CONDITION 4 would be adequate

to keep the core flooded under this condition. Small system leaks would

be detected by leakage inspections before significant inventory loss

occurred. Thus the consequences of previously analyzed accidents are

not increased by the proposed amendment.

Regarding the proposed P/T limit curve: The proposed change would

modify the P/T limit curves that are based on prevention of brittle

fracture of the reactor vessel. The proposed change would result in

plant operation closer to the actual brittle fracture condition of the

reactor vessel, potentially making a brittle fracture more likely. This

condition is offset by the slow heatup conducted using only pump heat,

which would result in lower stresses in the reactor vessel than are

assumed in the brittle fracture analyses. The resulting P/T limit curve

based on 8 EFPY would have sufficient conservatism from the actual

vessel brittle fracture condition to make vessel failure as unlikely as

the original 32 EFPY curve.

The potential reactor vessel failure mechanisms are not affected by

the proposed change, therefore the consequences of previously analyzed

accidents are unaffected by the proposed change.

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

kind of accident from any accident previously evaluated?

Regarding both the proposed Special Test Exception and the proposed

P/T limit curve: The proposed change introduces no new failure modes,

involves no physical modification to the plant or change in system

configurations, nor does it involve changes in plant, system, or

component operation. The proposed change, therefore, does not create

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

accident previously evaluated.

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

of safety?

Regarding the proposed Special Test Exception: The hydrostatic test

is conducted with low stored energy in the reactor, which is bounded by

the assumed decay heat in current safety analyses. In the unlikely

event that a leak from the reactor coolant system were to occur, the

RPV would depressurize and the low pressure systems would be available

to keep the core flooded. This would ensure that the fuel peak clad

temperature would not exceed 2200 deg.F, which is the design basis that

provides the margin of safety for the reactor itself. In addition,

secondary containment will be maintained during the hydrostatic test,

which would ensure that any potential airborne activity that might

occur would be filtered through the SGTS. This would ensure that the

current margins to the 10 CFR Part 100 limits remain bounded by current

analyses. The proposed change would, therefore, not involve a

significant reduction in the margins of safety.

Regarding the proposed P/T limit curve: The proposed new curves

would allow plant operation closer to the actual brittle fracture

condition of the reactor vessel during hydrostatic test conditions

only. This would result in a reduced margin in the protection afforded

by the P/T curve. The new curves would, however, allow a lower

temperature for conduct of the hydrostatic test, which would increase

the heat sink available in the RCS, and increase the margin to decay

heat loads assumed in accident analyses. This would result in reduced

potential for extensive flow from any break, reduce the time for

initiation of low pressure ECCS systems, and reduce the available

radioactive decay products that are available for release during any

postulated accident condition. The overall impact of the conditions

increases the margin to 10 CFR Part 100 limits that are the design

margin of safety for postulated loss of coolant accidents. The overall

effect of the proposed change would not involve a significant reduction

in the overall margins of safety.

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, N.W., Washington, D.C. 20005-3502

NRC Project Director: Theodore R. Quay

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: February 23, 1994

Description of amendment request: The proposed amendment would

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

6.8.c by removing the requirement to conduct a biennial review of plant

procedures in accordance with American National Standards Institute

(ANSI) N18.7-1976. The licensee proposes using alternate programs, that

are already in place, to ensure that procedures are periodically

reviewed and maintained current. A biennial review of the Integrated

Plant Emergency Operating Procedures (IPEOPs), however, would continue.

The requirements for these alternate programs and for the IPEOP review

would be added to the Operational Quality Assurance Program Description

(OQAPD).

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 were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by eliminating the periodic reviews of routine

administrative and technical procedures. Sufficient controls are

established to ensure that procedures impacting safety-related

structures, systems, and components are maintained current,

accurate, and usable. This TS change will therefore not impact the

function or method of operation of plant equipment. Thus, a

significant increase in the probability of a previously analyzed

accident does not result due to this change. No systems, equipment,

or components are affected by the proposed changes. Thus, the

consequences of a malfunction of equipment important to safety

previously evaluated in the Updated Safety Analysis Report (USAR)

are not increased by this change. The proposed changes do not affect

equipment or its operation, and, thus, do not affect the

probabilities or consequences of an accident. Therefore, WPSC

concludes that this change does not significantly increase the

probability or consequences of an accident.

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

from any accident previously evaluated.

The proposed changes do not involve changes to the physical

plant or operations. Since periodic procedure reviews do not

contribute to accident initiation, a change related to such an

activity does not produce a new accident scenario or produce a new

type of equipment malfunction. Also, this change does not alter any

existing accident scenarios. The proposed changes do not affect

equipment or its operation, and thus, do not increase the

possibility of a new or different kind of accident.

3) involve a significant reduction in the margin of safety.

The proposed changes do not affect equipment or its operation,

and thus, do not involve any reduction in the margin of safety.

Therefore, use of the proposed Technical Specification would not

involve any reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

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

54301.

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

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

NRC Project Director: John N. Hannon

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.

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 amendment: February 17, 1994

Brief description of amendment request: The proposed amendments

would revise the combined Technical Specifications (TS) for the Diablo

Canyon Power Plant Unit Nos. 1 and 2 to revise TS 3/4.3.2, ``Engineered

Safety Feature Actuation System Instrumentation,'' as follows: (1)

Table 3.3-3, functional unit 6.c.2), channels to trip, would be changed

from 2/steam generator in one steam generator to 2/steam generator in

any 2 steam generators to correct an administrative error. (2) Table

3.3-4 would be changed as follows: a. functional unit 4.6., Negative

Steam Pressure Rate - High, trip setpoint and allowable value, would be

changed from -100 psi/sec and -105.4 psi/sec to 100 psi and 105.4 psi,

respectively; b. a note would be added stating that the time constants

utilized in the rate-lag controller for Negative Steam Pressure Rate -

High, are equal to 50 seconds.

Date of individual notice in Federal Register: March 1, 1994 (59 FR

9789)

Expiration date of individual notice: March 31, 1994

Local Public Document Room location: California Polytechnical State

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

Department, San Luis Obispo, California 93407.

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 4, 1994

Brief description of amendment request: The proposed amendment

would add a new Section 3/4.10.8, ``Inservice Leak and Hydrostatic

Testing,'' and the Bases. The new section would allow Hope Creek to

remain in OPERATIONAL CONDITION 4 with reactor coolant temperatures up

to 212 *F to facilitate inservice leak and hydrostatic testing.

Date of publication of individual notice in Federal Register: March

16, 1994 (59 FR 12384)

Expiration date of individual notice: April 15, 1994

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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: April 28, August 12, and November 17,

1993, and February 2, 1994

Brief description of amendment request: The proposed changes

increase the spent fuel pool capacities for Salem 1 and 2 from the

current 1170 fuel assemblies to 1632 fuel assemblies. Also, the decay

time for refueling operations is being extended from 100 hours to 168

hours.

Date of publication of individual notice in Federal Register: March

4, 1994 (59 FR 10440)

Expiration date of individual notice: April 4, 1994

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

West Broadway, Salem, New Jersey 08079.

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 20555, and at the local public document

rooms for the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert

Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of application for amendment: September 3, 1993, as

supplemented February 1, 1994

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

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