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

Federal RegisterJul 30, 1997

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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.

[[Page 40844]]

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 3, 1997, through July 18, 1997. The

last biweekly notice was published on July 16, 1997.

Notice of Consideration of Issuance of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By August 29, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

be filed with

[[Page 40845]]

the Secretary of the Commission, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention: Docketing and Services Branch, or

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

Building, 2120 L Street, NW., Washington DC, by the above date. A copy

of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

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

Date of amendments request: May 23, 1997

Description of amendments request: The proposed amendment would

revise Technical Specification 3/4.4.4 to allow the installation of

ABB/CE welded sleeves, in accordance with ABB/CE Topical Report CEN-

630-P, ``Repair of 3/4 Inch Outer Diameter Steam Generator Tubes Using

Leak Tight Sleeves,'' Revision 1, in the Palo Verde Units 1, 2 and 3

steam generators.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below: 1. The proposed change does

not involve a significant increase in the probability or consequences

of an accident previously evaluated.

The proposed amendment to permit the use of steam generator tube

sleeves as an alternative to tube plugging is a safe and effective

repair procedure that does not result in removing a tube from

service. Mechanical strength, corrosion resistance, installation

methods, and inservice inspection techniques of sleeves have been

shown to meet NRC acceptance criteria.

Analytical verifications were performed using design and

operating transient parameters selected to envelope loads imposed

during normal operating and accident conditions. Fatigue and stress

analysis of sleeved tube assemblies were completed in accordance

with the requirements of Section III of the ASME Code. The results

of qualification testing, analysis and plant operating experience at

other facilities demonstrates that the sleeving process is an

acceptable means of maintaining steam generator tube integrity. The

sleeve configuration has been designed and analyzed in accordance

with the structural margins specified in Regulatory Guide 1.121 (RG

1.121). Furthermore, the installed sleeve will be monitored through

periodic inspections on a sample basis with eddy current techniques.

A sleeve-specific plugging margin, per the recommendations of

Regulatory Guide 1.121, has been specified with appropriate

allowances for NDE uncertainty and defect growth rate. Therefore,

since the sleeve provides the same protection against a tube rupture

as the original tube, the use of sleeves does not involve a

significant increase in the probability of an accident previously

evaluated.

Recently, industry experience with forced shutdown events

associated with tube failures at sleeve junctions was assessed by

APS and ABB-CE. The root cause of these events has been attributed

to the lack of proper post-installation stress relief and/or the

imposition of high stresses due to tube growth restrictions at

locked tube supports. The material and design of the PVNGS steam

generator supports minimizes the potential for locked supports. The

tube supports are of eggcrate design and are constructed of ferric

stainless steel. The large flow area in the eggcrate design provides

better irrigation and reduces the potential for steam blanketing,

therefore, the tube-to-tube support crevices are less likely to be

blocked by crud, boiler water deposits and corrosion products. Since

the support material is type 409 ferric stainless steel, it is not

susceptible to magnetite corrosion which has resulted in denting and

lockup at plants with carbon steel supports. These conclusions have

been substantiated via tube pull activities conducted in PVNGS Unit

2. Although ABB/CE does not require post-weld heat treatment in all

applications, APS will require that a post-weld stress relief be

conducted for sleeve installations. Therefore, with proper sleeve

installation the proposed change will not involve a significant

increase in the probability of an accident previously evaluated.

The consequences of accidents previously analyzed are not

increased as a result of sleeving activities. The hypothetical

failure of the sleeve would be bounded by the current steam

generator tube rupture analysis contained in the PVNGS UFSAR. Due to

the slight reduction in diameter caused by the sleeve wall

thickness, it is expected that the primary release rates would be

less than assumed for the steam generator tube rupture analysis,

and, therefore, would result in lower primary fluid mass release to

the secondary system. Additionally, further conservatism is

introduced if the break were postulated to occur at a location on

the tube higher than the location where a sleeve is installed. The

overall effect would be reduced steam generator tube rupture release

rates. The minimal reduction in flow area associated with a tube

sleeve has no significant affect on steam generator performance with

respect to heat transfer or system flow resistance and pressure

drop. The installation of sleeves rather than plugging also

maintains a greater heat transfer surface in the steam generator. In

any case, the impacts are bounded by evaluations which demonstrate

the acceptability of tube plugging, which totally removes the tube

from service.

Therefore, in comparison to plugging, tube sleeving is

considered a significant improvement with respect to steam generator

performance. Therefore, based on the above, the proposed amendment

does not significantly increase the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

A sleeved steam generator tube performs the same function in the

same passive manner as an unsleeved steam generator tube. Tube

sleeves are designed and qualified to the stress and pressure limits

of Section III of the ASME Code and Regulatory Guide 1.121.

The installation of the sleeve, including weld and welder

qualification and nondestructive examination (NDE), meets or exceeds

the requirements of ASME Section XI. Three types of NDE are

conducted. Ultrasonic Testing (UT) is performed to verify the

adequacy of the tube to sleeve weld assuring proper fusion. Eddy

Current testing (ECT) is performed following each installation to

establish baseline data for each sleeve in order to monitor future

degradation of the primary to secondary pressure boundary. Visual

inspections will be performed to verify or ascertain the mechanical

and structural condition of a weld. Critical conditions which are

checked include weld width and completeness, and the absence of

visibly noticeable indications such as cracks, pits, and burn

through.

ABB Combustion Engineering, Inc., Report CEN-630-P, Revision 01,

``Repair of 3/4'' O.D. Steam Generator Tubes Using Leak Tight

Sleeves'' dated November, 1996, demonstrates that the repair of

degraded steam generator tubes using tube sleeves will result in

tube bundle integrity consistent with the original design basis.

Extensive analyses and testing have been performed on the sleeve and

sleeve to tube joints to demonstrate that the design criteria are

met. The proposed amendments have no significant effect on the

configuration of the plant, and the change does not affect the way

in which the plant is operated. Therefore, reactor operation with

sleeves installed in the steam generator tubes does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

[[Page 40846]]

Evaluation of the sleeved tubes indicates no detrimental effects

on the sleeve-tube assembly resulting from reactor coolant system

flow, coolant chemistries, or thermal and pressure conditions.

Structural analyses have been performed for sleeves which span the

tube at the top of the tube sheet and which span the flow

distribution plate or eggcrate support. Mechanical testing has been

performed to support the analyses. Corrosion testing of typical

sleeve-tube assemblies has been completed and reveals no evidence of

sleeve or tube corrosion considered detrimental under anticipated

service conditions.

Steam generator tube integrity is maintained under the same

limits for sleeved tubes as for unsleeved tubes, ie., Section III of

the ASME Code and Regulatory Guide 1.121. The portions of the

installed sleeve assembly which represents the reactor coolant

pressure boundary can be monitored for the initiation and

progression of sleeve/tube wall degradation, thus satisfying the

requirements of Regulatory Guide 1.83. The degradation limit at

which a sleeve/tube boundary is considered inoperable has been

analyzed in accordance with Regulatory Guide 1.121 and is specified

in the proposed amendment. Eddy current detectability of flaws has

been verified by ABB Combustion Engineering. Additionally, the

Technical Specifications continue to require monitoring and

restriction of primary- to- secondary system leakage through the

steam generators. The minimal reduction in RCS flow due to sleeving

results in an insignificant impact on RCS operation during normal or

accident conditions and is bounded by tube plugging evaluations.

Based upon the testing and analyses performed, the installation

of tube sleeves will not result in a significant reduction in a

margin of safety.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

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

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

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: William H. Bateman

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

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendments request: 3April 30, 1997

Description of amendments request: The proposed amendments would

revise Surveillance Requirements (SRs) 4.7.2.b.2 and 4.7.2.c in the

Technical Specifications for the Brunswick Steam Electric Plant, Units

1 and 2. These SRs require periodic testing of the control room

emergency ventilation system charcoal filters. The proposed amendments

would revise the temperature and relative humidity conditions under

which the testing is performed. The revised conditions were selected to

approximate operating or accident conditions. Testing at the revised

conditions is more conservative than testing at the currently required

conditions. Additionally, the proposed amendments would relax the

acceptance criterion for filtration efficiency from 95% to a value

corresponding to a filtration efficiency of 90%. The 90% value is the

filtration efficiency assumed in the current bounding calculations for

control room dose under accident conditions.

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 amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendments revise Surveillance Requirements

4.7.2.b.2 and 4.7.2.c to require testing of the control room

emergency ventilation system (CREVS) charcoal in accordance with

ASTM D3803-1989, ``Standard Test Method for Nuclear-Grade Activated

Carbon.'' Currently, Surveillance Requirements 4.7.2.b.2 and 4.7.2.c

to [sic] require testing in accordance with the criteria of

Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1,

1976. The purpose of the CREVS is to mitigate an accident. It is not

associated with any initiating events and, therefore, cannot affect

the probability of any accident.

ASTM D3803-1989 is an industry accepted standard for charcoal

filter testing. The conditions employed by this standard were

selected to approximate operating or accident conditions of a

nuclear reactor which would severely reduce the performance of

activated carbons. The ASTM D3803-1989 testing is more stringent

than that required by the criteria of Regulatory Position C.6.a of

Regulatory Guide 1.52, Revision 1, 1976. Specifically, the testing

temperature of ASTM D3803-1989 is 30.0 [plus or minus] 0.2 deg.C

versus 80 deg.C for the Regulatory Guide 1.52 testing. Also, ASTM

D3803-1989 requires a relative humidity of 93 to 96% versus [greater

than or equal to] 70% for the Regulatory Guide 1.52 testing. Both

these parameters result in the ASTM D3803-1989 test being a more

conservative test [than] that required by the criteria of Regulatory

Position C.6.a of Regulatory Guide 1.52, Revision 1, 1976.

The proposed changes to Surveillance Requirements 4.7.2.b.2 and

4.7.2.c require that charcoal samples tested in accordance with the

methodology of ASTM D3803-1989 meet the acceptance criteria of , instead of Note

is an administrative change and thus will not

change the probability or consequences of an accident.

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

from any accident previously evaluated because:

The changes to the definition of Channel Calibration and

correction of the other miscellaneous errors in the TS and TS Bases

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

accident, because the changes will not affect the design or

operation of any structure, system, or component in the plant.

3) Involve a significant reduction in the margin of safety

because:

a. The definition of Channel Calibration is being changed to be

like the definition in NUREG 1434, Standard Technical Specifications

General Electric Plants, BWR/6, Revision 1. The primary changes

involve requiring only an inplace qualitative assessment of

thermocouple and RTD sensors. These sensors are not adjustable and

not susceptible to setpoint drift. Thus the appropriate check of the

sensors is a qualitative assessment only. The inplace qualitative

assessment assures operability of the sensors. Therefore there is no

reduction in the margin of safety.

b. The remaining miscellaneous changes are corrections due to

errors in the TS. The corrections will make the associated TS

consistent with the design and licensing basis of LaSalle or correct

typographical errors. 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: July 17, 1996, as supplemented by

letters dated June 3, and July 7, 1997.

Description of amendment request: The proposed change request

modifies Waterford Steam Electric Station, Unit 3, Technical

Specifications (TSs) 3/4.7.1.3, ``CONDENSATE STORAGE POOL,'' by

increasing the minimum Condensate Storage Pool (CSP) level from 82

percent to 91 percent in Modes 1, 2, and 3. The July 7, 1997,

supplement proposes to expand the applicability of TS 3.7.1.3 to

include Mode 4 operational requirements and maintains the 91 percent

minimum CSP level previously requested for Modes 1, 2, and 3. The staff

previously issued No Significant Hazard Considerations notice on March

26, 1997 (62 FR 14461).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?Response: No.

Increasing the minimum required Condensate Storage Pool (CSP)

level to 91 percent will insure that the minimum required 170,000

gallons of water is available to supply the Emergency Feedwater

System and that 3,500 gallons of water is available for use by the

Component Cooling Water Makeup System in Modes 1, 2, and 3.

Maintaining a minimum required CSP level of 11 percent will insure

that 3,500 gallons of water is available for use by the Component

Cooling Water Makeup System in Mode 4. Maintaining the minimum

required water volume will not increase the probability of any

accident previously evaluated. Additionally, it will not affect the

consequences of any accident. Maintaining a minimum required CSP

level will ensure that the system remains within the bounds of the

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

significant increase in the probability or consequences of any

accident previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

Increasing the minimum water volume of the CSP from 82 percent

to 91 percent in Modes 1, 2, and 3 does not create a possibility for

a new or different kind of accident. Maintaining a minimum water

volume of the CSP at 11 percent in Mode 4 does not create a

possibility for a new or different kind of accident. The CSP will be

operated in the same manner as previously evaluated. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

Operation in accordance with this proposed change will ensure

that the minimum contained water volume of the CSP will remain

adequate under all conditions. This will improve the present margin

of safety. Therefore, the proposed change will not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: James W. Clifford, Acting Director

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

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

Date of amendment request: May 29, 1997

Description of amendment request: The proposed amendments will

improve consistency throughout the Technical Specifications and their

related Bases by removing outdated material, incorporating minor

changes in text, making editorial corrections, and resolving other

inconsistencies identified by the licensee's plant operations staff.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments consist of administrative changes to the

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

amendments will implement minor changes in text to rectify

reference, typographic, spelling, and/or consistency-in-format

errors; update the TS Bases; and/or otherwise improve consistency

within the TS for each unit. The proposed amendments do not involve

changes to the configuration or method of operation of

plantequipment that is used to mitigate the consequences of an

accident, nor do the changes otherwise affect the initial conditions

or conservatisms assumed in any of the plant accident analyses.

Therefore, operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

[[Page 40850]]

The proposed administrative revisions will not change the

physical plant or the modes of plant operation defined in the

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

addition or modification of equipment nor do they alter the design

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

in accordance with the proposed amendments would not create the

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

previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendments are administrative in nature and do not

change the basis for any technical specification that is related to

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

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

proposed amendments would not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: December 2, 1996

Description of amendment request: The proposed amendment would

relocate the audit frequency requirements from the plant Technical

Specifications to the Quality Assurance Plan. In addition, the maximum

interval between certain types of audits will be extended. This change

would make the TMI-2 technical specifications consistent with the

Technical Specifications for Three Mile Island, Unit 1.

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:

10 CFR 50.92 provides the criteria which the Commission uses to

perform a No Significant Hazards Consideration. 10 CFR 50.92 states

that an amendment to a facility license involves No Significant

Hazards if operation of the facility in accordance with the proposed

amendment would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, or

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

from any accident previously evaluated, or

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

The proposed change to the technical specifications is

administrative and does not involve any physical changes to the

facility. No changes are made to operating limits or parameters, nor

to any surveillance activities. Based on this, GPU Nuclear has

concluded that the proposed change does not:

1. Involve a significant increase in the probability of

occurrence of the consequences of an accident previously evaluated.

The proposed amendment is administrative and does not affect the

function of any system or component. Therefore this change does not

increase the probability of occurrence or the consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed change is administrative and no new failure modes

or potential accident scenarios are created.

3. Involve a change in the margin of safety.

This change is administrative in nature and does not affect any

safety settings, equipment, or operational parameters.

Based on the above analysis it is concluded that the proposed

changes involve no significant safety hazards considerations as

defined by 10 CFR 50.92.

The NRC staff has reviewed the analysis of the licensee and, based

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

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

NRC Project Acting Director: Marvin M. Mendonca

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

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

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

Matagorda County, Texas

Date of amendment request: July 16, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification Table 2.2-1 and 3/4.2.5 to allow the

reactor coolant system total flow to be determined using cold leg elbow

tap differential pressure measurements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10[]CFR[]50.92 each application for amendment to an

operating license must be reviewed to determine if the proposed

change involves a Significant Hazards Consideration. The amendment,

as defined below, describing the Technical Specification change

associated with the change has been reviewed and determined to not

involve Significant Hazards Considerations. The basis for this

determination follows.

Proposed Change: The current Technical Specification Table 2.2-1

(page 2-4) ``Reactor Trip System Instrumentation Trip Setpoints,''

provides the Trip Setpoint and Allowable Value for the RCS [reactor

coolant system] Flow-Low trip. The Allowable Value will be changed

to reflect the increased uncertainty associated with the correlation

of the elbow taps to a previous baseline calorimetric. In addition,

Technical Specification 3.2.5 (page 3/4.2-11), ``Power Distribution

Limits, DNB Parameters'', will be changed to allow the RCS total

flow to be measured by the elbow tap [delta]p method. These changes

will include the modification of surveillance requirement 4.2.5.3,

which currently requires performance of a precision heat balance

every 18 months, to allow use of the elbow tap [delta]p method for

RCS flow measurement. Appropriate Technical Specification Bases

sections will also be revised to reflect use of the elbow tap

[delta]p method for flow measurement and to provide clarification.

The revised Technical Specifications are in Appendix C.

Background: The 18-month total RCS flow surveillance is

typically satisfied by a secondary power calorimetric-based RCS flow

measurement. In recent cycles, South Texas Project has experienced

apparent decreases in flow rates which have been attributed to

variations in hot leg streaming effects. These effects directly

impact the hot leg temperatures used in the precision calorimetric,

resulting in the calculation of low RCS flow rates. The apparent

flow reduction has become more pronounced in fuel cycles which have

implemented aggressive low leakage loading patterns. Evidence that

the flow reduction was apparent, but not actual, was provided by

elbow tap measurements. The results of this evaluation, including a

detailed description of the hot leg streaming phenomenon, are

documented in Westinghouse report SAE/FSE-TGX/THX-0152, ``RCS Flow

Verification Using Elbow Taps.''

South Texas Project intends to begin using an alternate method

of measuring RCS flow using the elbow tap [delta]p measurements. For

this alternate method, the RCS elbow tap measurements are correlated

to precision

[[Page 40851]]

calorimetric measurements performed during earlier cycles which

decreased the effects of hot leg streaming.

The purpose of this evaluation is to assess the impact of using

the elbow tap [delta]p measurements as an alternate method for

performing the 18-month RCS flow surveillance on the licensing basis

and demonstrate that it will not adversely affect the subsequent

safe operation of the plant. This evaluation supports the conclusion

that implementation of the elbow tap [delta]p measurement as an

alternate method of determining RCS total flow rate does not

represent a significant hazards consideration as defined in

10[]CFR[]50.92.

Evaluation: Use of the elbow tap [delta]p method to determine

RCS total flow requires that the [delta]p measurements for the

present cycle be correlated to the precision calorimetric flow

measurement which was performed during the baseline cycle(s). A

calculation has been performed to determine the uncertainty in the

RCS total flow using this method. This calculation includes the

uncertainty associated with the RCS flow baseline calorimetric

measurement, as well as uncertainties associated with [delta]p

transmitters and indication via QDPS [qualified display processing

system] or the plant process computer. The uncertainty calculation

performed for this method of flow measurement is consistent with the

methodology recommended by the Nuclear Regulatory Commission (NUREG/

CR-3659, PNL-4973, 2/85). The only significant difference is the

assumption of correlation to a previously performed RCS flow

calorimetric. However, this has been accounted for by the addition

of instrument uncertainties previously considered to be zeroed out

by the assumption of normalization to a calorimetric performed each

cycle. Based on these calculations, the uncertainty on the RCS flow

measurement using the elbow tap method is 2.6% flow which results in

a minimum RCS total flow of 391,500 gpm and must be measured via

indication with QDPS or the plant process computer at approximately

100% power.

The specific calculations performed were for Precision RCS Flow

Calorimetrics for the specified baseline cycles, Indicated RCS Flow

(either QDPS or the plant process computer), and the Reactor Coolant

Flow - Low reactor trip. The calculations for Indicated RCS Flow and

Reactor Coolant Flow - Low reactor trip reflect correlation of the

elbow taps to baseline precision RCS Flow Calorimetrics. As

discussed above, additional instrument uncertainties were included

for this correlation.

The uncertainty associated with the RCS Flow - Low trip

increased slightly. It was determined that due to the availability

of margin in the uncertainty calculation, no change was necessary to

either the Trip Setpoint (91.8% flow) or to the current Safety

Analysis Limit (87% flow) to accommodate this increase. The

Allowable Value is to be modified to allow for the increased

instrument uncertainties associated with the [delta]p to flow

correlation.

Since the flow uncertainty did not increase over the currently

analyzed value, no additional evaluations of the reactor core safety

limits must be performed. In addition, it was determined that the

current Minimum Measured Flow (MMF) assumed in the safety analyses

(389,200 gpm) bounds the required MMF calculated for the elbow tap

method (391,500 gpm).

Based on these evaluations, the proposed change would not

invalidate the conclusions presented in the UFSAR [Updated Final

Safety Analysis Report].

1. Does the proposed modification involve a significant increase

in the probability or consequences of an accident previously

evaluated?

Sufficient margin exists to account for all reasonable

instrument uncertainties; therefore, no changes to installed

equipment or hardware in the plant are required, thus the

probability of an accident occurring remains unchanged.

The initial conditions for all accident scenarios modeled are

the same and the conditions at the time of trip, as modeled in the

various safety analyses, are the same. Therefore, the consequences

of an accident will be the same as those previously analyzed.

2. Does the proposed modification create the possibility of a

new or different kind of accident from any accident previously

evaluated?

The proposed change revises the method for RCS flow measurement,

and therefore does not introduce any new accident indicators or

failure mechanisms.

No new accident scenarios have been identified. Operation of the

plant will be consistent with that previously modeled, i.e., the

time of reactor trip in the various safety analyses is the same,

thus plant response will be the same and will not introduce any

different accident scenarios that have not been evaluated.

3. Does the proposed modification involve a significant

reduction in a margin of safety[?]

There are no changes to the Safety Analysis assumptions.

Therefore, the margin of safety will remain the same.

The proposed change does not impact the results from any

accidents analyzed in the safety analysis.

Conclusion: Based on the preceding information, it has been

determined that this proposed change to allow an alternate RCS total

flow measurement based on elbow tap [delta]p measurements does not

involve a Significant Hazards Consideration as defined by 10 CFR

50.92(c).

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: James W. Clifford, Acting

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

Point Nuclear Station, Unit 1, Oswego County, New York

Date of amendment request: July 2, 1997

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 3/4.2.3 regarding reactor coolant

chemistry in accordance with a report by Electrical Power Research

Institute, Inc. (EPRI) TR-103515-R1, ``BWR Water Chemistry Guidelines,

1996 Revision,'' also known as Boiling Water Reactor Vessel and

Internals Project (BWRVIP)-29. Specifically, the amendment would define

new conductivity limits in TS 3.2.3a (when reactor coolant is 200

degrees F or more and reactor thermal power is no more that 10%), and

in TS 3.2.3b (when reactor thermal power exceeds 10%). The new

conductivity limits would be 1 micro-mho/cm, which is less than the

existing limits of 2 micro-mho/cm and 5 micro-mho/cm. The chloride ion

limit in TS 3.2.3a, 0.1 ppm, would remain at this value but would be

designated as 100 ppb. The chloride ion limit in TS 3.2.3b would be

changed from 0.2 ppm to 20 ppb. Sulfate ion limits would be added to TS

3.2.3a and TS 3.2.3b at 100 ppb and 20 ppb, respectively. In TS 3.2.3c,

the maximum conductivity limit would be changed from 10 micro-mho/cm to

5 micro mho/cm when reactor coolant temperature is 200 degrees F or

more; the maximum chloride ion concentration limit would be changed

from 0.5 ppm to 100 ppb (when reactor thermal power exceeds 10%) and

200 ppb (when reactor coolant temperature is 200 degrees F or more and

reactor thermal power is no more than 10%); and the maximum sulfate ion

concentration of 100 ppb (when reactor thermal power exceeds 10%) and

200 ppb (when reactor coolant temperature is 200 degrees F or more and

reactor thermal power is no more than 10%) would be added. The

requirement to place the reactor in the cold shutdown condition as

currently specified in TS 3.2.3d (when TSs 3.2.2a, b, and c are not

met) and TS 3.2.3e (when the continuous conductivity monitor is

inoperable for more than 7 days) would be changed to require that the

reactor coolant temperature be reduced to below 200 degrees F. TS 4.2.3

would be revised to add that the samples taken and analyzed for

conductivity and chloride ion content are also to be analyzed for

sulfate ion content. TS Bases 3/4.2.3 would also be changed to

[[Page 40852]]

reflect that the purpose of TS 3/4.2.3 is to limit crack growth rates

to values consistent with Unit 1 core shroud analyses in accordance

with an NRC letter dated May 8, 1997.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the proposed amendment, will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The changes to the conductivity and chloride ion action levels

and the addition of sulfate ion levels as an action level in reactor

water chemistry are being made to make the TS and its Bases

consistent with the values used in the core shroud vertical weld

cracking evaluations. These new values reflect the BWR water

chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29)

and are equal to or more restrictive than the present TS values. No

physical modification of the plant is involved and no changes to the

methods in which plant systems are operated are required. None of

the precursors of previously evaluated accidents are affected and

therefore, the probability of an accident previously evaluated is

not increased. These changes to the coolant chemistry TS are more

restrictive limits and no new failure modes are introduced.

Therefore, these changes will not involve a significant increase in

the consequences of an accident previously evaluated.

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

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

different kind of accident from any accident previously evaluated.

The changes to the conductivity and chloride ion action levels

and the addition of sulfate ion levels as an action level in reactor

water chemistry are being made to make the TS and its Bases

consistent with the values use in the core shroud vertical weld

cracking evaluations. The new values reflect the BWR water chemistry

guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29) and are

equal to or more restrictive than the present TS values. No physical

modification of the plant is involved and no changes to the methods

in which plant systems are operated are required. The change does

not introduce any new failure modes or conditions that may create a

new or different accident. Therefore, this change does not create

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

accident] previously evaluated.

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

the proposed amendment, will not involve a significant reduction in

a margin of safety.

The changes to the conductivity and chloride ion action levels

and the addition of sulfate ion levels as an action level in reactor

water chemistry are being made to make the TS and its Bases

consistent with the values used in the core shroud vertical weld

cracking evaluations. These new values reflect the BWR water

chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29)

and are equal to or more restrictive than the present TS values. No

physical modification of the plant is involved and no changes to the

methods in which plant systems are operated are required. This

change does not adversely affect any physical barrier to the release

of radiation to plant personnel or the public. Therefore, the change

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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126

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

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

NRC Project Director: Alex Dromerick, Acting

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

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

County, Connecticut

Date of amendment request: June 19, 1997

Description of amendment request: Technical Specification Table

2.2-1 Notes 1 and 3 define the values for the constants used in the

Overtemperature Delta-T and Overpower Delta-T reactor trip system

instrumentation setpoint calculators. The proposed amendment would make

changes to the notes as well as the associated Bases section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

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

significant hazards consideration (SHC). The basis for this

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

satisfied. The proposed revision does not involve an SHC because the

revision would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to Technical Specification Table 2.2-1

Notes 1 and 3 for the addition of the inequalities ensure that the

constants used for [Overtemperature Delta-T] and [Overpower Delta-T]

will be set conservatively with respect to the assumptions in the

accident analysis. The effect on the turbine

runback function has been evaluated with respect to the Loss of

External Electrical Load And/Or Turbine Trip analysis and it has

been determined that this change does not increase the probability

of this transient. The change was also reviewed to determine if it

produced an increase in the probability of an unnecessary or

spurious reactor trip and it was determined that it did not. This

change does not increase the probability of any previously evaluated

accident.

The consequences of previously evaluated accidents, including

Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power, Rod

Cluster Control Assembly Misalignment, Uncontrolled Boron Dilution,

Loss of External Electrical Load And/Or Turbine Trip, Excessive Heat

Removal Due To Feedwater System Malfunctions, Excessive Load

Increase Incident, Accidental Depressurization Of The Reactor

Coolant System, Accidental Depressurization Of The Main Steam

System, Loss of Reactor Coolant From Small Ruptured Pipes Or From

Cracks In Large Pipes Which Actuate ECCS [emergency core cooling

system], or Major Secondary System Pipe Ruptures have not changed.

The administrative changes have no impact on the design or

operation of Millstone Unit 3.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification Table 2.2-1

Notes 1 and 3 do not alter the design, construction, operation,

maintenance or method of testing of equipment. The proposed changes

alter the Technical Specification description of [an]

[Overtemperature Delta-T] and [Overpower Delta-T] setpoint functions

and requires only slight changes to the actual setpoints in the

field. The [Overtemperature Delta-T] and [Overpower Delta-T]

functions serve to mitigate the effects of accidents by opening the

Reactor Trip breakers or reduce power by ``running back'' turbine

electrical load. The change does not create any new interfaces to

plant control or protection systems and therefore, no new mechanism

for accident initiation has been introduced. The proposed change

does not introduce the possibility of an accident of a different

type than previously evaluated.

Therefore, the proposed revision does not create the possibility

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

evaluated.

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

The proposed changes to Technical Specification Table 2.2-1

Notes 1 and 3 do not affect the integrity of any physical fission

protective boundaries, increase the delays in actuation of safety

systems beyond that assumed in the safety analysis or reduce the

[[Page 40853]]

margin of safety of any system. These changes ensure that actuation

of Overtemperature [Delta-T] and Overpower [Delta-T] reactor trips

will occur conservatively with respect to the assumptions of the

accident analysis.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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 Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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

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

County, Connecticut

Date of amendment request: June 19, 1997

Description of amendment request: Technical Specification 3/4.7.1.3

requires sufficient water to be available for the auxiliary feedwater

(AFW) system to maintain the reactor coolant system at hot standby for

10 hours before cooling down to hot shutdown in the next 6 hours. The

proposed amendment would increase the required volume of water when the

condensate storage tank is used, make editorial changes, and expand the

description in the appropriate Bases section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

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

significant hazards consideration (SHC). The basis for this

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

satisfied. The proposed revision does not involve an SHC because the

revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillance

4.7.1.3.2 will account for the unusable Condensate Storage Tank

(CST) inventory by increasing the required combined CST and

Demineralized Water Storage Tank (DWST) inventory to 384,000

gallons. The increased required water volume is consistent with the

design of the CST and will provide assurance that sufficient water

is available to maintain the reactor coolant system at Hot Standby

for 10 hours before cooling down to Hot Shutdown in the next 6

hours.

The proposed changes to reword Technical Specification 3/

4.7.1.3, expand the description in Bases Section B3/4.7.1.3 and

modify the description in Bases Section B3/4.7.1.2 are to update and

clarify the requirements.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes to Technical Specification 3/4.7.1.3 do not

change the use of DWST or CST during normal or accident evaluations.

The proposed changes to reword Technical Specification 3/

4.7.1.3, Bases Section B3/4.7.1.3 and Bases Section B3/4.7.1.2 are

to update and clarify the requirements.

Therefore, the proposed revision does not create the possibility

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

evaluated.

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

The proposed changes to Technical Specification Surveillance

4.7.1.3.2 will increase the required inventory for the combined CST

and DWST to account for an additional 50,000 gallons of unusable

inventory due to the CST discharge line location, other physical

characteristics, and measurement uncertainty. The proposed change to

the surveillance requirement will increase the required volume of

the combined CST and DWST inventory to 384,000 gallons. The proposed

change ensures that sufficient water is available to maintain the

Reactor Coolant System at Hot Standby conditions for 10 hours with

steam discharge to the atmosphere, concurrent with a total loss-of-

offsite power, and with an additional 6-hour cool down period to

reduce reactor coolant temperature to 350 [degrees] F.

The proposed changes to Technical Specification 3/4.7.1.3 and

Bases Section 3/4.7.1.3 are to clarify the requirements. The

proposed changes to the Bases Section 3/4.7.1.2 update and expands

the description of the design bases accidents for which AFW System

is credited for accident mitigation. This additional information is

consistent with the current AFW System design bases.