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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from February 16, 1995, through March 3, 1995.

The last biweekly notice was published on March 1, 1995.

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

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

intervene is discussed below.

By April 14, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above. [[Page 14016]]

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

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

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

Commission by a toll-free telephone call to Western Union at 1-(800)

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

should be given Datagram Identification Number N1023 and the following

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

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

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: January 31, 1995

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TSs) for Calvert Cliffs, Unit Nos.

1 and 2, to increase the amount of Trisodium Phosphate Dodecahydrate

(TSP) located in the containment sump baskets required to be verified

by TS surveillance. The requested change is the result of an reanalysis

of the amount of TSP necessary to maintain the appropriate pH in the

containment sump water subsequent to a Loss of Coolant Accident.

Specifically, the request would change the TS value of TS 4.5.2.e.3

from the existing amount of 100 ft3 to 289 ft3. TS 4.5.2.e.4

would also be changed by moving the amounts of TSP and refueling water

storage tank water to be used in the required tests to the TS Bases

Section 3/4.5.2 and 3/4.5.3. These Bases sections would also be changed

by modifying the test methods.

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. Would not involve a significant increase in the probability

orconsequences of an accident previously evaluated.

Trisodium Phosphate Dodecahydrate (TSP) is stored in the

containment lower level to raise the pH of the sump and spray water

following a Loss of Coolant Accident (LOCA). As the pH of the water

increases, more radioactive iodine is kept in solution and the

possibility of airborne radioactivity leakage is decreased. An

additional advantage of a higher pH is the beneficial reduction in

chloride stress corrosion cracking of metal components in the

containment following an accident.

This chemical is an accident mitigator, not an accident

initiator in that it is not used until after an accident has

occurred. At the time it goes into solution, the accident has

occurred, containment spray has been activated and water has

collected in the containment sump. Therefore, increasing the

Technical Specification minimum amount verified to be in each

containment will not involve a significant increase in the

probability of an accident previously evaluated.

Updated Final Safety Analysis Report, Chapter 14.24, ``Maximum

Hypothetical Accident'', uses an assumption of a pre-RAS minimum

containment spray pH of 5.0 for the iodine removal calculation and a

post-RAS sump pH of 7.0 for iodine retention. Raising the pH to 7.0

does not increase the consequences of an accident previously

evaluated.

The proposed change to Technical Specification 4.5.2.e.4 would

remove the amounts of chemical and water used in the test to the

Bases. This relocation will not alter the test method or acceptance

criteria, but will allow adjustments to the ratio of TSP and borated

water under the controls of 10 CFR 50.59 to reflect changes in plant

conditions. In the Bases, the amount of TSP used in the test is

changed to reflect the ratio of TSP to water that would be found in

the containment following a LOCA. The specified concentration of

boron in the test reflects the highest concentration that could be

found in the containment following a LOCA. The test temperature is

changed to 120 deg.F which is well below the temperature expected to

be found in the containment sump following a LOCA. The decanting of

the solution does not change the intent of the test method since the

dissolving period will still be conducted without agitation.

Therefore, this change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated. [[Page 14017]]

The addition of more TSP does not represent a significant change

in the configuration or operation of the plant. Trisodium Phosphate

Dodecahydrate is currently present in the containment lower level.

There are no physical changes which result from the increase in

volume. The proposed change to Technical Specification 4.5.2.e.4

would move the amounts of chemical and water used in the test to the

Bases. This relocation will not alter the test method or acceptance

criteria, but will allow adjustments to the ratio of TSP and borated

water under the controls of 10 CFR 50.59 to reflect changes in plant

conditions. In the Bases, the amount of TSP used in the test is

changed to reflect the ratio of TSP to water that would be found in

the containment following a LOCA. The specified concentration of

boron in the test reflects the highest concentration that could be

found in the containment following a LOCA. The test temperature is

changed to 120 deg.F which is well below the temperature expected to

be found in the containment sump following a LOCA. The decanting of

the solution does not change the intent of the test method since the

dissolving period will still be conducted without agitation.

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

or different type of accident from any accident previously

evaluated.

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

safety.

Trisodium Phosphate Dodecahydrate is stored in the containment

lower level to raise the pH of the sump and spray water following a

LOCA. As the pH of the water increases, more radioactive iodine is

kept in solution and the possibility of airborne radioactivity

leakage is decreased. Additionally, a higher pH has a beneficial

effect on chloride stress corrosion cracking of metal components in

the containment.

Technical Specification 4.5.2.e.3 requires verification that a

minimum volume of TSP is contained in the storage baskets in each

containment. This change proposes to increase that volume. The

increased volume will ensure the containment sump, when filled with

water, will have an acceptable pH following a LOCA.

The proposed change to Technical Specification 4.5.2.e.4 would

move the amounts of chemical and water used in the test to the

Bases. This relocation will not alter the test method or acceptance

criteria, but will allow adjustments to the ratio of TSP and borated

water under the controls of 10 CFR 50.59 to reflect changes in plant

conditions. In the Bases, the amount of TSP used in the test is

changed to reflect the ratio of TSP to water that would be found in

the containment following a LOCA. The specified concentration of

boron in the test reflects the highest concentration that could be

found in the containment following a LOCA. The test temperature is

changed to 120 deg.F which is well below the temperature expected to

be found in the containment sump following a LOCA. The decanting of

the solution does not change the intent of the test method since the

dissolving period will still be conducted without agitation.

Therefore, this change 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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Ledyard B. Marsh

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

Station, Plymouth County, Massachusetts

Date of amendment request: February 9, 1995

Description of amendment request: The proposed amendment would

increase the Reactor High Water Level Trip Level Setting for the Group

1 isolation. The change will allow an increase to the main steam

isolation valve (MSIV) high water level isolation setpoint.

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:

In accordance with 10 CFR 50.91, Boston Edison submits the

following analysis addressing the no significant hazards

consideration. The proposed changes do not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Operation of the station in accordance with the proposed Trip

Level Setting will not significantly increase the probability or

consequences of an accident previously evaluated. The MSIV high

water level isolation signal is provided to protect against rapid

depressurization due to a pressure regulator malfunction during

plant startup. The high water level isolation signal is not

functional when the mode switch is in the RUN position. A high water

level in the reactor vessel indicates that fuel is covered.

Increasing the Trip Level Setting will have minimal effect on

moisture carryover in the event of a pressure regulator failure at

low reactor power. MSIV closure (Group 1) is initiated by low

reactor pressure (810 psig) approximately 30 seconds into the event.

The resulting reactor water level swell is not sufficient to reach

the bottom elevation of the main steam lines.

The proposed Technical Specification allowable value for the

Reactor Low Level Trip Level Setting and the Reactor Low Low Water

Level Trip Level setting does not involve significant increase in

the probability or consequence of an accident.

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

accident from any previously analyzed.

The proposed change does not affect the Group 1 isolation safety

function. The change does not involve any plant hardware changes

that could introduce any new failure modes or effects; thus, the

change can not create the possibility of a new or different kind of

accident from any previously analyzed.

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

The proposed change does not affect the Group 1 isolation safety

function. The proposed change is consistent with the FSAR [Final

Safety Analysis Report] and Technical Specification basis associated

with reactor vessel inventory control and main steam line flooding.

The proposed change to the instrument calibration range does not

affect the margin of safety for systems or components affected by

the change. Operating Pilgrim in accordance with the proposed Trip

Level Setting 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

NRC Project Director: Walter R. Butler

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 6, 1995

Description of amendment request: The change proposes to relocate

the cycle specific core operating limits of Figure 3.1-1, Shutdown

Margin Versus Boron Concentration, from Technical Specification (TS)

3.1.1.2, Shutdown Margins - Modes 3, 4, and 5, to the Core Operating

Limits Report (COLR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change of relocating TS Figure 3.1-1, Shutdown

Margin Versus Boron Concentration to the COLR has no influence

[[Page 14018]] or impact to the probability or consequences of an

accident. The revised TS will continue to implement the shutdown

margin limits through reference to the Shutdown Margin Curve in the

COLR. In addition, the COLR is subject to the existing controls of

TS 6.9.1.6. Given that this change is an administrative relocation

of the Shutdown Margin Curve to another TS controlled document,

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.

No safety-related equipment, safety function, or plant operation

will be altered as a result of this proposed change. The TS will

continue to require operation within the required core operating

limits. 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.

Relocation of the Shutdown Margin Curve to the TS controlled

COLR has no effect on the core operating limits currently in force

in TS 3.1.1.2. Future revisions to the Shutdown Margin Curve are

governed by TS 6.9.1.6 which stipulates the specific TS that

reference the COLR limits and the methodologies utilized in

developing those limits. Given that the change is an administrative

relocation of the Shutdown Margin Curve to another TS controlled

document, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

NRC Project Director: William H. Bateman

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

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

Date of amendment request: January 12, 1995

Description of amendment request: The proposed amendments would

revise and clarify portions of Technical Specification (TS) Section

6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee

nuclear stations. The licensee submitted a combined amendment request

covering the three Duke Power nuclear stations. The proposed changes

are described below.

1. Remove the specific assignment of responsibilities for the

review, distribution, and approval activities contained in the

Technical Review and Control Section of each station's TS. The proposed

specifications state that these activities will be performed by a

knowledgeable individual/organization. Approval of the affected

documents is to be at the appropriate manager/superintendent level as

specified in Duke administrative controls.

2. Move the requirement for the review of proposed changes in the

stations' TS and Operating Licenses by the Duke Nuclear Safety Review

Board (NSRB) to Duke administrative procedures (Selected Licensee

Commitments documents) and change the wording of the requirements

covering NSRB meeting frequency. The Oconee TS covering the NSRB are

being rewritten to be consistent with McGuire and Catawba.

3. Add Technical Review and Control Program implementation and

Plant Operations Review Committee (PORC) implementation to the list of

required procedures and programs for each nuclear station.

4. Change or clarify certain TS administrative requirements

covering technical review and control activities or records retention

requirements.

5. For Oconee only, under ``Station Operating Procedures,'' revise

the TS requirements covering the review and approval of station

procedures and temporary procedure changes such that these are now

consistent with the corresponding requirements for McGuire and Catawba.

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:

(It should be noted that the licensee submitted a combined

analysis that covers McGuire, Catawba, and Oconee nuclear stations.)

Standard 1. The proposed amendments will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The provisions of these proposed amendments concern

administrative changes in the stations' Technical Specifications

involving the Technical Review and Control, Procedures and Programs/

Station Operating Procedures, and Records Retention/Station

Operating Records portions of the Administrative Controls Section.

The requested changes primarily affect review and control

activities, but also include other administrative changes affecting

the approval of station procedures (Oconee only), records retention,

and definition of the term ODCM [offsite dose calculation manual]

(McGuire and [Catawba]). The provisions of the proposed amendment

primarily involve the relocation of existing Technical

Specifications review, distribution, or approval requirements to

internal Duke administrative controls. However, implementation of

the proposed amendment does involve changes to several review/

distribution activities. Theses review/distribution activities are

primarily for: 1) Proposed changes to the stations' Technical

Specifications, 2) Proposed tests and experiments which affect

nuclear safety and are not addressed in the stations' FSAR [Final

Safety Analysis Report] or Technical Specifications, 3)

Environmental radiological procedures, 4) Reportable events

documentation and reports of violations of Technical Specifications,

5) Reports of special reviews and investigations, and 6) Reports of

unplanned onsite releases of radiological material to the environs.

Planned implementation of the proposed Technical Specifications

amendments utilizing Selected Licensee Commitments will result in

the above items being reviewed/received by a different

organizational unit in the future. The organizational unit is to be

either the recently initiated Plant Operations Review Committee

(PORC) or the General Manager, Environmental Services. Personnel

serving on the PORC, and the General Manager, Environmental Services

will be qualified based upon education and experience to review the

operational and technical considerations involved with the

applicable items listed above. No required reviews are being

eliminated by the requested amendments, only the organizational

units responsible for performing the reviews will be changed. Future

reviews of theses items under the auspices of the PORC or the

General Manager, Environmental Services will maintain a quality

level equivalent to that being currently achieved by Duke's

Qualified Reviewer Program, the Station Managers, or the

Duke Nuclear Safety Review Board as applicable. Consequently,

merely changing the organizational units performing future reviews,

or making the additional administrative changes described above,

results in no increase in the probability or consequences of an

accident previously evaluated because the review function will

continue to be conducted in an equivalent manner.

The implementing SLC will also permit proposed amendments to the

stations' Technical Specifications and Operating Licenses to be

approved for the Station Manager by a designee. However, this

individual will occupy a position equivalent to, or higher, in the

Duke organization as the Station manager.

Additionally, the proposed changes do not directly impact the

design or operation of any plant systems or components any more so

than the review and approval processes currently being conducted in

accordance [[Page 14019]] with existing approved Technical

Specifications.

Standard 2. The proposed amendments will not create the

possibility of a new or different kind of accident from any

previously evaluated.

The proposed changes are administrative in nature and primarily

cover the review, distribution, and/or approval function performed

for items identified in existing Technical Specifications. The

quality level of the future reviews will not decrease and the

ability of Duke to identify the possibility for the concurrence of

new or different kinds of accidents prior to implementation will be

maintained. Of specific interest in the consideration of Standard

2 is the review of proposed tests and experiments which

affect station nuclear safety and are not addressed in the FSAR or

Technical Specifications. The Technical Specifications required

reviews of these tests and experiments are not being proposed for

removal by these requested amendments. Only the organizational unit

conducting the review of proposed tests and experiments is being

changed by the requested amendments. The PORC, instead of the

Station Manager, is being assigned the responsibility for conducting

the reviews of proposed tests and experiments in the future. It is

believed that the combined expertise of the PORC membership will

enhance Duke's ability to identify potential situations which could

possibly involve a new, or different, kind of accident.

Standard 3. The proposed amendments will not involve a

significant reduction in any margin of safety.

The changes contained in the requested amendments are

administrative in nature and do not impact the design capabilities

or operation of any plant structures, systems, or components. There

will be no reduction in margin of safety as a result of implementing

these requested amendments. Impact upon margin of safety is a

consideration primarily included in the 10 CFR 50.59 evaluation

process conducted for station procedures, procedure changes, and

nuclear station modifications. The 10 CFR 50.59 evaluation process

in conducted under the auspices of the Duke Qualified Reviewer

Program and is not affected by these requested amendments. The

impact on margin of safety for future Technical Specifications and

Operating License changes will be reviewed by the PORC, but these

reviews will be equivalent in quality to the reviews presently

conducted by the Qualified Reviewers.

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: Atkins Library, University of

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

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: January 13, 1995

Description of amendment request: The proposed amendments would

increase the surveillance test intervals and allowed outage times for

Reactor Trip System (RTS) and Engineered Safety Features Actuation

System (ESFAS) equipment based upon analyses by Westinghouse for the

Westinghouse Owners Group and approved by the NRC. The proposed changes

to the RTS and ESFAS instrumentation are based upon WCAP-10271, its

supplements, and the NRC's safety evaluation reports.

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 - Operation of McGuire in accordance with the

proposed license amendment[s] [do] not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The determination that the results of the proposed changes are

within all acceptable criteria was established in the SERs prepared

for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2,

and WCAP-10271 Supplement 2, Revision 1 issued by letters dated

February 21, 1985, February 22, 1989, and April 30, 1990.

Implementation of the proposed changes is expected to result in an

acceptable increase in total RTS yearly unavailability. This

increase, which is primarily due to less frequent surveillance,

results in an increase of similar magnitude in the probability of an

Anticipated Transient Without Scram (ATWS) and in the probability of

core melt resulting from an ATWS and also results in a small

increase in core damage frequency (CDF) due to ESFAS unavailability.

Implementation of the proposed changes is expected to result in

a significant reduction in the probability of core melt from

inadvertent reactor trips. This is a result of a reduction in the

number of inadvertent reactor trips (0.5 fewer inadvertent reactor

trips per unit per year) occurring during testing of RTS

instrumentation. This reduction is primarily attributable to testing

in bypass and less frequent surveillance.

The reduction in core melt frequency from inadvertent reactor

trips is sufficiently large to counter the increase in ATWS core

melt probability resulting in an overall reduction in total core

melt probability.

The values determined by the WOG and presented in the WCAP for

the increase in CDF were verified by Brookhaven National Laboratory

(BNL) as part of an audit and sensitivity analysis for the NRC

staff. Based on the small value of the increase compared to the

range of uncertainty in the CDF, the increase is considered

acceptable.

Changes to surveillance test frequencies for the RTS [reactor

trip system] interlocks do not represent a significant reduction in

testing. The currently specified test interval for interlock

channels allows the surveillance requirement to be satisfied by

verifying that the permissive logic is in its required state using

the permissive annunciator window. The surveillance as currently

required only verifies the status of the permissive logic and does

not address verification of channel setpoint or operability. The

setpoint verification and channel operability are verified after a

refueling shutdown. The definition of the channel check includes

comparison of the channel status with other channels for the same

parameter. The requirement to routinely verify permissive status is

a different consideration than the availability of trip or actuation

channels which are required to change state on the occurrence of an

event and for which the function availability is more dependent on

the surveillance interval. The change in surveillance requirement to

at least once every refueling does not therefore represent a

significant change in channel surveillance and does not involve a

significant increase in unavailability of the RTS.The proposed

changes do not result in an increase in the severity or consequences

of an accident previously evaluated. Implementation of the proposed

changes affects the probability of failure of the RTS but does not

alter the manner in which protection is afforded nor the manner in

which limiting criteria are established.

Criterion 2 - The proposed license amendment[s] [do] not create

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

accident previously evaluated.

The proposed changes do not result in a change in the manner in

which the RTS provides plant protection. No change is being made

which alters the functioning of the RTS (other than in a test mode).

Rather, the likelihood or probability of the RTS functioning

properly is affected as described above. Therefore, the proposed

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

accident.

The proposed changes do not involve hardware changes except

those necessary to implement testing in bypass. Some existing

instrumentation is designed to be tested in bypass and current

Technical Specifications allow testing in bypass. Testing in bypass

is also recognized by IEEE standards. Therefore, testing in bypass

has been previously approved and implementation of the proposed

changes for testing in bypass does not create the possibility of a

new or different kind of accident from any previously evaluated.

Furthermore, since the other proposed changes do not alter the

functioning of the RTS, the possibility of a new or different kind

of accident from any previously evaluated has not been created.

Criterion 3 - The proposed license amendment[s] [do] not involve

a significant reduction in a margin of safety.

The proposed changes do not alter the manner in which safety

limits, limiting safety [[Page 14020]] system setpoints, or limiting

conditions for operation are determined. The impact of reduced

testing other than as addressed above is to allow a longer time

interval over which instrument uncertainties (e.g., drift) may act.

Experience has shown that the initial uncertainty assumptions are

valid for reduced testing.

Implementation of the proposed changes is expected to result in

an overall improvement in safety by:

1) Less frequent testing will result in fewer inadvertent

reactor trips and actuation of Engineered Safety Features Actuation

System components.

2) Higher quality repairs leading to improved equipment

reliability due to longer allowable repair times.

3) Improvements in the effectiveness of the operating staff in

monitoring and controlling plant operation. This is due to less

frequent distraction of the operator and shift supervisor to attend

to instrumentation testing.

The foregoing analysis demonstrates that the proposed

amendment[s] to McGuire's Technical Specifications [do] not involve

a significant increase in the probability or consequences of a

previously evaluated accident, [do] not create the possibility of a

new or different kind of accident, and [do] not involve a

significant reduction in a margin of safety.

Based upon the preceding analysis, Duke Power Company concludes

that the proposed amendment[s] [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: Atkins Library, University of

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

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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: January 12, 1995

Description of amendment request: The proposed amendments would

revise and clarify portions of Technical Specification (TS) Section

6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee

nuclear stations. The licensee submitted a combined amendment request

covering the three Duke Power nuclear stations. The proposed changes

are described below.

1. Remove the specific assignment of responsibilities for the

review, distribution, and approval activities contained in the

Technical Review and Control Section of each station's TS. The proposed

specifications state that these activities will be performed by a

knowledgeable individual/organization. Approval of the affected

documents is to be at the appropriate manager/superintendent level as

specified in Duke administrative controls.

2. Move the requirement for the review of proposed changes in the

stations' TS and Operating Licenses by the Duke Nuclear Safety Review

Board (NSRB) to Duke administrative procedures (Selected Licensee

Commitments documents) and change the wording of the requirements

covering NSRB meeting frequency. The Oconee TS covering the NSRB are

being rewritten to be consistent with McGuire and Catawba.

3. Add Technical Review and Control Program implementation and

Plant Operations Review Committee (PORC) implementation to the list of

required procedures and programs for each nuclear station.

4. Change or clarify certain TS administrative requirements

covering technical review and control activities or records retention

requirements.

5. For Oconee only, under ``Station Operating Procedures,'' revise

the TS requirements covering the review and approval of station

procedures and temporary procedure changes such that these are now

consistent with the corresponding requirements for McGuire and Catawba.

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:

(It should be noted that the licensee submitted a combined

analysis that covers McGuire, Catawba, and Oconee nuclear stations.)

Standard 1. The proposed amendments will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The provisions of these proposed amendments concern

administrative changes in the stations' Technical Specifications

involving the Technical Review and Control, Procedures and Programs/

Station Operating Procedures, and Records Retention/Station

Operating Records portions of the Administrative Controls Section.

The requested changes primarily affect review and control

activities, but also include other administrative changes affecting

the approval of station procedures (Oconee only), records retention,

and definition of the term ODCM [offsite dose calculation manual]

(McGuire and [Catawba]). The provisions of the proposed amendment

primarily involve the relocation of existing Technical

Specifications review, distribution, or approval requirements to

internal Duke administrative controls. However, implementation of

the proposed amendment does involve changes to several review/

distribution activities. These review/distribution activities are

primarily for: 1) Proposed changes to the stations' Technical

Specifications, 2) Proposed tests and experiments which affect

nuclear safety and are not addressed in the stations' FSAR [Final

Safety Analysis Report] or Technical Specifications, 3)

Environmental radiological procedures, 4) Reportable events

documentation and reports of violations of Technical Specifications,

5) Reports of special reviews and investigations, and 6) Reports of

unplanned onsite releases of radiological material to the environs.

Planned implementation of the proposed Technical Specifications

amendments utilizing Selected Licensee Commitments will result in

the above items being reviewed/received by a different

organizational unit in the future. The organizational unit is to be

either the recently initiated Plant Operations Review Committee

(PORC) or the General Manager, Environmental Services. Personnel

serving on the PORC, and the General Manager, Environmental Services

will be qualified based upon education and experience to review the

operational and technical considerations involved with the

applicable items listed above. No required reviews are being

eliminated by the requested amendments, only the organizational

units responsible for performing the reviews will be changed. Future

reviews of these items under the auspices of the PORC or the General

Manager, Environmental Services will maintain a quality level

equivalent to that being currently achieved by Duke's Qualified

Reviewer Program, the Station Managers, or the Duke Nuclear Safety

Review Board as applicable. Consequently, merely changing the

organizational units performing future reviews, or making the

additional administrative changes described above, results in no

increase in the probability or consequences of an accident

previously evaluated because the review function will continue to be

conducted in an equivalent manner.

The implementing SLC will also permit proposed amendments to the

stations' Technical Specifications and Operating Licenses to be

approved for the Station Manager by a designee. However, this

individual will occupy a position equivalent to, or higher, in the

Duke organization as the Station Manager.

Additionally, the proposed changes do not directly impact the

design or operation of any plant systems or components any more so

than the review and approval processes currently being conducted in

accordance with existing approved Technical Specifications.

Standard 2. The proposed amendments will not create the

possibility of a new or different kind of accident from any

previously evaluated.

The proposed changes are administrative in nature and primarily

cover the review, [[Page 14021]] distribution, and/or approval

function performed for items identified in existing Technical

Specifications. The quality level of the future reviews will not

decrease and the ability of Duke to identify the possibility for the

occurrence of new or different kinds of accidents prior to

implementation will be maintained. Of specific interest in the

consideration of Standard 2 is the review of proposed tests

and experiments which affect station nuclear safety and are not

addressed in the FSAR or Technical Specifications. The Technical

Specifications required reviews of these tests and experiments are

not being proposed for removal by these requested amendments. Only

the organizational unit conducting the review of proposed tests and

experiments is being changed by the requested amendments. The PORC,

instead of the Station Manager, is being assigned the responsibility

for conducting the reviews of proposed tests and experiments in the

future. It is believed that the combined expertise of the PORC

membership will enhance Duke's ability to identify potential

situations which could possibly involve a new, or different, kind of

accident.

Standard 3. The proposed amendments will not involve a

significant reduction in any margin of safety.

The changes contained in the requested amendments are

administrative in nature and do not impact the design capabilities

or operation of any plant structures, systems, or components. There

will be no reduction in margin of safety as a result of implementing

these requested amendments. Impact upon margin of safety is a

consideration primarily included in the 10 CFR 50.59 evaluation

process conducted for station procedures, procedure changes, and

nuclear station modifications. The 10 CFR 50.59 evaluation process

is conducted under the auspices of the Duke Qualified Reviewer

Program and is not affected by these requested amendments. The

impact on margin of safety for future Technical Specifications and

Operating License changes will be reviewed by the PORC, but these

reviews will be equivalent in quality to the reviews presently

conducted by the Qualified Reviewers.

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: Herbert N. Berkow

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January 27, 1995

Description of amendment request: The requested change would modify

Section 5.3.1, Fuel Assemblies, of the Waterford 3 technical

specifications. The requested change increases the maximum enrichment

for the spent fuel pool and containment temporary storage rack from 4.1

to 4.9 weight percent U-235 when fuel assemblies contain fixed poisons.

Waterford 3 plans to use higher enriched fuel in the next fuel cycle

(Cycle 8) to meet the energy plans and maintain a reload batch size

similar to that used in Cycles 6 and 7.

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 change will increase the fuel enrichment limit in

order to meetthe cycle energy requirements while maintaining fuel

batch sizes consistent with previous cycle designs. The calculated

k-effective, including uncertainties, demonstrate substantial margin

to criticality in the storage racks for both normal and accident

conditions. No changes to the facility are required. No new modes of

operating the fuel storage or transfer systems are required, except

a restriction to limit the use of the new fuel vault to fuel with a

maximum enrichment of 4.1 weight percent U-235. This restriction

will be implemented by administrative controls. Since the plant

equipment and operation are essentially the same, there is no

significant increase in the probability of a criticality accident.

Since a criticality event is demonstrated to be unfeasible, there

are no increased adverse consequences for such a postulated event.

As previously discussed, the proposed change will not result in

a physical change to the facility nor will it result in a

significant change to the operation of the facility; therefore, it

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

accident from any accident previously evaluated.

The proposed change has been analyzed to establish a k-

effective, including uncertainties, at or below the NRC criticality

acceptance criteria of k-effective below 0.95 including

uncertainties at the 95/95 probability/confidence level; therefore,

there is no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

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

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

Date of amendment request: January 16, 1995

Description of amendment request: The proposed amendment would

revise the TMI-1 Technical Specifications (TS) to incorporate certain

improvements from the Revised Standard Technical Specifications (TS)

for Babcock & Wilcox nuclear power plants (NUREG-1430). The amendment

would also change the bases incorporating the results of analyses to

support allowance for drift of the pressurizer code safety valve

setpoint. One of the proposed STS improvements involves a change to

Chapter 6, Administrative Controls, affecting both TMI-1 and TMI-2 TSs.

A separate notice of consideration of issuance of amendment to facility

operating license is being issued for the proposed TMI-2 TSs Change.

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

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated.

The proposed amendments involve a) an administrative change to

both the TMI-1 and TMI-2 Technical Specifications which is

consistent with the B&W Standard Technical Specifications (STS),

NUREG-1430, and b) changes to the TMI-1 Technical Specifications

which are consistent with the STS. This change does not involve any

change to system or equipment configuration. The proposed amendment

revises certain surveillance requirements, extends certain

surveillance intervals as evaluated above, or involves changes that

are purely

administrative. The reliability of systems and components relied

upon to prevent or mitigate the consequences of accidents previously

evaluated is not degraded by the proposed changes. Assurance of

system and equipment availability is maintained. Therefore, this

change does not increase the probability of occurrence or the

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The changes

only involve changes to surveillance requirements that are

consistent with STS and with the ASME Code. No new failure modes are

created and thus the changes are bounded by accidents previously

evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not [[Page 14022]] involve a significant reduction

in a margin of safety. Each of these changes is compatible with the

STS and has been evaluated to preserve the level of safety assured

by the current TS.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Phillip F. McKee

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 20, 1995

Description of amendment request: The proposed amendment would

revise the fire hazards analysis for the River Bend Station (RBS) by

allowing a deviation from 10 CFR 50, Appendix R, Section III.G.3 with

respect to the requirement for a fixed fire suppression system in fire

area C-17. This area houses the control building heating, ventilation

and air conditioning (HVAC) systems and the loss due to a fire could

cause the loss of main control room habitability. C-17 does not have a

fixed fire suppression system but depends upon the use of the existing

remote shutdown system as described in the updated safety analysis

report (USAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) The request does not involve a significant increase in the

probability or consequences of accident previously evaluated.

The event of concern is a fire in fire area C-17. The low fire

loading and sparse concentration of exposed combustible material in

fire area C-17 would limit fire spread. However, for this scenario

all equipment in fire area C-17 will be assumed lost. Fire area C-17

contains the air handling units for the main control room envelope.

The loss of both air handling units would cause the control building

chillers to stop running due to a logic tie requiring air flow

through the air handling equipment for the chilled water system to

operate during normal operation. The loss of the HVAC system in the

control building would cause the main control room and the equipment

rooms to begin heating up if exposed to design summer conditions.

Operator actions can be accomplished to minimize the heat up rates

for the rooms prior to the areas reaching equipment temperature

limits. This would allow the operators to begin the shutdown process

from the main control room. If the main control room continued to

heat up, the operators could accomplish the shutdown using the

remote shutdown system. HVAC for the remote shutdown panel is

located in fire area C-4 and would not be damaged by a fire in fire

area C-17. Operation of the control building HVAC system from the

remote shutdown panel bypasses the logic between the chilled water

system and the air handling system. This would allow restart of the

HVAC system for all areas except the main control room. The scenario

would conclude in a manner similar to that described in RBS USAR

Appendix 15A, Event 52, ``Reactor Shutdown From Outside Main Control

Room.''

In summary, the probability of a fire occurring in fire area C-

17 is not increased. However, if a fire were to occur in fire area

C-17 which caused the loss of main control room HVAC, the remote

shutdown system would provide an acceptable method of shutdown. The

low fire loading and sparse concentration of exposed combustible

material in fire area C-17 would limit fire spread. Therefore, this

request does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

2) The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

The event of concern is a fire in fire area C-17. Fire area C-17

does not have a fixed suppression system as required by 10 CFR 50,

Appendix R, Section III.G.3. Fire suppression systems are generally

used to limit fire spread, once the heat of the fire opens thermally

sensitive sprinklers. The low fire loading and sparse concentration

of exposed combustible material in fire area C-17 would limit fire

spread. However, for the purpose of event analysis, all equipment in

fire area C-17 is assumed lost. Thus a fire in fire area C-17 is

bounded by the same analysis with or

without a fixed suppression system in terms of equipment

availability.

The proposed method of shutdown for a fire in fire area C-17

will be changed in that the remote shutdown system will be credited.

Use of the remote shutdown system is bounded by RBS USAR Appendix

15A, Event 52, ``Reactor Shutdown From Outside Main Control Room.''

The HVAC for the remote shutdown panel is located in fire area C-4

and would be undamaged by a fire in fire area C-17. Operation of the

control building HVAC system from the remote shutdown panel bypasses

the logic between the chilled water system and the air handling

system. This would allow restart of the HVAC system for all areas

except the main control room.

In summary, if a fire were to occur in fire area C-17 which

caused the loss of main control room HVAC, the remote shutdown

system would provide an acceptable method of shutdown. Since, for

the purpose of event analysis, all equipment in fire area C-17 is

assumed lost, a fire in fire area C-17 is bounded by the same

analysis with or without a fixed suppression system in terms of

equipment availability. Therefore, this request does not create the

possibility of occurrence of a new or different kind of accident

from any accident previously evaluated.

3) The request does not involve a significant reduction in a

margin of safety.

In this case, the margin of safety is implicit rather than being

explicitly expressed as a numerical value. An implicit margin of

safety involves conditions for NRC acceptance. Since the RBS

Technical Specification Bases do not specifically address a margin

of safety for fire protection, the SAR, the NRC's Safety Evaluation

Report (SER), and appropriate other licensing basis documents were

reviewed to determine if the proposed change would result in a

reduction in a margin of safety. As stated, in part, in Attachment 4

to NPF-47:

EOI shall implement and maintain in effect all provisions of the

approved fire protection program as described in the Final Safety

Analysis Report for the facility through Amendment 22 and as

approved in the SER dated May 1984 and Supplement 3 dated August

1985 subject to provisions 2 and 3....

As discussed in the Reason for Request, SSER 3 dated August 1985

states, in part:

On the basis of its evaluation the staff finds that the

applicant's fire protection program with approved deviations is in

conformance with the guidelines of BTP CMEB 9.5-1, sections III.G,

III.J, and III.O of Appendix R to 10CFR50, and GDC 3, and is,

therefore, acceptable.

Thus, the margin of safety in this case can be defined as

conformance with the specified fire protection guidelines. 10 CFR

50, Appendix R, Section III.G.3, requires, in part, that alternative

shutdown capability be provided for areas where adequate separation

of redundant safe shutdown components cannot be provided. In

addition, fire detection and a fixed fire suppression system must be

installed in the area, room, or zone under consideration. Since fire

area C-17 does not have a fixed suppression system, use of the

remote shutdown system for a fire in this fire area would deviate

from the requirements of 10 CFR 50, Appendix R, Section III.G.3.

However, as discussed previously, the low fire loading and sparse

amount of exposed combustibles compensate for the lack of a fixed

fire suppression system. There is no adverse impact on the ability

to achieve and maintain safe shutdown. Therefore, this request does

not involve a significant reduction in a margin of safety.

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

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

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

[[Page 14023]] amendment request involves no significant hazards

consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

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

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

Connecticut

Date of amendment request: July 28, 1994

Description of amendment request: The proposed amendment would add

a footnote to Technical Specifcaiton 3.5.C. The footnote would state

that the operability of the feedwater coolant injection (FWCI) system

be independent of its seismic 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:

NNECO has reviewed the proposed change in accordance with

10CFR50.92 and concluded that the change does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed change does not involve an SHC because the

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

Any postulated failure in the non-seismic portion of the FWCI

subsystem may result in a loss of feedwater flow transient. However

comparing the probability of occurrence of a seismic event, any

increase in the probability of occurrence of a loss of feedwater

event would be small. The proposed change would have no impact on

the probability of occurrence of any other accident, including LOCAs

[loss of coolant accidents].

The FWCI subsystem will continue to be maintained as QA Category

1 (except for the seismic attribute). Therefore, it will remain

available for accident mitigation for most scenarios. Nevertheless,

LOCA analyses have been reevaluated to demonstrate that FWCI is not

necessary to show compliance with 10CFR50.46. Potentially limiting

LOCA scenarios have been analyzed without the FWCI subsystem using

an approved LOCA methodology. An active single failure was

postulated in addition to not taking credit for the FWCI subsystem.

Based on the results of these analyses, the current design basis

large and small break LOCAs remain bounding. Moreover, FWCI is not

credited in mitigating any of the non-LOCA transients/accidents.

Safe shutdown following a seismic event can be achieved using

the LPCI [low pressure coolant injection] and ESW [emergency service

water] systems, and the SRVs [safety relief valves], which are all

seismically qualified. Therefore, the FWCI system is not required to

mitigate a seismic event.

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

from any previously analyzed.

Seismic reclassification of portions of FWCI does not create the

possibility of a new kind of an accident. The portion of the piping

up to the second isolation valve (from the RPV [reactor pressure

vessel]), is seismically qualified and will remain classified as

seismic. This ensures that a postulated failure in the non-seismic

portion of piping or components does not degrade containment

integrity or result in a blowdown of the RPV. Consequential and

environmental effects of a FW [feedwater] piping failure have been

analyzed in the HELB [high energy line break] program and have been

found to be acceptable.

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

All accidents, including LOCAs, can be mitigated without using

FWCI. FWCI is also not necessary for safe shutdown following a

seismic event. The intended function of the FWCI subsystem is to

reduce the likelihood of core uncovery during the lifetime of the

plant. The CS [core spray] and LPCI subsystems provide redundant and

diverse means of injecting water to the RPV. The FWCI subsystem

provides an additional diverse means to inject water. Since FWCI

will be maintained QA Category 1 (except for the seismic attribute),

it will continue to provide the additional diversity to the

injection systems. Considering the intended function of the

subsystem and the credit taken in the accident analysis,

reclassifying FWCI to be non-seismic does not significantly reduce

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

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

London Turnpike, Norwich, CT 06360.

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

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

06141-0270.

NRC Project Director: Phillip F. McKee

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: January 9, 1995, as supplemented

February 7, 1995

Description of amendment requests: The proposed amendments would

revise Prairie Island Nuclear Generating Plant Technical Specification

(TS) 4.12, ``Steam Generator Tube Surveillance,'' to incorporate

revised acceptance criteria for steam generator tubes with degradation

in the tubesheet roll expansion region. These criteria for steam

generator tube acceptance were developed by Westinghouse Electric

Corporation and are known as F* (F-Star'') and L*

(L-Star''). These criteria would be utilized to avoid

unnecessary plugging and sleeving of steam generator tubes.

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[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

F* Steam Generator Tube Repair Criteria

The supporting technical and safety evaluations of the subject

criterion demonstrate that the presence of the tubesheet will

enhance the tube integrity in the region of the hardroll by

precluding tube deformation beyond its initial expanded outside

diameter. The resistance to both tube rupture and tube collapse is

strengthened by the presence of the tubesheet in that region. The

results of hardrolling of the tube into the tubesheet is an

interference fit between the tube and the tubesheet. Tube rupture

cannot occur because the contact between the tube and tubesheet does

not permit sufficient movement of tube material. The radial preload

developed by the rolling process will secure a postulated separated

tube end within the tubesheet during all plant conditions. In a

similar manner, the tubesheet does not permit sufficient movement of

tube material to permit buckling collapse of the tube during

postulated LOCA [loss-of-coolant accident] loadings.

The F* length of roll expansion is sufficient to preclude tube

pullout from tube degradation located below the F* distance,

regardless of the extent of the tube degradation. The existing

Technical Specification leakage rate requirements and accident

analysis assumptions remain unchanged in the unlikely event that

significant leakage from this region does occur. As noted above,

tube rupture and pullout is not expected for tubes using the F*

criterion. Any leakage out of the tube from [[Page 14024]] within

the tubesheet at any elevation in the tubesheet is fully bounded by

the existing steam generator tube rupture analysis included in the

Prairie Island Plant USAR [Updated Safety Analysis Report]. For

plants with partial depth roll expansion like Prairie Island, a

postulated tube separation within the tube near the top of the roll

expansion (with subsequent limited tube axial displacement) would

not be expected to result in coolant release rates equal to those

assumed in the USAR for a steam generator tube rupture event due to

the limited gap between the tube and tubesheet. The proposed

plugging criterion does not adversely impact any other previously

evaluated design basis accident.

Leakage testing of roll expanded tubes indicates that for roll

lengths approximately equal to the F* distance, any postulated

faulted condition primary to secondary leakage from F* tubes would

be insignificant.

L* Steam Generator Tube Repair Criteria

The presence of the tubesheet enhances steam generator tube

integrity in the region of the hardroll by precluding tube

deformation beyond its initial expanded outside diameter. The

resistance to both tube rupture and tube collapse is strengthened by

the presence of the tubesheet in that region. The result of the

hardroll of the tube into the tubesheet is an interference fit

between the tube and the tubesheet. Tube rupture cannot occur

because the contact between the tube and tubesheet does not permit

sufficient movement of tube materials. In a similar manner, the

tubesheet does not permit sufficient movement of tube material to

permit buckling collapse of the tube during postulated LOCA

loadings.

The type of degradation for which the L* criteria has been

developed (cracking with an axial or near axial orientation) has

been found not to significantly reduce the axial strength of a tube.

An evaluation including analysis and testing has been done to

determine the strength reduction for the axial loads with simulated

axial and near axial cracks. This evaluation provided the basis for

the acceptance criteria for tube degradation subject to the L*

criteria.

The length of roll expansion above L* is sufficient to preclude

significant leakage from tube degradation located below the L*

distance. The existing Technical Specification leakage rate

requirements and accident analysis assumptions remain unchanged in

the unlikely event that significant leakage from this region does

occur. As noted above, tube rupture and pullout is not expected for

tubes using the alternate plugging criteria.

Any leakage out of the tube from within the tubesheet at any

elevation in the tubesheet is fully bounded by the existing steam

generator tube rupture analysis included in the Prairie Island

Updated Safety Analysis Report. The proposed alternate plugging

criteria do not adversely impact any other previously evaluated

design basis accident.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

F*

Implementation of the proposed F* criterion does not introduce

any significant changes to the plant design basis. Use of the

criterion does not provide a mechanism to initiate an accident

outside of the region of the expanded portion of the tube. Any

hypothetical accident as a result of any tube degradation in the

expanded portion of the tube would be bounded by the existing tube

rupture accident analysis. Tube bundle structural integrity will be

maintained. Tube bundle leaktightness will be maintained such that

any postulated accident leakage from F* tubes will be negligible

with regards to offsite doses.

L*

Implementation of the proposed alternate tubesheet tube plugging

criteria does not introduce changes to the plant design basis. Use

of the criteria does not provide a mechanism to result in an

accident outside of the region of the tubesheet expansion. Any

hypothetical accident as a result of any tube degradation in the

expanded portion of the tube would be bounded by the existing tube

rupture accident analysis.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

F*

The use of the F* criterion has been demonstrated to maintain

the integrity of the tube bundle commensurate with the requirements

of Reg Guide 1.121 [Bases for Plugging Degraded PWR Steam

Generator Tubes] (intended for indications in the free

span of tubes) and the primary to secondary pressure boundary under

normal and postulated accident conditions. Acceptable tube

degradation for the F* criterion is any degradation indication in

the tubesheet region, more than the F* distance below the bottom of

the transition between the roll expansion and the unexpanded tube.

The safety factors used in the verification of the strength of the

degraded tube are consistent with the safety factors in the ASME

Boiler and Pressure Vessel Code used in steam generator design. The

F* distance has been verified by testing to be greater than the

length of roll expansion required to preclude both tube pullout and

significant leakage during normal and postulated accident

conditions. Resistance to tube pullout is based upon the primary to

secondary pressure differential as it acts on the surface area of

the tube, which includes the tube wall cross-section, in addition to

the inner diameter based area of the tube. The leak testing

acceptance criteria are based on the primary to secondary leakage

limit in the Technical Specifications and the leakage assumptions

used in the USAR accident analysis.

Implementation of the tubesheet plugging criterion will decrease

the number of tubes which must be taken out of service with tube

plugs or repaired with sleeves. Both plugs and sleeves reduce the

RCS (reactor coolant system) flow margin; thus, implementation of

the F* criterion will maintain the margin of flow that would

otherwise be reduced in the event of increased plugging or sleeving.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the USAR or the Technical Specification

Bases.

L*

The use of the alternate tubesheet plugging criteria has been

demonstrated to maintain the integrity of the tube bundle

commensurate with the requirements of Reg. Guide 1.121 for

indications in the free span of tubes and the primary to secondary

pressure boundary under normal and postulated accident conditions.

Acceptable tube degradation for the L* criteria is any degradation

indication with axial or nearly axial cracking in the tubesheet

region, more than the L* distance below the bottom of the transition

between the roll expansion and the unexpended tube. For tubes with

axial or nearly axial cracks the strength of the tube relative to an

axial load would not be reduced below the strength required to

resist potential axial loads. The safety factors used in the

verification of the strength of the degraded tube are consistent

with the safety factors in the ASME Boiler and Pressure Vessel Code

used in steam generator design. The L* distance has been verified by

testing to be greater than the length of roll expansion required to

preclude significant leakage during normal and postulated accident

conditions. The leak testing acceptance criteria are based on the

primary to secondary leakage limit in the Technical Specifications

and the leakage assumptions used in the USAR accident analyses.

Implementation of the proposed tubesheet plugging criteria will

decrease the number of tubes which must be taken out of service with

tube plugs or repaired with sleeves. Both plugs and sleeves reduce

the RCS flow margin, thus implementation of the alternate plugging

criteria will maintain the margin of flow that would otherwise be

reduced in the event of increased plugging or sleeving.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the Updated Safety Analysis Report or the

bases of the Technical Specifications.

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. This

notice supersedes the staff's previous notice which was published in

the Federal Register February 1, 1995 (60 FR 6307).

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: Cynthia Carpenter, Acting [[Page 14025]]

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 23, 1995

Description of amendment requests: The proposed amendments would

revise the wording in the Prairie Island technical specifications to

allow implementation of exemptions to the schedule requirements of 10

CFR Part 50, Appendix J. A related exemption request would grant

temporary relief from the requirements of 10 CFR Part 50, Appendix J,

Section III.D.1.(a) which requires Prairie Island Unit 2 to perform a

Type A test in the May 1995 refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment is an administrative change which allows

implementation of approved exemptions to the regulations and by

itself does not change any retest schedules.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by the proposed amendment.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

The proposed amendment is an administrative change which allows

implementation of approved exemptions to the regulations and by

itself does not change any retest schedules.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be created

by the proposed amendment.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety

The proposed amendment is an administrative change which allows

implementation of approved exemptions to the regulations and by

itself does not change any retest schedules.

Therefore, a significant reduction in the margin of safety would

not be involved with the proposed amendment.

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: Cynthia Carpenter, Acting

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

Station,Unit No. 1, Washington County, Nebraska

Date of amendment request: February 10, 1995

Description of amendment request: The proposed amendment to the

technical specifications (TSs) would relocate the requirements for the

incore instrumentation (ICI) system from the TS to the Updated Safety

Analysis Report (USAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Incore Instrumentation (ICI) System is used to measure core

power distribution for the purpose of Limiting Conditions for

Operation (LCO) monitoring of Technical Specification (TS) limits on

linear heat rate, unrodded planer radial peaking factor, unrodded

integrated radial peaking factor, and azimuthal power tilt. The ICI

System has no safety purpose itself; it measures parameters which

have safety significance. No change to the monitored parameters is

proposed. The proposed changes will relocate requirements on the

number and distribution of incore detectors used by the ICI System

when measuring these parameters from the TS to the Updated Safety

Analysis Report (USAR). Changes to the requirements can be made

without NRC approval when the changes meet the criteria of 10 CFR

50.59. Changes to the ICI System requirements that do not meet the

criteria of 10 CFR 50.59 must be approved by the NRC by license

amendment.

Relocation of the requirements on the ICI System from the TS to

the USAR does not increase the probability or consequences of any

accident previously analyzed because the ICI System is neither a

precursor nor a mitigator for any analyzed accident. The ICI System

is used to ensure that operation within the LCOs for linear heat

rate, unrodded planer radial peaking factor, unrodded integrated

radial peaking factor, and azimuthal power tilt is maintained.

However, its operation serves no mitigation function associated with

any USAR Section 14 accident analysis. The parameters measured by

the ICI System are important parameters in many accident analyses;

however, this proposed change does not remove or revise the limits

on these parameters.

Additionally, it is proposed to revise TS 2.10.4(1)(b) to

clarify its requirements. Currently TS 2.10.4(1) part (b) applies

while operating under the provisions of part (a) if the plant

computer incore detector alarms become inoperable. This is incorrect

in that part (a) applies when the linear heat rate is being

monitored by the ICI System and the linear heat rate is exceeding

its limits as indicated by valid detector alarms. Part (b) of this

specification should apply only if the linear heat rate is being

monitored by the ICI System, is within its limits, and the plant

computer incore detector alarms are inoperable.

Administrative changes are also proposed which correct grammar

and renumber/relocate portions of the TS and bases to other TS, to

correspond to the proposed change to relocate ICI System

requirements.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any previously evaluated.

The ICI System will continue to be used to monitor TS limits on

core power distribution. There will be no physical alterations to

the plant configuration, changes to setpoint values, or changes to

the implementation of setpoints or limits as a result of this

proposed change.

The proposed change to TS 2.10.4(1)(b) only clarifies its

requirements. The proposed change is more restrictive in that TS

2.10.4(1)(b), as currently written, could be interpreted to allow

continued operation for up to seven days with the linear heat rate

exceeding its limits. The proposed change clarifies this

specification to ensure that TS 2.10.4(1)(a) is applied if the

linear heat rate is exceeded while being monitored by the ICI

System. TS 2.10.4(1)(a) requires that the linear heat rate be

restored within one hour or a plant shutdown initiated.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

(3) The proposed changes do not involve a significant reduction

in a

margin of safety.

The ICI System is used to measure core power distribution

parameters which are a direct measure of the margin of safety. The

limits on these parameters are not changed. Therefore, the proposed

change (i.e., relocation of the ICI System operability requirements

to the USAR and/or plant procedures) does not involve a significant

reduction in a margin of safety.

The proposed change to TS 2.10.4(1)(b) helps ensure that the

margin of safety is maintained by clarifying when the TS is

applicable. This clarification ensures that the more restrictive

actions of TS 2.10.4(1)(a) are taken if the linear heat rate is

exceeded while being monitored by the ICI System. Therefore, the

proposed change does not involve a significant reduction in a margin

of safety. [[Page 14026]]

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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875

Connecticut Avenue, NW., Washington, DC 20009-5728

NRC Project Director: Theodore R. Quay

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

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

Obispo County, California

Date of amendment requests: December 30, 1994 (Reference LAR 94-12)

Description of amendment requests: The proposed amendments would

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

Nuclear Power Plant, Unit Nos. 1 and 2, to revise TS 2.2, 3/4.3.1, 3/

4.3.2, 3/4.3.3, 3/4.4.4, 3/4.4.9, 3/4.5.2, 3/4.8.1, 3/4.8.2, 3/4.9.2,

3/4.9.9, and 3/4.10.3. The specific TS changes proposed are as follows:

(1) The TS issued in License Amendments (LAs) 84/83 would be

changed to (a) revise the value of the overpower Delta-temperature

(OPDT) constant K6 in TS 2.2.1, Table 2.2-1, Note 3; (b) revise the

reactor coolant system (RCS) loop Delta-T function; and (c) make

editorial corrections for clarification and consistency to TS 2.2.1

(and TS 2.2.1 Bases), TS 3/4.3.1, and TS 3/4.3.2.

In revising the RCS loop Delta-T function, the licensee would (a)

incorporate the 0.99 multiplying factor listed in TS 2.2.1, Table 2.2-

1, Note 5, and TS 3/4.3.2, Table 3.3-4, Note 2, into constants B1

through B4; (b) change ``Steam Generator (SG) Water Level Low-Low'' in

TS 3/4.3.2, Table 3.3-3 and Table 4.3-2, Functional Unit 6.c,

``Auxiliary Feedwater'' (AFW), by deleting the Mode 3 applicability of

the RCS loop Delta-T function and by adding a footnote to the Mode 3

applicability of the SG water level low-low function requiring that the

trip time delay (TTD) associated with the SG water level low-low

channel be less than or equal to 464.1 seconds; (c) change TS 3/4.3.1,

Table 3.3-1, Action 27, and TS 3/4.3.2, Table 3.3-3, Action 29, by

allowing up to four RCS loop Delta-T channels to be inoperable with the

TTD threshold power level for zero seconds time adjusted to 0-percent

rated thermal power (RTP) and by allowing the affected SG water level

low-low channels to be placed in the tripped condition, with one

inoperable RCS loop Delta-T channel; and (d) change the Table 3.3-1 and

Table 3.3-3 ``Channels to Trip'' and ``Minimum Channels Operable''

columns to not applicable (N.A.).

(2) The TS issued in LAs 70/69 would be changed to (a) delete

references to the plant vent noble gas activity monitors (RM-14A and

RM-14B) and footnote references to applicability of the containment

ventilation exhaust radiation monitors (RM-44A and RM-44B) in TS Tables

3.3-3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3 and TS 4.9.9; and (b)

revise the ``Trip Setpoint and Allowable Values'' column in TS Table

3.3-4, Functional Unit 3.c.4), to reference the offsite dose

calculation procedure (ODCP).

(3) Cycle-specific information in TS 4.3.2.1, TS 3.3.3.6, TS

4.4.4.1, TS 4.5.2, TS 3.8.1.1, TS 3.8.2.1, and TS 3.8.2.2 that is no

longer necessary would be deleted.

(4) The word ``analog'' would be deleted from TS 4.4.9.3.1, TS

4.9.2, and TS 4.10.3.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change to the OPDT constant K6 is conservative and

will not cause any design or analysis acceptance criteria to be

exceeded. There is no effect on the structural and functional

integrity of any plant system. The OPDT function is part of the

accident mitigation response and is not itself an initiator for any

transient. This change does not affect the integrity of the fission

product barriers for mitigation of radiological dose consequences as

a result of an accident.

The proposed change to incorporate the 0.99 multiplier into the

TTD constants is an administrative change and has no effect on plant

operation. The proposed change to delete Mode 3 applicability of the

RCS Loop Delta-T function does not affect any design or analysis

results. Allowing up to 4 RCS Loop Delta-T channels to be inoperable

with the TTD threshold power level for zero seconds time delay

adjusted to 0% RTP is conservative with respect to ESFs [engineered

safety features] and reactor trip actuation time. Allowing the SG

[steam generator] water level low-low channels affected by the

inoperable RCS Loop Delta-T channels to be placed in the tripped

condition is also conservative with respect to reactor trip and AFW

pumps start. The change to the Channels to Trip and Minimum Channels

Operable columns is a clarifying change to reflect the proposed

changes to the action statements and identifies that the RCS Loop

Delta-T does not provide a reactor trip function. Therefore, the

proposed changes to the RCS Loop Delta-T function do not affect any

of the accident analysis results.

The proposed changes to revise Table 3.3-4, Functional Unit

3.c.4), and to delete cycle-specific TS, TS references to RM-14A and

RM-14B, and the word ``analog'' from the analog channel operation

test are administrative and have no effect on plant operation.

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 change to the OPDT constant K6 does not affect the

assumed accident initiation sequences. No new operating

configuration is being imposed by the change to K6 that would create

a new failure scenario. No new failure modes are being created for

any plant equipment.

The proposed changes to the RCS Loop Delta-T function do not

involve any physical modification to any plant system or change the

methodology by which any safety-related system performs its

function.

1The proposed changes to revise Table 3.3-4, Functional Unit

3.c.4), and to delete cycle-specific TS, TS references to RM-14A and

RM-14B, and the word ``analog'' from the analog channel operation

test are administrative, would not result in any physical alteration

to any plant system, and would not be a change in the method by

which any safety-related system performs its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed change to the OPDT constant K6 will not affect any

accident analysis assumptions, initial conditions, or results.

The proposed changes to the RCS Loop Delta-T function do not

affect any accident analysis assumptions, initial conditions, or

results.

The proposed changes to revise Table 3.3-4, Functional Unit

3.c.4), and to delete cycle-specific TS, TS references to RM-14A and

RM-14B, and the word ``analog'' from the analog channel operation

test are administrative and clarify the TS. These proposed changes

have no effect on current operating methodologies or actions that

govern plant performance.

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 [[Page 14027]] 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, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

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

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

Date of application for amendments: September 26, 1994

Description of amendment request: The proposed TS changes extend

surveillance test intervals and allowable out-of-service times for the

testing and/or repair of instrumentation that actuate the Reactor

Protection System, Primary Containment Isolation, Core and Containment

Cooling systems, Control Rod Blocks, Radiation Monitoring systems, and

Alternate Rod Insertion/Recirculation Pump Trip.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed TS changes increase the STIs and AOTs for actuation

instrumentation based on analyses described and justified in

Licensing Topical Reports (References 2 through 8) [see licensee's

September 26, 1994 application for reference information] which have

been evaluated in associated Safety Evaluation Reports. These

changes were incorporated into PBAPS Technical Specifications

consistent with NUREG-1433. TS requirements that govern Operability

or routine testing of plant instruments are not assumed to be

initiators of any analyzed event because these instruments are

intended to prevent, detect or mitigate accidents. Therefore, these

changes will not involve an increase in the probability of

occurrence of an accident previously evaluated. Additionally, these

changes will not increase the consequences of an accident previously

evaluated because the proposed change will not involve any physical

changes to plant systems, structures, or components (SSC), or the

manner in which these SSC are operated, maintained, modified, or

inspected. The changes will not alter the operation of equipment

assumed to be available for the mitigation of accidents or

transients by the plant safety analysis or licensing basis. As

justified in References 1 through 8, the proposed changes establish

or maintain adequate assurance that components are operable when

necessary for the prevention or mitigation of accidents or

transients and that plant variables are maintained within limits

necessary to satisfy the assumptions for initial conditions in the

safety analyses. These changes establish or modify time limits

allowed for operation with inoperable instrument channels based on

the analyses in References 1 through 8 and will not allow continuous

plant operation with plant conditions such that a single failure

will result in a loss of any safety function. Therefore, these

changes will not increase the consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

These proposed changes will not involve any physical changes to

SSC, or the manner in which these SSC are operated, maintained,

modified, tested, or inspected. Therefore, these changes will not

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

any accident previously evaluated. The changes in methods governing

normal plant operation are consistent with the current safety

analysis assumptions. Therefore, these changes will not create the

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

previously evaluated.

3) The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed TS changes increase the STIs and AOTs for actuation

instrumentation based on analyses described and justified in

Licensing Topical Reports (References 2 through 8) which have been

evaluated in associated Safety Evaluation Reports. These changes

were incorporated into PBAPS Technical Specifications consistent

with NUREG-1433. These changes can be classified into one of the

following three categories:

a. Changes to the minimum STIs and AOTs for the testing and/or

repair of instrumentation based on the results of generic analyses

in References 1 through 8;

b. Changes to conditions, required actions, and completion times

needed to make PBAPS TS requirements consistent with the assumptions

used in the analyses in References 1 through 8; and,

c. Changes that reformat, renumber, and/or reword existing

requirements to incorporate the changes above.

All of the proposed changes will be incorporated into the PBAPS

custom Technical Specifications using the same approach and specific

requirements used in Reference 12.

There is no significant reduction in the margin of safety

resulting from changes to the STIs and AOTs for the testing and/or

repair of instrumentation based on the results of the analyses in

References 1 through 8. These analyses determined that there is no

significant change in the availability and/or reliability of

instrumentation as a result of this change in STIs and AOTs. PECO

Energy performed reviews that confirmed these analyses are

applicable to PBAPS and that there would be no effect on the

identification of excessive instrument setpoint drift as a result of

increasing from monthly to quarterly the minimum interval between

instrument functional tests. The proposed required actions ensure

that actions to mitigate loss of single failure tolerance is

initiated within 24 hours (12 hours for RPS) in accordance with the

results of the analyses in References 1 through 8 and action to

mitigate a loss of instrument function is initiated within 1 hour.

The proposed changes which replace the shutdown actions

associated with inoperable instrumentation with actions to declare

the supported system inoperable does not involve a reduction in a

margin of safety. The proposed changes ensure that appropriate

compensatory measures are taken commensurate with approved TS

Actions for the affected systems and the safety analyses. In

addition, the proposed changes provide the benefit of avoiding an

unnecessary shutdown transient when appropriate measures are

available to compensate for the inoperable instrumentation.

There is no significant reduction in the margin of safety

resulting from changes that reformat, renumber, and/or reword

existing requirements to incorporate the changes 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: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas

Company,Delmarva Power and Light Company, and Atlantic City

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

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

Date of application for amendments: November 17, 1994

Description of amendment request: The proposed changes to the

Technical Specifications (TS) are being requested to support

modifications 5384 and 5386 which upgrade the Main Stack and Vent Stack

Radiation Monitoring Systems. [[Page 14028]]

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.

Neither the Main Stack nor the Vent Stack Radiation Monitoring

Systems serve as an initiator or contributor to any accidents

previously evaluated. The systems provide indication and detection

of radioactivity and effluent release in the main and vent stacks.

The new systems perform the same function as the old, and have equal

or better performance characteristics. Installation and operation of

the new radiation monitoring systems do not degrade any active or

passive equipment that responds to an accident.

The proposed increase in the surveillance test interval of the

subject radiation monitoring systems from 12 to 18 months is

consistent with vendor recommendations, and is based on operating

experience with instrumentation of a similar design.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any previously evaluated.

Both modifications replace obsolete radiation monitoring

equipment and have the same failure modes as the existing equipment.

The upgraded systems are considered enhancements to the existing

systems and are considered neither a contributor nor initiator of

any accidents previously evaluated.

Based on the above, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

Neither the accuracy nor the responsiveness of the existing

radiation monitoring equipment will be degraded as a result of the

installation of modifications 5384 and 5386. Revisions to the

calibration and surveillance frequencies are based on vendor

information and experience with instrumentation of similar design.

The changes associated with setpoints and the lower limit of

detection are in the conservative direction. The upgraded main stack

system continues to provide a non-safety related trip signal to

Group III isolation valves during purging of the containment through

the SBGTS [standby gas treatment system]. The revisions to parameter

descriptions and instrument designation are considered

administrative.

Therefore, based on the above, the proposed changes do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-311, Salem

Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

Date of amendment request: February 3, 1994, supplemented September

19, 1994, and November 23, 1994

Description of amendment request: The proposed amendment revises

the Technical Specifications to reflect a reduction in the Reactor

Coolant System flow.

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 analyzed.

No component modification, system realignment, or change in

operations will occur which could affect the probability of any

accident or transient. The proposed reduction in RCS loop and total

flow rates will not change the probability of a challenge to any

Engineered Safeguard Feature or other device. The consequences of

previously analyzed accidents have been found to remain within

acceptable licensing basis limits when the reduced flow rates are

assumed. The system transient response is not affected by the

initial RCS flow assumption, unless the initial assumption is so low

as to impair the steady-state core cooling capability or steam

generator heat transfer capability. This is clearly not the case

with a 1% reduction in RCS flow. The proposed change to the wording

of the parameter title on Table 3.2-1 is editorial for clarity.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

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

accident.

No component modification, system realignment, or change in

operating procedure will occur which could create the possibility of

a new event not previously considered. The proposed reduction in RCS

loop and total flow rates will not initiate any new events.

Therefore, the proposed changes would not create the possibility of

a different or new kind of accident.

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

The proposed decrease in RCS loop and total flow rates has been

analyzed and found to have an insignificant effect on the applicable

transient analyses found in the FSAR. The proposed change to the

wording of the parameter title on Table 3.2-1 is editorial for

clarity. Therefore, the proposed changes would not involve a

significant reduction in any margin of safety.

Therefore, based on the information presented above, PSE&G has

concluded there is 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: January 30, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 4.6.1.2.a and associated Bases for

3/4.6.1.2 to state that Type A tests for overall integrated containment

leakage rate shall be conducted in accordance with the requirements

specified in Appendix J of 10 CFR 50, as modified by NRC-approved

exemptions. Additionally, TS 4.6.1.2.b would be revised to eliminate

the reference to the schedule contained in TS 4.6.1.2.a.

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:

Toledo Edison has reviewed the proposed change and determined

that a significant hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in

accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiators,

[[Page 14029]]

conditions or assumptions are significantly affected by the

proposed changes.

The proposed change would revise Technical Specification (TS)

Surveillance Requirement (SR) 4.6.1.2.a to allow overall integrated

containment leakage rate (Type A) testing to be scheduled in

accordance with 10 CFR 50 Appendix J, as modified by approved

exemptions, and would make associated administrative changes to TS

SR 4.6.1.2.b and to TS Bases 3/4.6.1.2. As stated above, none of

these proposed changes involve accident initiators, conditions, or

assumptions.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because no accident conditions or

assumptions are affected by the proposed changes.

The results of the previous Type A testing demonstrate a high

degree of containment integrity. The Type B and C testing performed

since the last Type A test provides confidence that the high degree

of containment integrity will be maintained during the interval to

the next Type A test. Therefore, the proposed changes do not alter

the source term, containment isolation, or allowable releases, and

will not increase the radiological consequences of a previously

evaluated accident.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new or

different accident initiators or assumptions are introduced by the

proposed changes. The proposed changes do not affect the design or

operation of any plant system, structure, or component. The proposed

changes do not affect any accident initiators and are not initiators

themselves. The proposed changes do not alter any accident

scenarios.

3. Not involve a significant reduction in a margin of safety.

The initial conditions and methodologies used in the accident

analyses remain unchanged. As described above, the proposed changes

do not significantly reduce or adversely affect the confidence that

the present high degree of containment integrity will be 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 request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: Leif J. Norrholm

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of amendment request: January 30, 1995

Description of amendment request: The proposed amendment would

provide new Reactor Coolant Pressure Boundary (RCPB) pressure-

temperature limit curves that are applicable up to 21 effective full

power years (EFPY).

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:

Toledo Edison had reviewed the proposed change and determined

that a significant hazards consideration does not exist because

operation of Davis-Besse Nuclear Power Station, Unit 1, in

accordance with this change would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because: (1) revision of the pressure-

temperature curves and the extended applicability of the pressurizer

level/RCS pressure limit curves for periods when relief valve DH4849

is inoperable will continue to provide the same level of protection

of the RCPB as was previously evaluated, and (2) the revision to

License Condition 2.C(3)(d) is administrative to reflect the

validity of the present analyses to 21 EFPY and (3) the revision to

the Technical Specification Bases

to reflect the extension to 21 EFPY is administrative and does

not affect any previously analyzed accidents.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because: (1) revision of the pressure-

temperature curves and the extended applicability of the pressurizer

level/RCS pressure limit curves for periods when relief valve DH4849

is inoperable will continue to provide the same level of protection

of the RCPB as was previously evaluated, and (2) the revision to

License Condition 2.C(3)(d) is administrative to reflect the

validity of the present analyses to 21 EFPY and (3) the revision to

the Technical Specification Bases to reflect the extension to 21

EFPY is administrative and does not affect any previously analyzed

accidents.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because: (1)

revision of the pressure-temperature curves and the extended

applicability of the pressurizer level/RCS pressure limit curves

will continue to provide protection against reactor vessel failure

due to brittle fracture concerns under all postulated circumstances,

and (2) the revision to License Condition 2.C(3)(d) is

administrative to reflect the validity of the present analyses to 21

EFPY and (3) the revision to the Technical Specification Bases to

reflect the extension to 21 EFPY is an administrative change and

does not affect any activities or equipment in plant operation.

3. Not involve a significant reduction in a margin of safety

because: (1) revision of the pressure-temperature curves and the

extended applicability of the pressurizer level/RCS pressure limit

curves maintains the present margin of safety from reactor vessel

brittle fracture as required by 10 CFR 50, Appendix G, and (2) the

revision to License Condition 2.C(3)(d) and the Bases revision are

administrative and do not affect any analyses which provide the

basis for the Technical Specifications.

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 Toledo Library,

Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: Leif J. Norrholm

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

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

Date of amendment request: February 14, 1995

Description of amendment request: The proposed change revises

Technical Specification 4.4.D to reference the testing requirements of

10 CFR Part 50, Appendix J, and to state that the Nuclear Regulatory

Commission-approved exemptions to the applicable regulatory

requirements are permitted.

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:

Virginia Electric and Power Company has performed an evaluation

of ... the proposed administrative Technical Specification change,

in accordance with 10 CFR 50.91(a)(1) regarding no significant

hazards considerations using the standards in 10 CFR 50.92(c). A

discussion of these standards as they relate to this ... amendment

request follows.

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change ... revises Technical Specification 4.4.D to

reference the testing frequency requirements of 10 CFR 50 Appendix J

and to state that NRC approved exemptions to the applicable

regulatory [[Page 14030]] requirements are permitted. The current

Technical Specification requires retests in accordance with Section

III.D.1(a) of Appendix J. The proposed administrative change simply

includes the statement ``as modified by NRC approved exemptions.''

No new requirements are added, nor are any existing requirements

deleted. Any specific changes to the requirements of Section

III.D.1(a) will require a submittal from Virginia Electric and Power

Company under 10 CFR 50.12 and subsequent review and approval by the

NRC prior to implementation. The proposed change is stated

generically to avoid the need for further Technical Specification

changes if different exemptions are approved in the future.

The proposed change, in itself, does not affect reactor

operations or accident analyses and has no radiological

consequences. The change provides clarification so that future

Technical Specifications changes will not be necessary to correspond

to applicable NRC approved exemptions from the requirements of

Appendix J.

Therefore, this proposed change does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

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

Different Kind of Accident from any Previously Evaluated.

The proposed Technical Specification amendment provides

clarification to a specification that paraphrases a codified

requirement.

Since the proposed change would not change the design,

configuration or method of operation of the plant, it would not

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

any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The proposed Technical Specification change is administrative

and clarifies the relationship between the requirements of TS 4.4.D,

Appendix J, and any approved exemptions to Appendix J. It does not,

in itself, change a safety limit or [a] Limiting Condition for

Operation. The NRC will directly approve any proposed change or

exemption to III.D.1(a) of Appendix J prior to implementation.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

this review, it appears that the three standards of 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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

Williams, Riverfront Plaza, 951 E. Byrd Street, Richmond, Virginia

23219.

NRC Project Director: David B. Matthews

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.

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

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

Illinois

Date of amendment request: February 14, 1995

Brief description of amendment request: The amendment request

proposes changes to Technical Specification 3.8.2, ``AC Sources-

Shutdown;'' 3.8.5, ``DC Sources-Shutdown;'' and 3.8.8, ``Inverters-

Shutdown.'' The proposed changes would revise the operability

requirements for the Division 3 diesel generator and the Division 3 and

4 batteries, battery chargers, and inverters to apply only when the

high pressure core spray system is required to be operable.Date of

publication of individual notice in Federal Register: February 17, 1995

(60 FR 9412).

Expiration date of individual notice: March 20, 1995

Local Public Document Room location: Vespasian Warner Public

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

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

Callaway County, Missouri

Date of amendment request: September 8, 1994

Brief description of amendment request: The amendment request

proposes changes to Technical Specification Section 3/4.9.1 to

establish administrative controls to address a possible boron dilution

event directly from the reactor makeup water system.

Date of publication of individual notice in Federal Register: March

1, 1995 (60 FR 11151).

Expiration date of individual notice: March 31, 1995

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo

Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

Date of application for amendment: November 30, 1994, as

supplemented by letter dated January 27, 1995

Brief description of amendment: The amendment changed the

pressurizer code safety valve lift setting from 2500

[[Page 14031]] psia to 2475 psia. The lift setting is being changed to

permit Unit 2 to operate with up to 1500 plugged tubes in each steam

generator.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 78

Facility Operating License No. NPF-74: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

496) The additional information contained in the January 27, 1995,

supplemental letter was clarifying in nature and thus within the scope

of the initial notice and did not affect the NRC staff's proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated March 1, 1995.No significant hazards consideration comments

received: No.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

Station,Plymouth County, Massachusetts

Date of application for amendment: September 6, 1994

Brief description of amendment: The proposed amendment relocates

the alarms for the drywell to suppression chamber vacuum breaker to a

different annunicator panel.

Date of issuance: February 16, 1995 Effective date: To be

implemented prior to startup from refueling outage 10.

Amendment No.: 158

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

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

53839) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated No significant hazards

consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of application for amendments: January 5, 1994, as

supplemented by letters dated April 26, 1994, September 30, 1994, and

January 12, 1995.

Brief description of amendments: The amendments change the

Braidwood Technical Specifications to remove the requirement to verify,

every 18 months, that the control room ventilation can be manually

isolated.

Date of issuance: February 28, 1995

Effective date: February 28, 1995

Amendment Nos.: 60 and 60

Facility Operating License Nos. NPF-72 and NPF-77: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 25, 1995 (60 FR

4930). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 28, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: Wilmington Township Public

Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.

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

Will County, Illinois

Date of application for amendments: August 31, 1993, as

supplemented July 19, 1994.

Brief description of amendments: The amendments revise the

technical specifications by increasing the allowed outage time for an

inoperable chiller only in MODES 1 through 4, adding an optional ACTION

statement in MODES 5 and 6, and adding a surveillance requirement for

the control room ventilation system.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment Nos.: 70, 70, 61 and 61

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 25, 1995 (60 FR

4932). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 2, 1995.No significant

hazards consideration comments received: No

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

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

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481.Commonwealth Edison Company, Docket

Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3,

Grundy County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, IllinoisDate

of application for amendments: July 29, 1992, as supplemented January

14, 1993, and February 16, 1993

Brief description of amendments: Dresden and Quad Cities Technical

Specification Upgrade Program. Date of issuance: February 16,

1995Effective date: Immediately, to be implemented by December 31,

1995.

Amendment Nos.: 131, 125, 152, and 148

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

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34071) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 16, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: For Dresden, The Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities,

The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

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

Date of application for amendments: July 29, 1992, as supplemented

January 14, 1993, February 16, 1993 and January 27, 1995

Brief description of amendments: The July 29, 1992, application, is

one of twelve applications which have been submitted by Commonwealth

Edison Company (ComEd) in an effort to upgrade the existing custom

Technical Specifications (TS) to the Boiling Water Reactor (BWR)

Standard Technical Specifications (STS). Dresden has recently

rescheduled the Unit 2 refueling outage from March 4, 1995, until June

1995. Currently, the surveillance frequency for certain Inservice

Testing (IST) requirements expires on February 21, 1995. The current

TSs do not make provisions for a grace period for surveillance

frequencies of the IST program. In accordance with BWR STS guidance,

the TSs regarding IST proposed in the July 29, 1992, application, allow

the flexibility to perform these tests appropriately during refueling

outages (where applicable) by providing a 25 percent extension to IST

surveillance intervals. The January 27, 1995, supplement requested the

staff to review and approve just that portion of the July 29, 1992,

application dealing with the implementation of the IST program in

Section 3.0/4.0 of the proposed TS. [[Page 14032]]

Date of issuance: February 22, 1995Effective date: February 22,

1995

Amendment Nos.: 132 and 126

Facility Operating License Nos. DPR-19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 23, 1993 (58 FR

34071) The January 27, 1995, letter did not change the initial proposed

no significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation datedFebruary 22, 1995.No significant hazards consideration

comments received: No

Local Public Document Room location: Morris Public Library, 604

Liberty Street, Morris, Illinois 60450.

Connecticut Yankee Atomic Power Company and Northeast Nuclear

Energy Company, Docket Nos. 50-213 and 50-245, Haddam Neck Plant

and Millstone Nuclear Power Station, Unit 1, Middlesex County and

New London County, Connecticut

Date of application for amendments: October 31, 1994, as

supplemented February 14, 1995.

Brief description of amendments: The amendments renew the existing

license conditions for both plants to implement and maintain Integrated

Implementation Schedule Program Plans (the Program Plans). The Program

Plans provide a methodology to be followed for scheduling plant

modifications and engineering evaluations.

Date of issuance: February 23, 1995

Effective date: February 23, 1995

Amendment Nos.: 183 for Haddam Neck, 80 for Millstone 1

Facility Operating License Nos. DPR-61 and DPR-21. Amendments

revise the Licenses.

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

63117)The February 14, 1995, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated February 23,

1995.No significant hazards consideration comments received: No.

Local Public Document Room locations: Russell Library, 123 Broad

Street, Middletown, CT 06457, for the Haddam Neck Plant, and the

Learning Resource Center, Three Rivers Community-Technical College,

Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for

Millstone Unit 1.

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

Buren County, Michigan

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

February 10, 20, and 22, 1995.

Brief description of amendment: The amendment revises primary

coolant system (PCS) pressure-temperature limits, power-operated relief

valve setting limits, and primary coolant pump starting limits to

accommodate reactor vessel fluence for an additional 4 effective full

power years. The amendment also revises the emergency core cooling

system technical specifications to render two high-pressure safety

injection pumps incapable of injecting into the PCS when the PCS is

below 300 deg.F rather than rendering both inoperable below 260 deg.F.

In addition, it revises the pressurizer heatup to achieve consistency

between design assumptions and technical specifications limits.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment No.: 163

Facility Operating License No. DPR-20. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

501) The February 10, 20, and 22, 1995, submittals provided

clarifyinginformation which was within the scope of the initial

application and did not affect the staff's initial proposed no

significant hazards consideration findings. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

March 2, 1995.No significant hazards consideration comments received:

No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of application for amendments: January 10, 1994, as

supplemented March 21 and September 15, 1994, and January 5, 1995

Brief description of amendments: The amendments revised Technical

Specification Table 2.2-1 and TS 4.2.5 to allow a change in the method

for measuring reactor coolant system (RCS) flow rate from the

calorimetric heat balance method to a method based on a one-time

calibration of the RCS cold leg elbow differential pressure taps.

Date of issuance: February 17, 1995

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

issuance

Amendment Nos.: 128 and 122

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

revised the Technical Specifications.

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

3743) for Unit 1; and March 1, 1994 (59 FR 9785) for Unit 2

The March 21 and September 15, 1994, and January 5, 1995, letters

provided additional information that did not change the initial scope

of the January 10, 1994, application and the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 17, 1995. No significant hazards

consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear

One,Unit No. 1, Pope County, Arkansas

Date of amendment request: August 30, 1994

Brief description of amendment: The amendment revised the Technical

Specifications to address the installation of two battery chargers on

each 125 vdc power train in lieu of the ``swing'' battery charger that

is currently used.

Date of issuance: February 17, 1995

Effective date: February 17, 1995

Amendment No.: 176

Facility Operating License No. DPR-51. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 17, 1995 (60 FR

3439) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 17, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear

One,Unit No. 1, Pope County, Arkansas

Date of amendment request: June 22, 1994.

Brief description of amendment: The amendment extends the allowable

outage time for one inoperable train of emergency feedwater from 36

hours to 72 hours, clarifies the specifications and their associated

bases, and relocates information within the specifications.

Date of issuance: March 1, 1995

Effective date: 30 days following the date of

issuance. [[Page 14033]]

Amendment No.: 177

Facility Operating License No. DPR-51. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994, (59 FR

42339) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 1, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January 19, 1995

Brief description of amendment: The amendment changed the Appendix

A technical specifications (TSs) by adding TS 3.0.5 and its associated

Bases. This new specification will allow equipment removed from service

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

under administrative controls soley to perform testing required to

demonstrate its OPERABILITY or the OPERABILITY of other equipment.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 101

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

Technical Specifications.

Date of initial notice in Federal Register: January 27, 1995 (60 FR

5441) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 1, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 11, 1994, as supplemented by

letter dated December 2, 1994.

Brief description of amendment: The amendment revised the Technical

Specifications for the Waterford Steam Electric Station, Unit 3, by

modifying the specifications having cycle-specific parameter limits by

replacing the values of those limits with a reference to a core

operating limits report for the values of those limits. These changes

are in accordance with the requirements of Generic Letter 88-16.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 102

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

Technical Specifications.

Date of initial notice in Federal Register: December 21, 1994 (59

FR 65812) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 1, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 19, 1994, as supplemented by

letter dated October 14, 1994.

Brief description of amendment: The amendment changed the Appendix

A technical specification (TSs) by removing the Limiting Condition For

Operation (LCO) 3/4.3.4, the associated surveillance requirements, and

Bases information from the TSs. This information and requirements will

be incorporated into the Waterford 3 Updated Final Safety Analysis

Report (UFSAR) and maintained under the provisions of 10 CFR 50.59.

Date of issuance: March 2, 1995

Effective date: March 2, 1995

Amendment No.: 103

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

Technical Specifications.

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

45023) The additional information contained in the supplemental letter

dated October 14, 1994, was clarifying in nature and thus, within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated March 2, 1995.No significant hazards consideration comments

received: No.

Local Public Document Room location: University of New Orleans

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

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

Mississippi Electric Power Association, and Mississippi Power &

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

1, Claiborne County, Mississippi

Date of application for amendment: April 21, 1993

Brief description of amendment: The amendment revised the

requirement for control rod testing to increase the ``notch testing''

surveillance interval for partially withdrawn control rods from once

per 7 days to once per 31 days. The change is consistent with the

format and content of the Improved Standard Technical Specifications

(NUREG-1434, Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 115

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

Technical Specifications.

Date of initial notice in Federal Register: May 12, 1993 (58 FR

28055) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 16, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

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

Mississippi Electric Power Association, and Mississippi Power &

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

1, Claiborne County, Mississippi

Date of application for amendment: July 14, 1993

Brief description of amendment: The amendment revised technical

specification requirements for the hydrogen ignition system (HIS). The

amendment also removed several tables related to the HIS in accordance

with guidance contained in Generic Letter 91-08, ``Removal of Component

Lists From Technical Specifications.''

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 116

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

Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46232) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 16, 1995. No

significant hazards consideration comments received: No [[Page 14034]]

Local Public Document Room location: Judge George W. Armstrong

Library, 220 S. Commerce at Washington, Natchez, Mississippi 39120.

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

Mississippi Electric Power Association, and Mississippi Power &

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

1, Claiborne County, Mississippi

Date of application for amendment: August 11, 1993

Brief description of amendment: The amendment deleted the

requirements of Limiting Condition for Operation (LCO) 3.3.3.9 and

Surveillance Requirement 4.3.3.9 related to loose-part detection

instrumentation. The deleted requirements will be relocated to

documents that are controlled by the licensee under the provisions of

10 CFR 50.59. The change is consistent with the format and content of

the Improved Standard Technical Specifications (NUREG-1434, Revision

0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 117

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

Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46232) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 16, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

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

Mississippi Electric Power Association, and Mississippi Power &

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

1, Claiborne County, Mississippi

Date of application for amendment: August 11, 1993

Brief description of amendment: The amendment deleted certain

accident monitoring instruments from Technical Specification Table

3.3.7.5-1 ``Accident Monitoring Instrumentation'' and deleted the

corresponding Surveillance Requirements from Table 4.3.7.5-1,

``Accident Monitoring Instrumentation Surveillance Requirements.'' The

deleted requirements will be relocated to documents that are controlled

by the licensee under the provisions of 10 CFR 50.59. The change is

consistent with the format and content of the Improved Standard

Technical Specifications (NUREG-1434, Revision 0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 118

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

Technical Specifications.

Date of initial notice in Federal Register: September 1, 1993 (58

FR 46234) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 16, 1995.No significant

hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

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

Mississippi Electric Power Association, and Mississippi Power &

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

1, Claiborne County, Mississippi

Date of application for amendment: October 22, 1993, as

supplemented by letters dated February 10, and 14, 1995.

Brief description of amendment: The amendment modified the testing

frequencies for the drywell bypass test and the airlock test, relocated

certain drywell airlock tests from the technical specifications to

administrative procedures, and incorporates various improvements from

the Improved Standard Technical Specifications (NUREG-1434, Revision

0).

Date of issuance: February 16, 1995

Effective date: February 16, 1995

Amendment No: 119

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

Technical Specifications.

Date of initial notice in Federal Register: December 8, 1993 (58 FR

64607) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 16, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: Judge George W. Armstrong

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

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

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

Date of application for amendment: May 17, 1993, as supplemented on

December 23, 1994

Brief description of amendment: The amendment changes the action

statement for inoperable degraded grid and loss of voltage relays and

their associated auxiliary relays and timers.

Date of issuance: January 31, 1995

Effective date: January 31, 1995

Amendment No.: 193

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

FR 59750). The December 23, 1994, letter provided additional

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of this amendment is contained in a Safety Evaluation dated

January 31, 1995. No significant hazards consideration comments

received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, PA 17105. The above Notice was to be

published in the Federal Register of February 15, 1995. The notice that

was inadvertently published at 60 FR 8762 relates to a licensing action

which has not been completed.

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

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

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

Matagorda County, Texas

Date of amendment request: November 7, 1994, as supplemented by

letters dated December 20, 1994, and January 23, 1995.

Brief description of amendments: The amendments changed the number

of standby diesel generators (SDGs) (emergency power source) required

to be operable during Mode 6 with greater than or equal to 23 feet of

water above the reactor vessel flange, from two to one. The amendment

also allows limited substitution of an alternate onsite emergency power

source for one of the two required SDGs, in Mode 5, and in Mode 6 with

less than 23 feet of water. In addition, certain system specifications

that are affected by the changes for the emergency power source were

also changed.

Date of issuance: February 14, 1995

Effective date: February 14, 1995, to be implemented within 31 days

of issuance.

Amendment Nos.: Unit 1 - Amendment No. 34; Unit 2 - Amendment No.

20

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

revised the Technical [[Page 14035]] Specifications.Public comments

requested as to proposed no significant hazards consideration: Yes (60

FR 5739, dated January 30, 1995). The notice provided an opportunity to

submit comments on the Commission's proposed no significant hazards

consideration determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by March 1, 1995,

but stated that, if the Commission makes a final no significant hazards

consideration determination, any such hearing would take place after

issuance of the amendments.

The Commission's related evaluation of the amendments, finding of

exigent circumstances, and final determination of no significant

hazards consideration is contained in a Safety Evaluation dated

February 14, 1995.

Local Public Document Room location: Wharton County Junior College,

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

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

Linn County, Iowa

Date of application for amendment: August 15, 1994, as supplemented

on December 21, 1994, and January 20, 1995. The licensee's submittals

of December 21, 1994, and January 20, 1995, provided clarification and

did not change the original no significant hazards consideration.

Brief description of amendment: The proposed amendment would revise

the Technical Specifications by increasing the allowable main steam

isolation valve (MSIV) leakage and deleting the requirements applicable

to the MSIV leakage control system.

Date of issuance: February 22, 1995

Effective date: February 22, 1995 and to be implemented within 90

days.

Amendment No.: 207

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

Technical Specifications.

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

FR 47169) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 22, 1995.No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

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

Illinois

Date of application for amendment: August 12, 1994, as supplemented

on October 14, 1994 and February 6, 1995.

Brief description of amendment: The amendment modifies Clinton

Power Station Technical Specification 3.6.5.1, ``Drywell,'' to permit a

one-time only change to forego performance of the drywell bypass

leakage rate test during the fifth refueling outage scheduled to begin

in March 1995.

Date of issuance: March 1, 1995

Effective date: March 1, 1995

Amendment No.: 96

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

Technical Specifications.

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

FR 49427). The October 14, 1994, and February 6, 1995, submittals

consisted of revisions and clarifications which did not change the

staff's initial proposed no significant hazards consideration

determination or expand the scope of the original notice.The

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

Safety Evaluation dated March 1, 1995. No significant hazards

consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

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

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

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

Michigan

Date of application for amendments: Novembe

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