Biweekly Notice

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from February 18, 1994, through March 4, 1994.

The last biweekly notice was published on March 2, 1994 (59 FR 9999).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue,

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

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

Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.

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

intervene is discussed below.

By April 15, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

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

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

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

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

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the

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

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

the Commission by a toll-free telephone call to Western Union at 1-

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

operator should be given Datagram Identification Number N1023 and the

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

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

publication date and page number of this Federal Register notice. A

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

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

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

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

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

Date of amendment requests: January 20, 1994

Description of amendment requests: The proposed amendment would

change the departure from nucleate boiling ratio (DNBR) in Safety

Limits, Section 2.1.1.1, and the associated Bases, as well as the DNBR

- Low Trip Setpoint in Table 2.2-1, and the associated Bases, from a

value of 1.24 to 1.30. In addition, the amendment would add a

methodology supplement entitled,System 80GT1TMInlet

Flow Distribution,'' to the list of methods used to determine the core

operating limits.

Basis for proposed no significant hazards consideration

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

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Involve a significant increase in the probability

or consequences of an accident previously evaluated.

The purpose of the proposed TS amendment is to provide a revised

DNBR Safety Limit and Low DNBR Trip Setpoint to ensure that no

anticipated operational occurrence or postulated accident will

result in core conditions exceeding DNBR Safety Limit.

The change in the DNBR Safety Limit from 1.24 to 1.30 can be

accommodated directly by increasing the limit (including the DNBR

Trip Setpoint) or by an increase in the DNBR overall uncertainty

factors for core operating limit supervisory system (COLSS) (EPOL2

and EPOL4) and core protection calculator (CPC) (BERR1). Using the

1.24 DNBR Safety Limit will result in larger uncertainty factors,

and conversely using the increased DNBR Safety Limit of 1.30 will

result in lower uncertainty factors. Therefore, plant operation for

COLSS and CPC are not significantly affected by the choice of the

DNBR Safety Limit and the Trip Setpoint as long as the corresponding

overall uncertainty factors are calculated and implemented. PVNGS

will implement the 1.30 DNBR Safety Limit and its corresponding

overall uncertainty factors in the reload safety analysis for Unit 3

Cycle 5 and in subsequent reload safety analyses for Units 1 and 2.

The proposed amendment changes only the DNBR Safety Limit and

associated Trip Setpoint, and does not in any way impact the

operation of the plant. Safety and setpoint analyses will be

performed consistent with the increased DNBR limit of 1.30. The core

power distribution during all phases of normal and anticipated

operational occurrences will remain bounded by the initial

conditions assumed in Chapter 15 of the PVNGS Updated Safety

Analysis Report (UFSAR). Furthermore, the UFSAR Chapter 15 analysis

remains bounding because the margins of safety will be maintained.

Therefore, the proposed change to Sections 2.1.1.1 and 2.2.1 (Table

2.2-1) will not significantly increase the probability or

consequences of an accident previously evaluated.

The proposed change to Section 6.9.1.10 does not involve a

significant increase in the probability or consequences of an

accident previously evaluated. The proposed change is administrative

in nature and does not involve any change to the configuration or

method of operation of any plant equipment that is used to mitigate

the consequences of an accident. Also, the proposed change does not

alter the conditions or assumptions in any of the UFSAR accident

analyses. Since the FSAR accident analyses remain bounding, the

radiological consequences previously evaluated are not adversely

affected by the proposed change. Therefore, it can be concluded that

the proposed change to Section 6.9.1.10 will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

of accident from any accident previously evaluated.

The proposed amendment is limited to changing the DNBR Safety

Limit and Low DNBR Trip Setpoint and does not involve any physical

change to plant systems or to the COLSS and the CPC algorithms.

These changes will not affect any safety-related equipment used in

the mitigation of anticipated operational occurrences or design

basis accidents. Therefore, this change to Section 2.1.1.1 and 2.2.1

(Table 2.2-1) will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed change to Section 6.9.1.10 does not create the

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

previously evaluated. The proposed change is administrative in

nature and does not involve any change to the configuration or

method of operation of any plant equipment that is used to mitigate

the consequences of an accident. Accordingly, no new failure modes

have been defined for any plant system or component important to

safety nor has any new limiting failure been identified as a result

of the proposed change. Therefore, it can be concluded that the

proposed change to Section 6.9.1.10 will not create the possibility

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

evaluated.

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

safety.

The DNBR Safety Limit specified in TS 2.1.1.1 and the Low DNBR

Trip Setpoint specified in TS 2.2.1 (Table 2.2-1) ensure that

operation of the reactor is prevented from exceeding the DNBR Safety

Limit during normal operation and design basis anticipated

operational occurrences. Therefore, operating within the increased

DNBR Safety Limit will ensure that no anticipated operational

occurrence or postulated accident will result in core conditions

exceeding the specified DNBR Safety Limit. The UFSAR Chapter 15

analysis remains bounding because the margins of safety will be

maintained. Additionally, the COLSS and the CPC overall uncertainty

factors will be calculated and implemented consistent with the

increased DNBR Safety Limit of 1.30. Therefore, this change to

Section 2.1.1.1 and 2.2.1 (Table 2.2-1) will not result in a

significant reduction in a margin of safety.

The proposed change to Section 6.9.1.10 does not involve a

significant reduction in a margin of safety. The proposed change is

administrative in nature and does not adversely impact the plant's

ability to meet applicable regulatory requirements. Therefore, it

can be concluded that the proposed change to Section 6.9.1.10 does

not involve a significant reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004Attorney for licensees: Nancy

C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public

Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona

85072-3999

NRC Project Director: Theodore R. Quay

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

Station, Plymouth County, Massachusetts

Date of amendment request: January 9, 1994

Description of amendment request: The proposed amendment would make

changes to the Technical Specifications and License. These changes

consist of revised wording for the license, clarify wording to aid

operators in selecting the correct pressure/temperature curve during

startup and shutdown operations, and removal of certain obsolete

mechanical snubber acceptance criteria.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The first proposed change will modify License DPR-35 to

eliminate the need to issue a new page 3 to identify the latest

amendment number. The second change will provide the correction of

an error of omitting the reference to the subcritical mode of

operation, in relation to the pressure/temperature curves. The third

change will remove the unnecessary mechanical snubber functional

test acceptance criterion to determine if drag force has increased

more than 50% since the last functional test.

Modification of License DPR-35 for Pilgrim Nuclear Power Station

to remove the need to update page 3 whenever a new amendment is

approved will reduce an administrative burden. This license change

also precludes a possible administrative error if the correct

reference is somehow missed. This change does not affect plant

operation or design and is considered an administrative change and

as such does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The second change corrects an error of omission made in an

earlier amendment by inserting a reference to the subcritical

reactor operation phase. This proposal will enhance the procedure

changes and training already accomplished as short term corrective

actions. This change does not affect plant operation or design and

is considered an administrative change and therefore does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The third change removes an acceptance criterion for mechanical

snubber testing not required by the ASME Boiler and Pressure Vessel

Code, Section XI, Subsection IWF nor recommended by the vendor for

mechanical snubbers in use at Pilgrim.

This change will not result in any physical modification to

Pilgrim. The mechanical snubbers will continue to be tested in

accordance with existing plant procedures which reference the ASME

Code Section XI, Subsection IWF. Therefore, this is considered an

administrative change and as such, operation of Pilgrim will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

different kind of accident than previously evaluated because they

are administrative in nature and require no physical alterations of

plant configuration or changes to setpoints or operating parameters.

3. The operation of Pilgrim in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety.

Because these changes do not alter plant operation or design and

are considered administrative in nature, they 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 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 Nos. 50-325 and 50-

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

County, North Carolina

Date of amendments request: April 13, 1993

Description of amendments request: The proposed amendments would

revise design features information pertaining to the elevation at which

the spent fuel pool is designed to prevent inadvertent draining. The

proposed amendment would revise this elevation from 116 feet 4 inches

to ll5 feet 11 inches based on the actual spent fuel pool design.

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 current value in Specification 5.6.2 is incorrect. No basis

can be determined for including this value in Specification 5.6.2

other than incorrectly characterizing the normal fuel pool water

level as the design level to be maintained to prevent inadvertent

draining of the fuel pool. This value was incorrectly incorporated

into the initial standard Brunswick Technical Specifications, Water

Level- Spent Fuel Storage Pool.

The 115' 11'' elevation is equal to 20' 10-7/8'' above the top

of the spent fuel rods seated in the storage racks. This level is

still in excess of the minimum level required (20' 6'') by Technical

Specification 3.9.9.

The accident discussed in UFSAR [Updated Final Safety Analysis

Report] Section 9.1.2.3.2.4.2, Loss of Spent Fuel Pool Cooling, is

not impacted by this change since the spent fuel pool safety

functions are not impacted and Technical Specification minimum fuel

pool levels (Specification 3.9.9) are not changed. As such, the

proposed amendments do not involve a significant increase in the

probability of an accident previously evaluated.

The radiological consequences of this accident are discussed in

UFSAR Section 9.1.2.3.2.5. This analysis assumes spent fuel pool

boiling. In addition, the facilities description of the spent fuel

storage pool (Section 9.1.2.2.1), states that the surface of the

water will be maintained at Elevation 116.3 ft, which is the normal

water level of the pool. Therefore, the (lower) designed level to

prevent inadvertent pool draining is not relevant within this

analysis. As such, the proposed amendments do not involve a

significant increase in the consequences of an accident previously

evaluated.

2. The proposed amendments do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

This amendment request corrects a mischaracterization of the

design features and does not involve a change in fuel pool

operations. Specification 5.6.2 states that the fuel pool is

designed to prevent inadvertent draining of the pool below elevation

116'4''. However, it is possible that the pool could drain below

this level by draining through piping connected to the pool coupled

with no flow into the pool. It is not possible, however, with the

fuel pool gates installed, that the fuel pool could be inadvertently

drained below the bottom of the pool overflows to the skimmer surge

tanks. The elevation at the bottom of the overflows to the skimmer

surge tanks is 115'11''. Therefore, this is the correct value to

cite in the design features section of Technical Specifications.

The proposed 115'11'' elevation will not result in new drain

pathways, nor will the minimum fuel pool water level required by the

Technical Specifications be impacted by this change. Therefore, the

proposed amendments do not create the possibility of a new or

different type of accident from any accident previously evaluated.

3. The proposed amendments do not involve a significant

reduction in the margin of safety.

The proposed amendments do not change safety limits, setpoints,

or plant operations. The plant is actually designed to prevent

inadvertent draining of the fuel pool below elevation 115'11'' as

discussed above. This change is not an actual design change; it is a

design clarification correcting the level at which inadvertent

draining of the spent fuel pool is prevented. As such, the proposed

amendments do not involve a significant reduction in the margin of

safety at Brunswick.

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 North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

NRC Project Director: S. Singh Bajwa

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

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

Date of amendment request: February 14, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) in response to Generic Letter

93-08 issued by the NRC and dated December 29, 1993, by relocating the

reactor trip system (RTS) and engineered safety feature activation

system (EFAS) response time limits to the updated Final Safety Analysis

Report (FSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes are administrative in nature and do not

involve any change to the configuration or method of operation of

any plant equipment used to mitigate the consequences of an

accident.

The proposed changes do not alter the conditions or assumptions

in any accident previously evaluated.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

The proposed changes are administrative in nature and do not

involve any change to the configuration or method of operation of

any plant equipment used to mitigate the consequences of an

accident. No new

accident initiators or failure modes are created by relocating

the RTS and ESFAS instrumentation response time limits.

Therefore, the proposed changes 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 will in no

way affect the TS adequacy in ensuring the response times for the

RTS and ESFAS instrumentation do not exceed the limits assumed in

the accident analyses. The proposed changes will have no impact on

the protective boundaries, safety limits, or margin of safety.

Therefore, the proposed changes will not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: July 22, 1993, as supplemented February

4, 1994

Description of amendment request: The proposed amendment would

modify the Facility Operating License (OL) and Technical Specifications

(TSs) to permit uprated power operation. The plant is currently

licensed for operation at 3323 megawatts thermal (MWt), although many

of the original analyses were performed at a design power level of 3467

MWt. The proposed changes would redefine rated thermal power to be 3467

MWt, which represents an approximately 4.3 percent increase over the

currently licensed power level. Implementation of the power uprate

would require minor modifications, such as, resetting of the low set

safety relief setpoints, as well as the recalibration of plant

instrumentation to reflect the uprated power. The proposed changes

follow the generic guidelines for boiling water reactor power uprate

described in General Electric Topical Report, NEDC-31897P-1, ``Generic

Guidelines for General Electric Boiling Water Reactor Power Uprate,''

June 1991.

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:

Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

OL2C(1), TS 1.34 - Increase in Rated Thermal Power to

3467 MWt.

The changes in the OL and TS were evaluated and it was

determined that the probability (frequency) of a DBA [design-basis

accident] or other licensing event occurring is not a significant

function of the power level because the design and regulatory

criteria originally established for plant equipment (ASME [American

Society of Mechanical Engineers] code, IEEE [Institute of Electrical

and Electronics Engineers] standards, NEMA [National Electrical

Manufacturers Association] standards, Regulatory Guide criteria,

etc.) are still imposed for the uprated power level. Scram setpoints

are established such that there will be no significant increase in

scram frequency due to power uprate.

The consequences of hypothetical accidents which would occur

from 102% of the uprated power, as opposed to that previously

evaluated from 102% of the original power, are in all

cases insignificant, since the accident evaluations from 102% of

uprated power do not result in exceeding the NRC-approved acceptance

limits. A spectrum of hypothetical accidents and transients has been

investigated for uprated conditions and the bounding events have

been shown to meet the same regulatory criteria to which they are

currently licensed. In the area of core design, for example, the

fuel operating limits such as Maximum Average Planar Linear Heat

Generation Rate (MAPLHGR) and Safety Limit Minimum Criteria Power

Ratio (SLMCPR) are still met at the uprated power level.

The analysis of all limiting events (Section 9) [of General

Electric Topical Report, NEDC-31994P, ``Power Uprate Licensing

Evaluation for Nine Mile Point Nuclear Power Station Unit 2,''

Revision 1, May 1993] and cycle specific reload analyses will show

plant transients meet the criteria accepted by the NRC as specified

in NEDO-24011, GESTAR II. Challenges to fuel or ECCS [emergency core

cooling system] performance have been evaluated (Section 9.2) [of

NEDC-31994P] and shown to still meet the criteria of 10CFR50.46

using the methodology defined by Appendix K (Regulatory Guide 1.70,

USAR [Updated Safety Analysis Report] Section 6.3). Challenges to

the containment have been evaluated for uprated power (Section 4.1)

[of NEDC-31994P] and still meet 10CFR[Part]50 Appendix A Criterion

38, Long Term Cooling and Criterion 50, Containment. Radiological

Release events have been evaluated [Sections 8.4 and 8.5] [of NEDC-

31994P] and shown to be a small fraction of the criteria of

10CFR[Part]100 (Regulatory Guide 1.70 USAR Chapter 15).

The results of these analyses as discussed above demonstrate

that operation of [at] the power uprate level does not significantly

increase the probability or consequences of any accident previously

evaluated.

OL2C(7) - Change in Allowable Feedwater Temperature

This change is made to maintain an equivalent 20 deg.F allowable

operating range of final feedwater temperature for uprated power

(405 to 425 deg.F) as compared to the presently licensed range (400

to 420 deg.F). No change is made to the current method and criteria

for operation of the feedwater heating systems. The limiting

transient (Feedwater Controller Failure - Maximum Demand) has been

evaluated at a feedwater temperature of 405 deg.F (Section 9.1.3)

[of NEDC-31994P] to demonstrate compliance with all current thermal

limits criteria. Previous feedwater nozzle evaluations have shown

that operation with the proposed feedwater temperature operating

range is acceptable. Therefore, there is no significant increase in

the probability or consequences of an accident previously evaluated.

TS Table 2.2.1-1 - Reactor Protection System Instrument

Setpoints

The increases in steam dome high pressure scram instrument

setpoints are made to ensure that there is no significant increase

in the frequency of scrams due to operation at the higher pressure.

The increase of the high pressure scram setpoint by the same amount

as the increase in the planned operating pressure maintains the same

level of trip avoidance for scram as originally provided. The high

pressure scram is used as a backup to other scram signals. It has

been shown that this role is still adequate for uprated operation

with the revised setpoints (e.g., vessel overpressure protection).

Since the backup protectionfunctions and the current margins to trip

avoidance are maintained with the revised setpoints, there is no

significant increase in the probability or consequences of an

accident previously evaluated.

TS Bases Table B2.1.2-2 - Add footnote to applicability

of table to uprated operation.

The parameters listed in TS Bases Table B2.1.2-1 come from the

original statistical analysis performed for BWR4/5 core designs

(including NMP2) [Nine Mile Point Nuclear Station, Unit 2]. As

discussed in Section 3.2 of LTR2 (Reference 11-1) [General Electric

Topical Report NEDC-31984P, ``Generic Evaluation of General Electric

Boiling Water Reactor Power Uprate,'' July 1991], the uprated

average bundle power is used to determine the applicability of the

generic Safety Limit Minimum Critical Power Ratio (SLMCPR) basis for

each plant. The average bundle power for NMP2 after power uprate is

4.538 MWt per bundle. This value is acceptable for application of

the generic SLMCR statistical analysis to uprated NMP2. The generic

analysis is documented through NEDC-24011-P-A (GESTAR II) and NEDC-

31152P (GE Fuel Bundle Design). Therefore, there is no significant

increase in the probability or consequences of an accident

previously evaluated.

TS 4.1.5.c and TS 4.1.5.d2 - Increases in SLCS [standby

liquid control system] Surveillance Test Pressure and SLCS pump

discharge relief valve setpoint.

The standby liquid control system test pressure increase ensures

continued ability of the system to pump the required amount of

sodium pentaborate at the higher operating pressure associated with

power uprate. This test pressure increase is consistent with the

ATWS [anticipated transient without scram] analysis provided in

Section 9.3 [of NEDC-31994P]. The higher pressure setpoint for the

SLCS relief valve does not exceed the design capability of the SLCS

components. Surveillance testing at the increased pressure will

maintain the system's design capability for operation at uprated

conditions. These changes therefore do not increase the probability

or consequences of a previously evaluated accident.

TS Table 3.3.1-1 - Note (i), footnote (**), Action 6

footnote (*), and Table 3.3.4.2-1 - footnote (**)

The setpoints for the bypass of T/G [turbine generator] trip

scram and RPT [recirculation pump trip] at 30% of rated power are

changed to 125.8 psig and 136.4 psig to be consistent with uprated

power.

These changes reflect the redefinition of rated conditions. They

are consistent with the approach discussed in Section F.4.2(c) of

LTR1 (Reference 11-2) [General Electric Topical Report NEDC-31897P-

1, ``Generic Guidelines for General Electric Boiling Water Reactor

Power Uprate,'' June 1991]. There is no significant impact on the

transient safety analyses which establish core thermal operating

limit since T/G trips at this partial power setpoint are not

limiting. Therefore, there is no significant increase in the

probability or consequences of an accident previously evaluated.

TS Table 3.3.2-2 Item 1.C.3 - Increase in main

steamline high flow isolation differential pressure setpoint and

allowable value.

The main steamline high flow trip TS changes reflect the

redefinition of rated steam flow during uprated power operation and

the application of GE [General Electric Company] setpoint

methodology. The current analytical basis of 140% of rated steam

flow is maintained for uprated operation to ensure that an adequate

trip avoidance margin is maintained (e.g., for disturbances caused

by full closure testing of MSIVs [main steam isolation valves] or

turbine inlet valves). The revised setpoints ensure that there is no

effect on the probability of inadvertent isolation; they have no

effect on the probability of occurrence of a main steamline break.

The same isolation initiation function for the main steam line break

accident is maintained (Section 5.1.2.5) [of NEDC-31994P]. Therefore

these setpoint changes do not significantly increase the consequence

of the main steamline break accident.

TS Table 3.3.2-2 Item 1d - Increase the main steamline

tunnel temperature setpoints.

These isolation setpoints are changed to reflect the slight

increase (about 1 deg.F) in the operating temperature expected for

uprated operation. Margins between trip setpoints and operating

temperature are maintained. The increases will avoid unnecessary

trips. The revised trip setpoints were derived using the GE setpoint

methodology (documented in NEDC-31336). The setpoints still perform

their isolation function equivalent to current operation.

Therefore no significant increase in the probability or

consequences of an accident previously evaluated results from these

changes.

TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor

vessel high pressure trip and allowable setpoint.

The ATWS RPT high pressure setpoints are increased to correspond

to the increase in the steam dome operating pressure due to power

uprate. This increase maintains the current margin between the

operating condition and the trip setpoint to avoid unnecessary

trips. The capability of the system to adequately perform its ATWS

function with the new setpoints is shown in Section 9.3 [of NEDC-

31994P]. Therefore the change does not cause a significant increase

in the probability or consequences of an accident previously

evaluated.

TS Figure 3.4.1.1-1 - The figure is revised to reflect

new definition of rated thermal power in terms of megawatts thermal.

This change is made to be consistent with the new definition of

rated thermal power. The current restrictions on operation within

the restricted power/flow zone are unchanged. The basis for this

change is described in Section 3.2 of LTR2 (Reference 11-1) [NEDC-

31984P]. There is no significant change in the previously evaluated

potential for initiation of core thermal hydraulic instability.

Therefore, this TS change ensures that power uprate operation will

not cause a significant increase in the probability of [or]

consequences of an accident previously evaluated.

TS 3.4.2 - Increase of spring setpoints for the two

lowest set SRVs [safety relief valves].

The two low set SRV setpoints are increased to accommodate the

change in operating pressure after power uprate. This increase in

the SRV setpoints ensures that approximately the same difference is

maintained between the RPV [reactor pressure vessel] pressure and

the lowest SRV setpoint such that there is no increase in the number

of unnecessary SRV actuations. The increase in the spring setpoints

by the same amount as the increase planned for normal operation also

maintains acceptable simmer margin for the SRVs. The SRVs are

capable of operating at uprated temperatures and pressures as

evaluated generically in Section 4.6 of LTR2 [NEDC-31984P]. As

described in Sections 3.2 and 9.3.1 [of NEDC-31994P] a higher RPV

peak pressure results due to uprate conditions but it is maintained

well within the ASME Code allowable peak pressure of 1375 psig.

Therefore, no significant increase in the probability or

consequences of an accident previously evaluated is caused by this

change.

TS 4.4.6.1.3-1 - Revision to the neutron fluence lead

factor.

The increase in the lead factor from 0.41 to 0.46 reflects

updated calculations for the higher power level and projected

fluence distributions. This calculation accounted for the locations

of the NMP2 specimen capsules (at three locations on the vessel wall

around the core beltline region), and the projected uprated

equilibrium cycle spatial power distributions. Since the revised

lead factor is consistent with the requirements for vessel

surveillance, the change causes no significant increase in the

probability or consequences or [of] an accident previously

evaluated.

TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam

dome operating pressure limit.

This change to the dome operating pressure limit is consistent

with and meets the current design criteria used for evaluation of

steady state operating conditions and for the most limiting

transient and accident events, i.e., vessel overpressure protection

and a loss-of-coolant accident (Sections 3.2 and 4.3) [of NEDC-

31994P]. Therefore, there is no significant increase in the

probability or consequences of an accident previously evaluated.

TS 4.7.4b - Increase in RCIC [reactor core isolation

cooling] Surveillance Test Pressure.

The increase in the RCIC surveillance test pressure requires

system testing at the higher operating pressure with power uprate.

The RCIC system has been evaluated and demonstrated to be capable of

injecting its design flow rate at the higher reactor pressure

associated with power uprate as discussed in Section 4.2 of LTR2

(Reference 1) [NEDC-31984P]. This evaluation applies to NMP2 as

described in Section 3.8 [of NEDC-31994P]. Therefore, this TS change

ensures that power uprate operation will not cause a significant

increase in the probability or consequences of an accident

previously evaluated.

TS Bases 3/4.2 (References), and TS 6.9.1.9.b(1)

(Administrative Control) - Revised the reference for the LOCA [loss-

of-coolant accident] analysis methodology to the SAFER/GESTR-LOCA

methodology report.

These changes are made to incorporate the power uprate LOCA

licensing basis. Reference 1 of TS Bases 3/4.2 (References) and the

report noted in TS 6.9.1.9.b(1) are changed to reflect the improved

SAFER/GESTR-LOCA methodology used for the NMP2 loss-of-coolant

accident analysis for power uprate. This methodology has been

previously approved by the NRC. These changes are made for

documentation consistency and there is no significant increase in

the probability or consequences of an accident previously evaluated.

TS Bases Table B3.2.1-1 - Significant input parameters

used in the LOCA analysis.

The changes in the plant parameters used in the uprated LOCA

analysis are provided from the power uprate analysis (Reference 4-

15, Section 4) [General Electric Report NEDC-31830P, ``NMP2 SAFER/

GESTR-LOCA Loss-of-Coolant Accident Analysis,'' Revision 1, November

1990]. The power and steam flow values are consistent with

Regulatory Guide 1.49. The new analysis parameters will also be

included in USAR Table 6.3-1 as uprate is implemented. Detailed

information about application of the SAFER/GESTR methodology is

provided in Reference 1 of TS Bases 3/4.2 (Reference 4-14, Section

4) [General Electric Report NEDE-23785-1-PA, ``The GESTR/LOCA and

SAFER Models for the Evaluation of the Loss of Coolant Accident,''

Revision 1, October 1984]. The LOCA analysis (Section 4.3) [of NEDC-

31994P] shows that all required criteria are met for operation with

the uprated parameters.

Footnote (*) is revised to provide the correct reference for the

LOCA analysis parameters for NMP2 power uprate. Since the LOCA

analysis for uprated operation meets all required criteria, these TS

changes do not cause an increase in the probability or consequences

of an accident previously evaluated.

TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required

capability of the HPCS [high pressure core spray] pump and the

corresponding differential pressure.

The increase in the differential pressure for HPCS pump flow

accommodates the increase in SRV setpoint valves as discussed for

changes to TS 3.4.2 earlier. This change maintains the currently

designed functional capability of the HPCS system to provide coolant

inventory during isolation conditions after a loss of feedwater flow

transient (backup to RCIC) and during a main steam line break

(outside containment) accident.

The small change in the HPCS pump flow (517 versus 516) [gpm] is

made to be consistent with the value used in the NMP2 SAFER/GESTR

analysis for a loss of coolant accident. This small change corrects

the Technical Specification bases for this parameter.

Therefore, this TS change ensures that power uprate operation

will not cause a significant increase in the probability or

consequences of an accident previously evaluated.

TS Bases B3/4.6.1.2, B3/4.6.1.5, and B3/4.6.2 - Maximum

containment pressure for leakage testing.

The bases for the value currently in the TS for the maximum

containment pressure are reworded to clarify that the maximum

containment pressure after power uprate will be maintained below the

current value used for containment leak rate testing. Section 4.1

[of NEDC-31994P] documents the containment analysis for power uprate

and shows a peak DBA-LOCA calculated pressure of 36.8 psig (less

than the current testing requirement of 39.75 psig). There is no

impact on currently approved requirements and test procedures.

Therefore, there is no significant increase in the probability or

consequences of an accident previously evaluated.

Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The Operating License changes in power level and allowable

feedwater temperature, and the associated Technical Specification

changes (all listed in Table 11-1) [of NEDC-31994P] will not create

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

accident previously evaluated.

Equipment that could be impacted by power uprate has been

evaluated. No new operating mode, equipment lineup, accident

scenario, or equipment failure mode has been identified. The full

spectrum of accident considerations defined in Regulatory Guide 1.70

have been reviewed and no new or different kind of accident has been

identified. Power uprate uses already developed technology and

applies it within the capabilities of already existing plant

equipment in accordance with presently existing regulatory criteria

to include NRC approved codes, standards, and methods. GE has

designed BWRs of higher power levels than the uprated power of any

of the currently operating BWR fleet and no new power dependent

accidents have been identified.

The Technical Specifications changes required to implement power

uprate require minor changes to the configuration of the plant, and

all the Technical Specification changes have been evaluated and are

acceptable.

Will the change involve a significant reduction in a margin of

safety?

OL2C(1), TS 1.34 - Increase in Rated Thermal Power to

3467 MWt.

Power uprate will not involve a significant reduction in a

margin of safety, since the licensing evaluations were performed

either at plant conditions higher than the proposed uprate

conditions, or used approved methodologies which incorporate

appropriate allowances for uncertainties. As discussed throughout

this report (e.g., Section 11.1) [of NEDC-31994P] and in Section 5

of Reference 11-2 [NEDC-31897P-1], the safety margins prescribed by

the Code of Federal Regulations have been maintained by meeting the

appropriate regulatory criteria. Similarly, the margins provided by

the application of the American Society of Mechanical Engineers

(ASME) design acceptance criteria where applicable have been

maintained (e.g., see Section 3.2) [of NEDC-31994P]. Other margin-

assuring acceptance criteria have also been maintained.

All limiting accident and transient analyses have been

reperformed at uprated power operating conditions consistent with

the requested Technical Specification changes. The NRC-approved

SAFER/GESTR-LOCA methodology was used in the LOCA analysis.

Additionally, Reference 11-2 [NEDC-31897P-1] addresses the BWR

generic acceptability of analytical evaluations for the loss of

feedwater transient, stability, core spray distribution, safety

limit minimum critical power ratio, containment atmosphere

combustibility, materials and coolant chemistry, and anticipated

transients without scram (ATWS).

As discussed in Section 5.2.3 of Reference 11-2 [NEDC-31897P-1],

offsite doses for the DBA/LOCA will increase proportionally to

reactor power and can be compared on a consistent basis.

As evaluated in Sections 8.5 and 9.2 [of NEDC-31994P], the

results remain a small fraction of the acceptance criteria of

10CFR[Part]100.

The radiological doses resulting from the DBA/LOCA and MSLB

[main steam line break] accidents were initially analyzed at 3489

MWt (105% of 3323 MWt). For the power uprate program, the increase

in the analyzed power level is only 1.3% (from 3489 to 3536 MWt)

which provides the uncertainty factor (2%) required by Regulatory

Guide 1.49.

It is concluded that there is no significant decrease in a

margin of safety.

OL2C(7) - Change in Allowable Feedwater Temperature

The increase in the lower limit of the allowable operating range

of the final feedwater temperature for uprate power was evaluated by

reanalyzing the Feedwater Controller Failure - Maximum Demand

transient at a feedwater temperature of 405 deg.F (Section 9.1.3)

[of NEDC-31994P]. In addition, the reactor pressure vessel feedwater

nozzle has been evaluated for the 20 deg.F range of feedwater

temperature. The results of these evaluations demonstrate that

current fuel thermal limits criteria and ASME Code criteria are met.

Therefore, there is no significant decrease in a margin of safety.

TS Table 2.2.1-1 - Reactor Protection System Instrument

Setpoints

The increases in the steam dome high pressure scram instrument

setpoints for uprated power were evaluated by determining if the

high pressure scram, which is used as a backup to other scram

signals, provides adequate overpressure protection. The evaluation

demonstrates that the backup protection function, with the revised

setpoints, continues to provide adequate overpressure protection at

uprated power conditions by meeting the applicable ASME Code

criteria. It is concluded that there is no significant decrease in a

margin of safety.

TS Bases Table B2.1.2-2 - Add footnote to applicability

of table to uprated operation

The nominal values of parameters used in the statistical

analyses of the fuel cladding integrity safety limit were re-

evaluated at power uprate conditions. This evaluation demonstrates

that the average bundle power at uprated conditions for NMP2 is

acceptable for application of the generic safety limit minimum

critical power ratio statistical analysis. Therefore, it is

concluded that there is no significant decrease in a margin of

safety.

TS 4.1.5.c and TS 4.1.5.d2 - Increase in SLCS

Surveillance Test Pressure and SLCS pump discharge relief valve

setpoint.

The SLCS surveillance test pressure was increased to provide

periodic demonstration of the ability of the SLCS to provide the

required amount of sodium pentaborate solution at the higher

pressure associated with an ATWS event postulated to occur at power

uprate conditions. At this increased SLCS discharge pressure, the

system provides an adequate shutdown backup capability by having the

ability to bring the isolated reactor from full power to a cold,

Xenon-free shutdown condition, assuming that the withdrawn control

rods remain fixed in the uprated power pattern.

For power uprate, the capability of the SLCS to respond with

adequate margin to a postulated ATWS event was confirmed. The most

limiting ATWS events evaluated for peak vessel pressure and peak

suppression pool temperature were: (1) closure of all MSIVs and (2)

inadvertent opening of a relief valve. The peak pressure for the

MSIV closure event which included simulation of the higher relief

setpoints and two relief valves out of service, demonstrates that

the peak pressure, 1325 psig, remains below the ASME emergency

overpressure protection criteria of approximately 1500 psig, which

is applicable to an ATWS event. The reactor pressure is controlled

by the relief valves (after the initial peak) within the pressure

specified in this revised Technical Specification. SLCS injection

takes place during this period with the relief valves controlling

pressure.

The peak suppression pool temperature for the inadvertent

opening of a relief valve was demonstrated to remain below the ATWS

peak pool temperature criteria of 190 deg. for a Mark II containment

design, which is applicable to NMP2. Peak containment pressure was

well below the 45 psig containment design pressure. For this event,

SLCS injection will be at vessel pressures bounded by the revised

Technical Specification. The higher pressure setpoint of the SLCS

pump discharge relief valve provides adequate overpressure

protection of the SLCS pressure boundary by meeting the applicable

ASME Code criteria (equal to or less than the system piping design

pressure).

In summary, peak vessel pressure is below ASME code criteria,

and suppression pool temperature is below the ATWS peak pool

temperature criterion for Mark II containment design, peak

containment pressure is well below the containment design pressure,

the SLCS injection pressure during the bounding events is within the

new Technical Specification testing requirement, and the SLCS

pressure boundary is maintained in compliance with ASME Code

criteria. Therefore, it is concluded that there is no significant

decrease in a margin to safety.

TS Table 3.3.1-1 - Note (i), footnote (**), Action 6,

footnote (*), and Table 3.3.4.2-1 footnote (**)

The increase in the setpoints for the bypass of T/G trip scram

and RPT at 30% power are made to be consistent with uprated power.

These increased setpoints do not significantly reduce a margin of

safety since the T/G trips at this partial power setpoint continue

to be non-limiting events.

TS Table 3.3.2-2 Item 1.c.3 - Increase in main steam

line high flow differential pressure setpoint and allowable valve.

The increase in the main steam line high flow differential

pressure setpoint and allowable value reflect the redefinition of

rated conditions. The increased setpoint will maintain the same

inadvertent trip avoidance margin, thereby avoiding any increase in

the frequency of occurrence of isolation events. The closure of the

MSIV remains assured during the limiting event (the steam line break

accident). The break flow rate (controlled by the flow restrictor)

will be about 190% (Section 3.5) [of NEDC-31994P], so the setpoint

at less than 140% will sense the accident as effectively as for

current operation. It is concluded that this change does not result

in a significant decrease in a margin of safety.

TS Table 3.3.2-2 Item 1.d - Increase in main steam line

tunnel temperature setpoints.

These isolation setpoints are changed to reflect the slight

increase (about 1 deg.F) in the steam tunnel operating temperature

expected for uprated operation. The increase in these setpoints

ensures adequate inadvertent trip avoidance. The analytical upper

limits for these setpoints are not changed so that their safety

functions are not impacted by the Technical Specification changes.

For example, the instruments will act at the same setpoints assumed

in previous analysis for a main steamline break, ensuring that

offsite radiological doses remain a small fraction of 10CFR[Part]100

criteria and within GDC19 criteria for control room doses.

Therefore, there is no significant decrease in a margin of safety.

TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor

vessel high pressure trip and allowable setpoints

The purpose of the high pressure RPT is to reduce reactor power

level during a postulated pressurization transient with scram

assumed to fail (ATWS). The physical phenomenon involved is that the

void reactivity feedback from a pressurization transient adds

positive reactivity to the reactor system. However, the high

pressure RPT system trips both recirculation pumps to the low speed

condition, thereby increasing core void fraction and creating

negative reactivity to reduce the power transient. This enables the

safety/relief valves to maintain peak pressure within the ASME

overpressure emergency limit for the bounding ATWS case (Section

9.3.1) [of NEDC-31994P].

For power uprate, the capability of the SLCS to respond to a

postulated ATWS event with adequate margin was confirmed (Section

9.3.1) [of NEDC-31994P]. By reducing reactor power until the SLCS

can be injected to achieve full shutdown, the RPT also reduces

suppression pool temperature for isolation cases (also shown to be

acceptable for power uprate conditions in Section 9.3.1) [of NEDC-

31994P]. Therefore, it is concluded that there is no significant

decrease in a margin of safety.

TS Figure 3.4.1.1-1 - The figure is revised to reflect

the new definition of rated thermal power in terms of megawatts

thermal.

This change is made to be consistent with the new definition of

rated thermal power. As described in Section 3.2 of LTR2 (Reference

11-1) [NEDC-31984], the change to the power flow restricted zone is

made to maintain the same operating constraints and stability margin

that were established for the current power level. This change

avoids any increase in the possibility of occurrence or any increase

in the potential effects of power oscillations. Therefore, there is

no significant decrease in a margin of safety.

TS 3.4.2 - Increase of spring setpoints for the two

lowest set SRVs.

The two low set SRV setpoints are increased to accommodate the

change in operating pressure after power uprate. This change

maintains a simmer margin of greater than 120 psig. Power uprate

analysis shows that the revised SRVs still maintain the peak RPV

pressure within the ASME Code Upset limit of 1375 psig for the

limiting pressurization event (MSIV closure when credit is only

taken for the backup high neutron flux scram) and provide adequate

protection for postulated ATWS events. See Sections 3.2 and 9.3.1

[of NEDC-31994P] for further discussion. Therefore, it is concluded

that there is no significant decrease in a margin of safety.

TS 4.4.6.1.3-1 - Revision of the neutron fluence lead

factor.

The increase in the lead factor includes consideration of the

higher power level and projected spatial power distributions for an

uprated equilibrium cycle. The evaluation at uprated conditions

utilized the same calculational approach as performed for the

current neutron fluence lead factor, but using more precise input

parameters for uprated conditions. Therefore, it is concluded that

there is no significant decrease in a margin of safety.

TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam

dome operating pressure limit.

The change to the dome operating pressure limit is made to be

consistent with the new operating pressure for uprated thermal

power. This change is used as a direct initial condition analysis

input or sensitivity study parameter in the evaluation of steady

state operating conditions and for the most limiting transients and

accident events, i.e., vessel overpressure protection and LOCA. With

this revised limit, peak vessel pressure remains below ASME Code

criteria, and LOCA fuel performance satisfies the requirements of

10CFR50.46 and 10CFR[Part]50 Appendix K. Therefore, there is no

significant decrease in a margin of safety.

TS 4.7.4b - Increase in RCIC Surveillance Test

Pressure.

The RCIC surveillance test pressure was increased to provide

periodic demonstration of the ability of RCIC system to perform

consistent with the requirements of the analyses at the higher

operating pressure associated with power uprate conditions. An

evaluation of the RCIC system confirmed its ability to operate at

slightly higher turbine speed and provide its design flow rate at

power uprate conditions. RCIC system performance will be confirmed

during the initial power ascension to uprated conditions (and

periodically thereafter per the Technical Specification). Therefore,

it is concluded that there is no significant decrease in a margin of

safety.

TS Bases 3/4.2 (References) and TS 6.9.1.9.b (1)

(Administrative Control) - Revised the references for the LOCA

analysis methodology to the SAFER/GESTR-LOCA methodology report.

These changes are made to incorporate the power uprate LOCA

licensing basis. Since SAFER/GESTR-LOCA methodology has been

previously approved by the NRC and is acceptable for use for NMP2,

it is concluded this change does not significantly decrease a margin

of safety.

TS Bases Table B3.2.1-1 - Significant input parameters

used in the LOCA analyses.

The changes in the plant parameters reflect the power uprate

condition. These changes have been reflected as input parameters in

the LOCA analyses consistent with Regulatory Guide 1.49. Since the

LOCA analysis demonstrates that 10CFR50.46 and 10CFR[Part]50

Appendix K criteria are met for operation with the uprated

parameters, it is concluded that there is no significant decrease in

a margin of safety.

TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required

capability of the HPCS pump and the corresponding differential

pressure.

The increase in differential pressure accommodates the increase

in SRV setpoint values previously discussed to TS 3.4.2. The

increased differential pressure ensures there is sufficient HPCS

flow, assuming the two lowest setpoint SRVs are out of service, and

the commencement of flow as reflected in the analysis of isolation

events. The increase in HPCS pump flow is reflected in the LOCA

analyses (516 to 517 gpm). This small change corrects the Technical

Specification bases for this parameter. The revised parameters for

the HPCS pump differential pressure and flow are reflected as inputs

to the LOCA analyses and analyses of isolation events. Since the

LOCA analysis meets 10CFR50.46 criteria and 10CFR[Part]50 Appendix K

criteria, and the isolation events meet all required criteria (e.g.

top of fuel remains covered for the loss of feedwater transient) it

is concluded that there is no significant decrease in a margin of

safety.

TS Bases B3/4.6.1.2, B3/4.6.1.5 and B3/4.6.2 - Maximum

containment pressure for leakage testing.

The bases for the value currently in the TS for the maximum

containment pressure are reworded to clarify that the maximum

containment pressure for power uprate has been calculated to remain

below the current value used for containment leak rate testing.

Therefore, it is concluded that there is no significant decrease in

the margin of safety.

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

this review, it appears that the three standards of 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: Robert A. Capra

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

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

Pennsylvania

Date of amendment request: January 10, 1994

Description of amendment request: The proposed amendment would

relocate the seismic monitoring instrumentation Limiting Condition for

Operation, Surveillance Requirements and associated tables and Bases

contained in TS sections 3.3.7.2 and 4.3.7.2 to the Updated Final

Safety Analysis Report (UFSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The function of the seismic monitoring instrumentation system is

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

not initiate or mitigate an accident previously evaluated.

Furthermore, the proposed TS changes to relocate the seismic

monitoring instrumentation requirements from TS to the UFSAR are in

accordance with the criteria for determining those requirements that

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

statement, ``Final Policy Statement on Technical Specifications

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

seismic monitoring instrumentation LCO, SRs, and associated tables

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

continue to be implemented by administrative controls that will

satisfy the applicable requirements of TS section 6 ``Administrative

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

systems and equipment and to the applicable administrative controls

in accordance with the provisions of 10CFR50.59.

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

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

restriction that is an initial condition of a Design Basis Accident

or Transient Analysis that either assumes the failure of or presents

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

seismic monitoring instrumentation system is not a system that

monitors a process variable that is an initial condition for

accident or transient analyses. The seismic monitoring

instrumentation is also not a design feature or an operating

restriction that is an initial condition of a Design Basis Accident

or transient analyses since it only provides information regarding

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

earthquake. Therefore, the current LGS seismic monitoring

instrumentation TS requirements do not meet Criterion 2 of the July

22, 1993 NRC final policy statement.

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

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

primary success path and which functions or actuates to mitigate a

Design Basis Accident or Transient that either assumes the failure

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

barrier.'' The LGS seismic monitoring instrumentation system does

not provide a function or actuate in order to mitigate the

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

current LGS seismic monitoring instrumentation TS requirements do

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

statement.

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

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

experience or probabilistic safety assessment has shown to be

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

shown that the LGS seismic monitoring instrumentation system has no

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

policy statement. Furthermore, LGS specific probabilistic risk

assessment (PRA) does not credit the seismic monitoring

instrumentation system as a significant factor in the plant response

to an accident. Therefore, the current LGS seismic monitoring

instrumentation TS requirements do not meet Criterion 4 of the July

22, 1993 NRC final policy statement for determining those

requirements that should remain in TS. This conclusion is consistent

with the function of the seismic monitoring instrumentation system

stated above.

These proposed TS changes will maintain the current operation,

maintenance, testing, and system operability controls of the seismic

monitoring instrumentation system. Furthermore, any further changes

to the seismic monitoring instrumentation system will be evaluated

for the effect of the those changes on system reliability as

required by 10CFR50.59. The seismic monitoring instrumentation

system performance will not decrease due to these proposed TS

changes and the system will continue to be administratively

controlled in accordance with TS Section 6, including the

requirements of 10CFR50.59, thereby precluding a future decrease in

its performance.

In accordance with the current TS Section 3.3.7.2, with the

seismic monitoring instrumentation inoperable, the plant would not

be required to shut down and the provisions of TS Section 3.0.3

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

inoperability of this system and therefore the consequences of an

accident while this system is inoperable, was previously evaluated

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

condition.

Since the seismic monitoring instrumentation system does not

monitor a process variable that is an initial condition for an

accident or transient analyses, or actuates any accident mitigation

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

modification of the seismic monitoring instrumentation system will

continue to be administratively controlled, including the

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

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

in the probability or consequences of an accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The function of the seismic monitoring instrumentation system is

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

proposed TS changes to relocate the seismic monitoring instruments

requirements from TS to the UFSAR are in accordance with the

criteria for determining those requirements that should remain in

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

July 22, 1993. The seismic monitoring instrumentation system does

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

accident or transient analyses.

The seismic monitoring instrumentation is also not a design

feature or an operating restriction that is an initial condition of

a Design Basis Accident or transient analyses since it only provides

information regarding the magnitude of and the plant equipment

response to a Design Basis earthquake.

These proposed TS changes to relocate the TS requirements to the

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

which the seismic monitoring instrumentation system will perform its

function, and any future changes will continue to be

administratively controlled in accordance with TS

Section 6, including the requirements of 10CFR50.59.

These proposed TS changes will not impose new conditions nor

result in new types of equipment which will result in different

types of malfunctions of equipment important to safety than any type

previously evaluated.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

These proposed TS changes to relocate the seismic monitoring

instrumentation requirements from TS to the UFSAR are in accordance

with the criteria for determining those requirements that should

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

dated July 22, 1993.

Criterion 1 of the NRC final policy statement

states,Installed instrumentation that is used to detect,

and indicate in the control room, a significant abnormal degradation

of the reactor coolant pressure boundary.'' The NRC final policy

statement explains that ''...This criterion is intended to ensure

that Technical Specifications control those instruments specifically

installed to detect excessive reactor coolant leakage. This

criterion should not, however, be interpreted to include

instrumentation to detect precursors to reactor coolant pressure

boundary leakage or instrumentation to identify the source of actual

leakage (e.g., loose parts monitor, seismic instrumentation, valve

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

UFSAR, and TS Bases 3.3.7.2, the seismic monitoring instumentation

does not detect, and indicate in the control room, a significant

abnormal degradation of the reactor coolant pressure boundary.

Therefore, the current LGS seismic monitoring instrumentation TS

requirements do not meet Criterion 1. Furthermore, operating

experience has shown that the LGS seismic instrumentation system has

no impact on public health and safety as defined by the NRC final

policy statement. In addition, the LGS specific PRA does not credit

the seismic monitoring instrumentation system as a significant

factor in the plant response to accidents.

The seismic monitoring instrumentation LCO, SRs, and associated

tables and Bases proposed for relocation to the LGS UFSAR will

continue to be implemented by administrative controls that will

satisfy the applicable requirements of TS section 6 ``Administrative

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

the system and applicable administrative controls in accordance with

the provisions of 10CFR50.59.

Accordingly, based on the above discussion of NRC specific

guidance, operating experience, and continued imposition of

administrative controls, the proposed TS changes do not involve a

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller

South Carolina Electric & Gas Company, South Carolina Public

ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: December 13, 1993, as supplemented

February 2, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to allow for the use and

subsequent storage of fuel with an initial enrichment of 5.0 w/o

[weight percent] Uranium 235. The TS currently allow the use of fuel

with a maximum enrichment of 4.25 w/o Uranium 235. The proposed

amendment would also revise the restrictions on fuel storage in regions

1 and 2 of the spent fuel pool to ensure that the design basis for

preventing criticality is maintained in the event of absorber panel

shrinkage.

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 change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

There is no increase in the probability of an accident because

the physical characteristics of a fuel assembly are not changed when

fuel enrichment is increased. Fuel assembly movement will continue

to be controlled by approved fuel handling procedures.

There is no increase in the consequences of an accident because

fuel cycle designs will continue to be analyzed with NRC-approved

codes and methods to ensure the design bases for VCSNS [Virgil C.

Summer Nuclear Station] are satisfied. The double contingency

principle of ANSI/ANS 8.1-1983 can be applied to any postulated

accident in the spent fuel pool which could cause reactivity to

increase beyond the analyzed conditions. As shown in Attachment IV,

the level of boron in the VCSNS spent fuel pool is sufficient to

maintain Keff [effective neutron multiplication factor] less than or

equal to 0.95. There is no postulated accident which could cause

reactivity to increase beyond the analyzed conditions in the new

fuel rack.

The radiological consequence analyses [...] performed to support

the installation of replacement steam generators at VCSNS included

the development of source terms which bound fuel enrichments up to

5.0 w/o U235 [Uranium 235] and average discharge burnups up to

65,730 MWD/MTU [megawatt days per metric ton uranium], which bounds

the currently licensed burnup for fuel at VCSNS. These source terms

were used to calculate offsite doses for accidents that are

postulated to result in the release of fission products to the

environment, including the fuel handling accident. In all cases, the

dose results are within 10CFR100 limits.

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

different kind of accident from any accident previously evaluated.

The proposed Technical Specification changes do not involve any

physical changes to the plant or any changes to the method in which

the plant is operated. They do not affect the performance or

qualification of safety related equipment. Therefore the possibility

of a different type of accident or malfunction than previously

considered is not created.

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

margin of safety.

Criticality analyses [...] have been performed for the spent

fuel pool to allow for storage of fuel assemblies with enrichments

up to 5.0 without U-235. The proposed Technical specification

changes include those necessary to maintain Keff less than or equal

to 0.95, including conservative allowances for uncertainties and

biases, when the pool is flooded with unborated water.

The new fuel racks have been previously analyzed [...] for

storage of fuel assemblies with enrichments up to 5.0 w/o U-235. For

the flooded condition Keff does not exceed 0.95 including

conservative allowances for uncertainties and biases. For the

normally dry condition Keff does not exceed 0.98 for the low density

optimum moderation condition. However, the proposed Technical

Specification changes require fuel assemblies with enrichment above

4.0 w/o U-235 to contain integral fuel burnable absorbers such that

the maximum reference fuel [infinite neutron multiplication factor]

is less than or equal to 1.460 in unborated water at 68 deg.F due to

restrictions on spent fuel storage.

Since the proposed changes ensure that the design basis for

preventing criticality in the fuel storage areas is preserved and

since fuel cycle designs will continue to be analyzed with NRC-

approved codes and methods to ensure the design bases for VCSNS are

satisfied, there is no significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Fairfield County Library,

Garden and Washington Streets, Winnsboro, South Carolina 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: S. Singh Bajwa

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: February 16, 1994

Description of amendments request: The proposed amendment will

change the Technical Specifications to modify the description of fuel

and control rod assemblies in TS 5.3.1, Fuel Assemblies. The change to

the fuel assembly description will permit the limited substitution of

zirconium alloy, zircaloy-4, ZIRLOTM, or stainless steel filler

rods for fuel rods in accordance with the NRC-approved applications of

fuel rod configurations that have been analyzed with NRC-approved

methods. This change will allow timely removal of fuel rods that are

found to be a probable source of future leakage. The change will make

provisions for the loading of lead test assemblies into the reactor

without requiring a specific TS change. This amendment also allows the

use of ZIRLOTM clad fuel as lead test assemblies. The specific

descriptions of the fuel and control rod assemblies contained in the TS

which are restrictive due to the unnecessary details are being deleted.

The change will also make line item improvements in the Technical

Specifications in accordance with Generic Letter 90-02, Supplement 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:

1. The proposed change to the Technical Specifications allowing

reconstitution will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because it will not result in a change to any of the process

variables that might initiate an accident or affect the radiological

release for an accident. The operating limits will not be changed

and the analysis methods to demonstrate operation within the limits

will remain in accordance with NRC-approved methodology. Other than

the changes to the fuel assemblies, there are no physical changes to

the plant associated with this Technical Specification change. The

consequences of an accident previously evaluated will not be

increased because the safety analysis to be performed for each cycle

will continue to demonstrate compliance with all fuel safety design

bases. The ability to remove potentially leaking fuel rods should

result in a reduction in the radiological consequences of any

transients or accidents.

The probability or consequences of an accident previously

evaluated are not significantly increased with the use of

ZIRLOTM cladding. The VANTAGE 5 fuel assemblies containing

ZIRLOTM clad fuel rods meet the same fuel assembly and fuel rod

design bases as other VANTAGE 5 fuel assemblies. In addition, the 10

CFR 50.46 criteria will be applied to the ZIRLOTM clad fuel

rods. The use of these fuel assemblies will not result in a change

to the proposed Farley VANTAGE 5 reload design and safety analysis

limits. Since the original design criteria are being met, the

ZIRLOTM clad fuel rods will not be an initiator for any new

accident. The ZIRLOTM clad material is similar in chemical

composition and has similar physical and mechanical properties as

that of zircaloy-4. Thus, the cladding integrity is maintained and

the structural integrity of the fuel assembly is not affected. The

ZIRLOTM clad fuel rod improves corrosion performance and

dimensional stability. No concerns have been identified with respect

to the use of an assembly containing a combination of both zircaloy-

4 and selected ZIRLOTM clad fuel rods. Since the dose

predications in the Farley safety analyses are not sensitive to the

fuel rod cladding material used, the radiological consequences of

accidents previously evaluated in the Farley safety analysis remain

valid. Therefore the probability or consequences of an accident

previously evaluated are not significantly increased.

The proposed removal of detailed descriptions of fuel and

control rod assemblies will not involve a significant increase in

the probability or consequences of an accident previously evaluated

because it will not result in a change to any of the process

variables that might initiate an accident. The operating limits will

not be changed and the analysis methods to demonstrate operation

within the limits will remain in accordance with NRC-approved

methodology. The consequences of an accident previously evaluated

will not be increased because the safety analyses to be performed

for each cycle will continue to demonstrate compliance with all fuel

safety design bases.

2. This change to the Technical Specifications allowing

reconstitution will not create the possibility of a new or different

kind of accident from any accident previously evaluated because it

will only affect the assembly configuration and will be limited to

NRC-approved applications of fuel rod configurations. The other

aspects of plant design, operation, limitations and responses to

events will remain unchanged.

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

accident previously evaluated is not created by the use of

ZIRLOTM cladding since the VANTAGE 5 fuel assemblies containing

ZIRLOTM clad fuel rods will satisfy the same design bases as

that used for other VANTAGE 5 fuel assemblies. All design and

performance criteria will continue to be met and no new single

failure mechanisms have been defined. In addition, the use of these

fuel assemblies does not involve any alterations to plant equipment

or procedures that would introduce any new or unique operational

modes or accident precursors. Therefore, the possibility for a new

or different kind of accident previously evaluated is not created.

The removal of detailed descriptions of fuel and control rod

assemblies will not create the possibility of a new or different

kind of accident from any accident previously evaluated because they

will be limited to NRC-approved applications of fuel rod

configurations. The other aspects of plant design, operation,

limitations and responses to events will remain unchanged.

3. The use of zirconium alloy, zircaloy-4, ZIRLOTM, or

stainless steel filler rods in fuel assemblies will not involve a

significant reduction in a margin of safety because analyses using

NRC-approved methods will be performed for each configuration to

demonstrate continued operation within the limits that assure

acceptable plant response to accidents and transients. These

analyses will be performed using NRC-approved methods that have been

approved for application to the fuel configuration.

The margin of safety is not significantly reduced by the use of

ZIRLOTM clad [sic] since the VANTAGE 5 fuel assemblies

containing ZIRLOTM clad fuel rods do not change the proposed

Farley VANTAGE 5 reload design and safety analysis limits. The use

of these fuel assemblies will take into consideration the normal

core operating conditions allowed for in the Technical

Specifications. For each cycle reload core, the fuel assemblies will

be evaluated using NRC Staff-approved reload design methods. This

will include consideration of the core physics analysis peaking

factors and core average linear heat rate effects. Therefore, the

margin of safety as defined in the bases to the Farley Technical

Specifications and VANTAGE 5 Licensing Amendment Request is not

significantly reduced.

The removal of detailed descriptions of fuel assemblies will not

involve a significant reduction in a margin of safety because

analyses using NRC-approved methods will be performed for each

configuration to demonstrate continued operation within the limits

that assure acceptable plant response to accidents and transients.

These analyses will be performed using NRC-approved methods that

have been approved for application to the fuel configuration.

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

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

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

amendment request involves no significant hazards consideration.

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

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

Attorney for licensee: James H. Miller, III, Esq., Balch and

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

Alabama 35201

NRC Project Director: S. Singh Bajwa

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

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

Date of amendment request: February 7, 1994 TS 93-19

Description of amendment request: The proposed change would revise

Technical Specification 5.3.1 to allow the substitution of filler rods

for fuel rods in fuel assemblies. This would permit the timely removal

of fuel rods that are found to be leaking or are determined to be the

probable source of future leaks.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification change

and has determined that it does not represent a significant hazards

consideration based on criteria established in 10 CFR 50.92(c).

Operation of Sequoyah Nuclear Plant (SQN) in accordance with the

proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The substitution of filler rods will be justified using NRC-

approved methodology. This methodology will demonstrate that the

existing design limits and safety analyses criteria are met.

Therefore, the proposed change does not increase the consequences of

an accident previously analyzed.

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

from any previously analyzed.

The proposed change involves the substitution of filler rods for

fuel rods. This substitution requires the utilization of NRC-

approved methodology. This methodology will ensure that the specific

analyses will not cause any new or different kind of accident from

that previously analyzed.

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

The substitution of filler rods for fuel rods would result in

less active fuel in the core. Therefore, the amounts of radiological

effluents that may be released offsite would not increase. The NRC-

approved methodology by which any reanalyses would be performed

already accounts for the affects on grid strength or the mass,

stiffness, and fundamental frequency of the fuel assembly during

seismic and loss-of-cooling accident conditions. Thus, the margin of

safety is not reduced when substituting filler rods for fuel rods.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

NuclearPlant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: February 7, 1994 TS 93-11

Description of amendment request: The proposed change would revise

Surveillance Requirement (SR) 4.7.9.i, ``Snubber Service Life

Program,'' to replace the present wording that describes the service

life hydraulic snubber monitoring and evaluation program with that from

the Westinghouse Electric Corporation Standard Technical

Specifications, Revision 4a. This would eliminate the need to perform

an engineering evaluation for drag-force increases of 50 percent or

greater of the previously measured value and substitute a requirement

to establish a monitoring program. This program would require that a

maximum service life for the snubber components be determined and the

monitoring program be established to ensure that the maximum service

life is not exceeded based on test results and failure history. A

proposed change to SR 4.7.9.c would remove the wording that is

inconsistent with Generic Letter 90-09 by removing the term ``if

applicable'' for performance of an as-found functional test and the

requirement related to tests of hydraulic snubbers that have uncovered

fluid ports.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

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

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

TVA proposes to delete the current TS requirements to perform an

evaluation of snubber test data when there is a greater than 50

percent increase in drag force for mechanical snubbers. The 50

percent evaluation requirement is considered unnecessary where

snubbers have small drag forces during their previous test. During

subsequent testing, small increases in drag forces (when compared

with the rated load of the snubber) may exceed 50 percent of the

previous test value. The relative change in drag force is small when

compared with the overall rating of the snubber; however, under the

current TS, an engineering evaluation for impending failure will

still be required. Eliminating the current evaluation requirement

from SQN's TSs will reduce the burden associated with performing

unnecessary evaluations. The proposed change is consistent with the

standard TS (Revision 4a), and a Snubber Service Life Program

continues to exist at SQN. Therefore, there is no increase in the

probability or consequences of an accident previously evaluated.

In addition, the testing language associated with Visual

Inspection Performance and Evaluation has been deleted to provide

consistency with Generic Letter 90-09. It should be noted that SQN

tests its hydraulic snubbers in either direction as necessary. This

is more conservative than the present TS requirement; therefore,

this change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any previously analyzed.

The two proposed changes involve deleting a requirement to

perform unnecessary analyses and a potentially nonconservative

testing requirement. These changes do not alter any plant operation,

maintenance requirements, or system design or function. Therefore, a

new or different kind of accident is not created by this proposed

change.

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

The proposed changes to the visual inspection and the service

life section will not modify the plant or revise its mode of

operation or the present safety analysis. The trending criteria to

be utilized provide adequate assurance that snubber impending

failure will be predicted in a timely manner. The deleted sections

will not change the requirement to test and trend data for snubbers

to predict failure; therefore, there is not a reduction in any

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: February 8, 1994 TS 93-14

Description of amendment request: The proposed change would revise

the setpoints in Technical Specification Table 3.3-4, ``Engineered

Safety Feature Actuation System Instrumentation,'' for the pressure

switches used to control switchover of the motor-driven Auxiliary

Feedwater pump suction from the normal condensate storage tank supply

to the essential raw cooling water supply. The setpoints would be

changed from the present trip setpoint of 2 psig and allowable value of

1 psig, to new values of 3.21 psig and 2.44 psig, respectively.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

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

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The auxiliary feedwater (AFW) system is designed to mitigate the

effects of the design basis accidents and anticipated operational

transients listed below:

A. Loss of normal feedwater

B. Loss of offsite power to station auxiliaries

C. Accidental depressurization in the main steam system

D. Rupture of a main steam line

E. Major rupture of a main feedwater pipe

F. Steam generator tube rupture

G. Small break loss of coolant accident

The AFW system only provides mitigation of the events listed

above and cannot initiate design-basis accident. Therefore, the

proposed change in the low-pressure setpoint of the motor-driven AFW

pump supply line will not result in an increase in the probability

of a previously analyzed accident. In addition, the proposed change

does not affect the overall water supply to the AFW system. Instead,

the proposed change results in a transfer from the condensate

storage tanks (CST) to the essential raw cooling water system at a

slightly higher CST water level, thus enhancing the continuous

supply of water. Therefore, this change will not result in an

increase in the consequences of a previously analyzed accident.

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

from any previously analyzed.

As discussed previously, the AFW system provides only mitigation

functions. In addition, the proposed change does not affect the

overall function and operation of the AFW system or its associated

water supplies. Instead, this change will provide additional

assurance of the proper operation of the AFW system. Therefore, the

proposed revision of the low-pressure setpoint of the motor-driven

AFW pump supply line will 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 TS bases for the AFW system require that AFW be available to

ensure that the reactor coolant system (RCS) can be cooled down to

less than 350 degrees Fahrenheit from normal operating conditions in

the event of a total loss of offsite power. In addition, the TS

bases for the CST require that a minimum water volume be available

to maintain the RCS at hot standby condition for two hours with

steam discharge to the atmosphere concurrent with a total loss of

offsite power.

The proposed TS revision does not affect the overall operation

of either the AFW system or the CST. The proposed setpoint revision

does slightly reduce the usable volume of water in the CST. However,

sufficient margin remains to ensure compliance with the bases of the

SQN TSs.

Therefore, the proposed changes to the SQN TSs do not involve a

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: February 8, 1994 TS 93-14

Description of amendment request: The proposed change would revise

the setpoints in Technical Specification Table 3.3-4, ``Engineered

Safety Feature Actuation System Instrumentation,'' for the pressure

switches used to control switchover of the motor-driven Auxiliary

Feedwater pump suction from the normal condensate storage tank supply

to the essential raw cooling water supply. The setpoints would be

changed from the present trip setpoint of 2 psig and allowable value of

1 psig, to new values of 3.21 psig and 2.44 psig, respectively.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

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

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The auxiliary feedwater (AFW) system is designed to mitigate the

effects of the design basis accidents and anticipated operational

transients listed below:

A. Loss of normal feedwater

B. Loss of offsite power to station auxiliaries

C. Accidental depressurization in the main steam system

D. Rupture of a main steam line

E. Major rupture of a main feedwater pipe

F. Steam generator tube rupture

G. Small break loss of coolant accident

The AFW system only provides mitigation of the events listed

above and cannot initiate design-basis accident. Therefore, the

proposed change in the low pressure setpoint of the motor-driven AFW

pump supply line will not result in an increase in the probability

of a previously analyzed accident. In addition, the proposed change

does not affect the overall water supply to the AFW system. Instead,

the proposed change results in a transfer from the condensate

storage tanks (CST) to the essential raw cooling water system at a

slightly higher CST water level, thus enhancing the continuous

supply of water. Therefore, this change will not result in an

increase in the consequences of a previously analyzed accident.

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

from any previously analyzed.

As discussed previously, the AFW system provides only mitigation

functions. In addition, the proposed change does not affect the

overall function and operation of the AFW system or its associated

water supplies. Instead, this change will provide additional

assurance of the proper operation of the AFW system. Therefore, the

proposed revision of the low-pressure setpoint of the motor-driven

AFW pump supply line will 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 TS bases for the AFW system require that AFW be available to

ensure that the reactor coolant system (RCS) can be cooled down to

less than 350 degrees Fahrenheit from normal operating conditions in

the event of a total loss of offsite power. In addition, the TS

bases for the CST require that a minimum water volume be available

to maintain the RCS at hot standby condition for two hours with

steam discharge to the atmosphere concurrent with a total loss of

offsite power.

The proposed TS revision does not affect the overall operation

of either the AFW system or the CST. The proposed setpoint revision

does slightly reduce the usable volume of water in the CST. However,

sufficient margin remains to ensure compliance with the bases of the

SQN TSs.

Therefore, the proposed changes to the SQN TSs do not involve a

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: February 9, 1994 TS 93-21

Description of amendment request: The proposed change would revise

Technical Specification Table 3.3-11, ``Fire Detection Instruments,''

by adding one detector to Fire Zones 184, 185, 186, and 187 for each

Unit. These fire zones are located in the 6.9 kv shutdown board room

corridors in the auxiliary building.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in

accordance with the proposed amendment will not:

1.Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to the fire detection instrumentation adds

two additional cross-zone detectors in each of the Units 1 and 2

6900-volt shutdown board room corridors on Elevation 734 of the

auxiliary building. The additional fire detection instrumentation

provides additional assurance that the fire

1detection instrumentation will operate as required in the event

of a fire. However, neither the fire detection instrumentation nor

the equipment associated with this instrumentation is considered to

be the source of an accident. In addition, this equipment is not

taken credit for in the safety analysis. Therefore, there is no

increase in the probability or consequences of a previously

evaluated accident.

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

from any previously analyzed.

The fire detection and/or suppression functions affected by this

change enhance fire mitigation functions only and do not result in a

change in plant functions. Therefore, the proposed change does not

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

any preciously analyzed.

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

The equipment functions affected by the proposed changes are not

assumed for any accident in the SQN safety analysis and are not an

input to the TS margin of safety. Therefore, the proposed change

will not result in a reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library,1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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 31, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.1.1.2 to permit the reduction of

boron concentration of water within the reactor coolant system (RCS),

subject to certain restrictions, when the reactor is in Mode 5 and RCS

flow is less than 2800 gpm. The proposed amendment is related to

Amendment No. 176, which was issued by the NRC on December 8, 1992, and

incorporated a similar revision for Mode 6 operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:The Nuclear Regulatory

Commission has provided standards in 10 CFR 50.92(c) for determining

whether a significant hazard exists due to a proposed amendment to an

Operating License for a facility. A proposed amendment involves no

significant hazards if operation of the facility in accordance with the

proposed changes would: (1) Not involve a significant increase in the

probability or consequences of an accident previously evaluated; (2)

Not create the possibility of a new or different kind of accident from

any accident previously evaluated; or (3) Not involve a significant

reduction in a margin of safety. 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 Number 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,

conditions or assumptions are significantly affected by the proposed

changes. The proposed change to Technical Specification (TS) 3/

4.1.1.2 would revise an exception to make it applicable in Mode 5 as

well as Mode 6. The revised exception would allow water of a lower

boron concentration than the Reactor Coolant System (RCS) to be

added to the RCS, with the flowrate of reactor coolant through the

RCS less than 2800 gpm, provided that the water to be added meets

the requirements of TS 3.1.1.1 (Mode 5) or TS 3.9.1 (Mode 6). TS

3.1.1.1 requires that in Mode 5, the boron concentration of the RCS

be maintained such that the Shutdown Margin shall be less than or

equal to one percent delta k/k. TS 3.9.1 requires that in Mode 6,

the boron concentration of all filled portions of the RCS and the

refueling canal shall be maintained uniform and sufficient to ensure

that the more restrictive of two reactivity conditions is met. If

the RCS meets these reactivity condition requirements, and water is

added to the RCS that also meets the reactivity condition

requirements of TS 3.1.1.1 or TS 3.9.1, then the RCS is assured to

remain in compliance with the reactivity condition requirements. The

possibility that the added water may be of lower boron concentration

than the RCS, therefore, does not significantly increase the

probability of an accident previously evaluated.

The proposed change to TS 3/4.9.8.1 makes TS 3/4.9.8.1 and TS 3/

4.9.8.2 consistent with the current TS 3/4.1.1.2, and is considered

to be administrative in nature.

The proposed changes to TS Bases 3/4.1.1.2 and TS Bases 3/4.9.8

are considered to be administrative in nature.

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. As discussed in

item 1a. above, the proposed revision of the exception to TS 3/

4.1.1.2 will not cause a condition that would result in the RCS not

meeting the requirements of TS 3.1.1.1 or TS 3.9.1, as applicable.

The proposed changes do not alter the source term, containment

isolation, or allowable releases. The proposed changes, therefore,

will not increase the radiological consequences of a previously

evaluated accident. As also discussed in item 1a. above, the

proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases

3.4.9.8 are considered to be administrative in nature.

2a. Not create the possibility of a new kind of accident from

any accident previously evaluated because no new accident initiators

or assumption are introduced by the proposed changes. The proposed

changes do not alter any accident scenarios. As discussed in item

1a. above, the proposed revision of the exception to TS 3/4.1.1.2

will not cause a condition that would result in the RCS not meeting

the requirements of TS 3.1.1.1 or TS 3.9.1. The proposed changes to

TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases 3/4.9.8 are

considered to be administrative in nature. None of the proposed

changes creates the possibility of a new kind of accident.

2b. Not create the possibility of a different kind of accident

from any accident previously evaluated because no different accident

initiators or assumptions are introduced by the proposed changes.

The proposed changes do not alter any accident scenarios. As

discussed in item 1a. above, the proposed revision of the exception

to TS 3/4.1.1.2 will not cause a condition that would result in the

RCS not meeting the requirements of TS 3.1.1.1 or TS 3.9.1. The

proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases

3/4.9.8 are considered to be administrative in nature. None of the

proposed changes creates the possibility of a different kind of

accident from any accident previously evaluated.

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

because the proposed change to TS 3/4.1.1.2, as described above,

will not cause a condition that would result in the RCS not meeting

the requirements of TS 3.1.1.1 or TS 3.9.1. The margin of safety

will be maintained by adhering to the limits specified in these TSs.

The proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1 and TS

Bases 3/4.9.8 are considered to be administrative in nature.

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: John N. Hannon

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: July 14, 1993

Description of amendment request: The proposed amendment would

modify Sections 3.6 and 4.6 of the Technical Specifications to add

Reactor Coolant System leakage detection requirements to address

Generic Letter 88-01, ``NRC Position on Intergranular Stress Corrosion

Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1.The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed amendment would add a more conservative

requirement into the plant Technical Specifications, in addition to

those that presently exist. Hence, approval of this change will have

no affect on any previously evaluated accident scenario.

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

new or different kind of accident from any accident previously

evaluated. No physical changes are being made to the plant and now

new operating techniques or procedures are being proposed. The

proposed amendment would add an additional Limiting Condition for

Operation and an increased Surveillance

Requirement to plant Technical Specifications. Hence, approval

of this change will not create the possibility of a new or different

kind of accident.

3. The proposed amendment will not involve a significant

reduction in a margin of safety. The proposed change adds more

restrictive requirements into the Technical Specifications. Hence,

approval of this change would not reduce the margin of

safety.GI21The Commission has also provided guidance concerning the

application of these standards by providing certain examples (March

6, 1986, 51FR7751). An example of an amendment that is considered

not likely to involve a significant hazards consideration is Example

(ii) which is an additional limitation, restriction or control not

presently included in the Technical Specifications. This proposed

amendment provides for an additional Limiting Condition of Operation

and an increased Surveillance Requirement in the plant Technical

Specifications. Therefore, based on the above, it is determined this

change does not constitute a significant hazards consideration as

defined in 10CFR 50.92(c).

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: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624

NRC Project Director: Walter R. Butler

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

Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

Date of amendment request: December 27, 1993

Description of amendment request: The proposed change would revise

the Technical Specifications (TS) for the Surry Power Station, Units

No. 1 and No. 2 (SPS-1&2). The proposed changes revise the review

responsibilities of the Station Nuclear Safety and Operating Committee

(SNSOC) and the Management Safety Review Committee (MSRC).

The SPS-1&2 TS address the organization and responsibilities of

both the onsite and offsite review groups: SNSOC and MSRC,

respectively. The responsibilities of the SNSOC include the review of

new procedures and changes to procedures that affect nuclear safety.

The MSRC review responsibilities include the review of safety

evaluations and SNSOC meeting minutes and reports. The extent of these

review activities would be revised by the proposed changes to ensure

the two review groups are focusing on nuclear safety issues and not

spending an unnecessary amount of time on activities of minimal safety

significance. Specifically, the proposed changes would revise the

review responsibilities of SNSOC regarding procedure changes. Rather

than reviewing all procedure changes, SNSOC would only review procedure

changes that require a safety evaluation. The proposed changes also

would revise the review responsibilities of the MSRC. Rather than

reviewing all of the safety evaluations and SNSOC meeting minutes and

reports as presently required by the TS, the MSRC would only review a

representative sample of these documents.

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:

[T]he elimination of the SNSOC review of procedure changes that

do not require a safety evaluation, revising the wording for

approval of procedure changes, and the modification of the MSRC's

duties regarding their review of safety evaluations and SNSOC

meeting minutes and reports will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. As administrative

changes, the proposed Technical Specifications changes have no

direct or indirect effect on accident precursors. No plant

modifications are being implemented and operation of the plant is

unchanged. SNSOC review of new procedures and procedure changes that

require a safety evaluation ensures that activities that could

affect nuclear safety are being properly reviewed. The MSRC's

overview of representative samples of safety evaluations and SNSOC

meeting minutes and reports based on performance ensures these

programs are being properly implemented and nuclear safety is not

being compromised; or

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

from any accident previously evaluated since physical modifications

are not involved and systems and components will be operated as

before the change. The proposed changes are wholly administrative in

nature and have no impact on plant operations or accident

considerations. These changes modify the scope of SNSOC review of

procedure changes and MSRC's review functions concerning safety

evaluations and SNSOC meeting minutes and reports. Procedure changes

will continue to receive management review in accordance with

administrative procedures, however, only changes that require a

safety evaluation will require SNSOC approval. MSRC review of

representative samples of safety evaluations and SNSOC meeting

minutes and reports based on performance will continue to provide

adequate assurance that nuclear safety is being properly considered;

or

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

defined in the basis of any Technical Specification since the

responsibilities of the SNSOC and MSRC are not addressed by the

existing Technical Specification Bases, nor are review requirements

for procedures. The proposed changes are administrative in nature

and have no impact on, nor were they considered in, existing UFSAR

accident analyses. Safety significant procedure changes, i.e.,

changes that require a safety evaluation to be prepared, will

continue to be reviewed by SNSOC, as will new procedures. Procedure

changes still require cognizant management approval and preparation

of an activity screening to determine whether or not the change

impacts nuclear safety. This ensures activities important to nuclear

safety are being appropriately reviewed. The effectiveness of the

safety evaluation program, and the thoroughness of SNSOC meetings

and reports will be assured through the MSRC's plant overview

function which is based on observed performance.

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, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Herbert N. Berkow

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

Project No. 2, Benton County, Washington

Date of amendment request: December 6, 1993

Description of amendment request: The proposed amendment would

modify the testing requirements for the Main Steam Relief Valves

(MSRVs) in the Technical Specifications (TS). Specifically, the

proposed amendment would allow deferral of MSRV Position Indication

Channel Calibration, including the Channel Functional Test that is the

focus of the request, for 24 hours after the plant reaches conditions

that would allow the Channel Functional Test to be conducted under

operating conditions (above 10% rated reactor power). In addition, the

proposed change would extend the current deferral for two related TS,

one for MSRVs and the other for Automatic Depressurization System (ADS)

valves, from 12 hours to the proposed 24 hours after reaching

conditions that would allow the Channel Functional Test to be conducted

under operating 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. The staff's evaluation of the licensee's analysis of the

change that would defer Safety/Relief Valve Position Indicators Channel

Calibration, specifically the Channel Functional Test, is presented

below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

No credit is taken for the MSRV position indication in the

initiation or mitigation of any analyzed accident. The inoperability of

valve position indication does not affect the manual or automatic

actuation of the MSRVs. The analysis for inadvertent opening of an MSRV

(FSAR Section 15.1.4) assumes that the alarm function of the MSRV

discharge line temperature sensors and Reactor Pressure Vessel (RPV)

level control systems provide the signals for manual and automatic

system actuation, respectively. Therefore, because previously analyzed

accidents are mitigated without the use of the MSRV position

indication, this change that would allow temporary plant operation in

modes 1 and 2 with inoperable MSRV position indication to allow testing

does not increase the probability or consequences of a previously

analyzed accident.

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

kind of accident from any accident previously evaluated?

No new mode of operation of any equipment results from the delay of

the Channel Functional surveillance. The valve position indication

provides information for operator response to previously evaluated

accidents, and does not provide any automatic system actuations that

could initiate an accident or abnormal operating occurrence sequence.

This change would not, therefore, create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

of safety?

The margin of safety involved in this amendment is the time it

takes to identify inadvertent MSRV operation to initiate either

automatic or manual plant response. Since no credit is taken for MSRV

position indication in the WNP-2 safety analyses for initiation of

automatic or operator manual response, the lack of MSRV position

indication for a 24 hour time period will not affect the analyzed time

to identify inadvertent MSRV operation. The proposed change does not,

therefore, affect the margin of safety.

The staff's evaluation of the licensee's analysis of the change

that would allow an increase in the current deferral of Safety/Relief

Valve and Automatic Depressurization System (ADS) valve testing of 12

hours to the proposed 24 hours is presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change represents an additional 12 hours in which the

operability of the MSRV position indication would not be verified, and

could be inoperable. No credit is taken for the MSRV position

indication in the initiation or mitigation of any analyzed accident.

The inoperability of valve position indication does not affect the

manual or automatic actuation of the MSRVs as discussed in the

following: The analysis for inadvertent opening of an MSRV (FSAR

Section 15.1.4) assumes the alarm function of the MSRV discharge line

temperature sensors and RPV level control systems for manual and

automatic system actuation, respectively. Thus, the inoperability of

the MRSV position indication for an additional 12 hours would not

affect the probability or consequences of an accident previously

evaluated.

The proposed change represents an additional 12 hours in which the

operability of the ADS valves, which are also the MSRVs, would not be

verified, and could be inoperable. The cause of any potential

inoperability would also be undetermined, and any unidentified failure

mode affects the uncertainties assumed in the probability of

inadvertent MSRV opening, which is an analyzed accident (FSAR Section

15.1.4). In the worst case, the probability of inadvertent MSRV opening

would be increased due to an unidentified problem with the MSRVs. The

additional 12 hours, considering plant operation for approximately 5800

hours of power operation each year (assuming 60 day annual refueling

outage and 80% capacity factor for the remaining time), represents only

an estimated 0.2% increase in time at power without verifying

operability of the ADS valves, which contributes a very small increase

in the probability of an inadvertent MSRV opening due to an

undiscovered fault condition. Performing the surveillance testing using

other possible methods, either removal of the MSRV position indication

from the valve and testing separately, or testing in modes 3, 4, or 5

such that adequate steam back pressure does not exist, introduce other

uncertainties and potential for valve damage that would, using

engineering judgement, create a greater increase in probability of an

inadvertent MSRV opening than the additional 12 hours would contribute.

The proposed change would not, therefore, significantly increase the

probability of an accident previously evaluated.

Regarding the potential effect on the consequences of an

inadvertent MSRV opening, the accident analysis in FSAR Section 15.1.4

analyzes consequences assuming an MSRV sticks open. Thus, increasing

the time that a potential malfunction of a valve may go undetected does

not affect the consequences of an MSRV that is already considered open

as assumed in the accident analysis. Consequently, the increase in time

allowed for conducting the surveillance testing, and therefore the time

that the operability of the ADS-related MSRV is not verified does not

affect the consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

No new or modified mode of operation of any structure, system, or

component results from delaying verification of the status of

operability of the MSRVs for an additional 12 hours. The proposed

change does not, therefore, create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

of safety?

The margin of safety involved in this proposed amendment is the

time, established by the current TS, that operation in modes 1 and 2 is

allowed with the operability of the MSRVs and associated ADS valves

undetermined. The current TS allow 12 hours to conduct testing to

verify operability of these valves. The proposed TS would extend that

time to 24 hours. The additional 12 hours, considering plant operation

for approximately 5800 hours of power operation each year (assuming 60-

day annual refueling outage and 80% capacity factor for the remaining

time), represents only an estimated 0.2% increase in time at power

without verifying operability of the ADS valves, which the staff

considers, by engineering judgement, as a very small reduction in the

margin to safety affected by the proposed change. This change would,

therefore, not involve a significant reduction in the margin of safety

provided by the current TS.

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

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

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

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

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

NRC Project Director: Theodore R. Quay

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: January 26, 1994

Description of amendment request: The proposed amendments would

change Technical Specification Section 15.3.0, ``General

Considerations.'' This section specifies the actions which should be

taken for conditions not directly addressed in the action statements of

the Technical Specifications. The changes would provide more

operational flexibility. Changes to the applicable bases and editorial

changes are also proposed.

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 this facility under the proposed Technical

Specifications change will not create a significant increase in the

probability or consequences of an accident previously evaluated.

This proposed change will incorporate requirements contained in

NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved

Technical Specifications,'' into the Point Beach Technical

Specifications Section 15.3.0, ``General Considerations.'' The

proposed revisions will not remove any existing requirements.

Several new requirements will be added. However, the proposed

revisions will allow a longer period of time for the affected

unit(s) to be placed in hot shutdown should one of the applicable

Limiting Conditions for Operation not be met. This longer time limit

is identical to the time limit specified in NUREG 1431, Revision 0,

and permits the shutdown to proceed in a controlled and orderly

manner that is well within the capabilities of the unit(s), assuming

that only the minimum required equipment is operable. This reduces

thermal stresses on components of the Reactor Coolant System and the

potential for a plant transient that could challenge plant safety

systems. The amount of time to reach cold shutdown would decrease

from 48 hours to 37 hours. The slightly longer time to reach hot

shutdown is more than offset by the shorter time to cold shutdown,

thereby reducing the total amount of time a unit may be operated in

a condition in which a system or component required to mitigate the

consequences of an accident is unavailable, or that is otherwise

prohibited by the specifications.

Should a shutdown of both units be required, 15.3.0.A and

15.3.0.B allow an orderly and sequential shutdown of each unit to

take place. This ensures that the plant operators are not overloaded

during the shutdown process and allows the units shutdowns to

proceed in a controlled and orderly manner. The revised

specifications will clarify the requirements, enhancing the

effectiveness of the Point Beach Technical Specifications. There is

no physical change to the facility, its systems, or its operation.

Thus, an increased probability or consequences of an accident

previously evaluated cannot occur.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This proposed change will incorporate requirements contained in

NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved

Standard Technical Specifications,'' into the Point Beach Technical

Specifications Section 15.3.0, ``General Considerations.'' The

proposed revisions will not remove any existing requirements.

Several new requirements will be added.

However, the proposed revisions will allow a longer period of

time for the affected unit(s) to be placed in hot shutdown should

one of the applicable Limiting Conditions for Operation not be met.

This longer time limit is identical to the time limit specified in

NUREG 1431, Revision 0, and permits the shutdown to proceed in a

controlled and orderly manner that is well within the capabilities

of the unit(s), assuming that only the minimum required equipment is

operable. This reduces thermal stresses on components of the Reactor

Coolant System and the potential for a plant transient that could

challenge plant safety systems. The amount of time to reach cold

shutdown would decrease from 48 hours to 37 hours. The slightly

longer time to reach hot shutdown is more than offset by the shorter

time to cold shutdown, thereby reducing the total amount of time a

unit may be operated in a condition in which a system or component

required to mitigate the consequences of an accident is unavailable,

or that is otherwise prohibited by the specifications.

Should a shutdown of both units be required, 15.3.0.A and

15.3.0.B allow an orderly and sequential shutdown of each unit to

take place. This ensures that the plant operators are not overloaded

during the shutdown process and allows the units shutdown to proceed

in a controlled and orderly manner. The revised specifications will

clarify the existing specifications and will add some additional

requirements, enhancing the effectiveness of the Point Beach

Technical Specifications. There is no physical change to the

facility, its systems, or its operation. Thus, a new or different

kind of accident cannot occur.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety. This proposed change will incorporate requirements

contained in NUREG 1431, Revision 0, ``Westinghouse Owner's Group

Improved Standard Technical Specifications,'' into the Point Beach

Technical Specifications Section 15.3.0, ``General Considerations.''

The proposed revisions will not remove any existing requirements.

Several new requirements will be added.

However, the proposed revisions will allow a longer period of

time for the affected unit(s) to be placed in hot shutdown should

one of the applicable Limiting Conditions for Operation not be met.

This longer time limit is identical to the time limit specified in

NUREG 1431, Revision 0, and permits the shutdown to proceed in a

controlled and orderly manner that is well within the capabilities

of the unit(s), assuming that only the minimum required equipment is

operable. This reduces thermal stresses on components of the Reactor

Coolant System and the potential for a plant transient that could

challenge plant safety systems. The amount of time to reach cold

shutdown would decrease from 48 hours to 37 hours. The slightly

longer time to reach hot shutdown is more than offset by the shorter

time to cold shutdown, thereby reducing the total amount of time a

unit may be operated in a condition in which a system or component

required to mitigate the consequences of an accident is unavailable,

or that is otherwise prohibited by the specifications.

Should a shutdown of both units be required, 15.3.0.A and

15.3.0.B allow an orderly and sequential shutdown of each unit to

take place. This ensures that the plant operators are not overloaded

during the shutdown process and allows the units shutdowns to

proceed in a controlled and orderly manner. The revised

specifications will clarify the existing specifications and will add

some additional requirements, enhancing the effectiveness of the

Point Beach Technical Specifications. There is no physical change to

the facility, its systems, or its operation. Thus, a significant

reduction in a margin of safety cannot occur.

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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

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

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

NRC Project Director: John N. Hannon

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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: November 4, 1993

Description of amendments request: The proposed amendments would

allow the removal of an orifice plate in the containment vent/purge

line to allow greater flow through the line. The restoration of full

flow capability will result in less time required to vent the

containment. A reanalysis of the maximum hypothetical accident, as

currently described in the Updated Final Safety Analysis Report, was

performed to support the requested amendments. The results of the

reanalysis indicate that the consequences of the accident previously

analyzed would be increased. Although the consequences result in an

increase in the fission product release, the total doses are well

within the limits of 10 CFR Part 100, ``Factors to be considered when

evaluating sites.''Date of Publication of Individual Notice in Federal

Register: February 25, 1994 (59 FR 9254)

Expiration date of individual notice: March 28, 1994

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: Robert A. Capra

Duke Power Company, Docket No. 50-414, Catawba Nuclear Station,

Unit No. 2, York County, South Carolina

Date of amendment request: January 10, 1994

Description of amendment request: The proposed amendment would

change the method of measuring the reactor coolant system flow rate

(Technical Specification 2.0 and 3/4.2) during the 18-month

surveillance for Catawba, Unit 2. Date of publication of individual

notice in Federal Register: March 1, 1994 (59 FR 9785)

Expiration date of individual notice: March 31, 1994

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730

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

Creek Generating Station, Coffey County, Kansas

Date of application for amendment: February 7, 1994

Brief description of amendment request: The proposed amendment

would allow an increase in reactor coolant temperature in order to

support operation at the rated thermal power of 3565 megawatts thermal

(MWt). The proposed amendment would change reactor protection system

setpoints by increasing the nominal reactor coolant average temperature

from 581.2 deg.F to 586.5 deg.F, changing the axial flux difference

penalties and setpoint uncertainty allowances. The proposed amendment

also increases the maximum indicated reactor coolant system average

temperature from 585.0 deg.F to 590.5 deg.F.

Date of individual notice in Federal Register: February 15, 1994

(59 FR 7269)

Expiration date of individual notice: March 17, 1994

Local Public Document Room location: Emporia State University,

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

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

66621

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with

these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC 20555, and at the local public document

rooms for the particular facilities involved.

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

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

Date of application for amendment: November 1, 1993, as

supplemented on February 1, 1994

Brief description of amendment: The amendment revises the heatup

and cooldown curves which will allow operation beyond the current 12

effective full-power years (EFPY) to approximately 13.8 EFPY. The

increase in this EFPY will allow Unit 2 to operate until its next

refueling outage (RFO-10) in accordance with the requirements of 10 CFR

Part 50, Appendix G.

Date of issuance: March 1, 1994

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

30 days.

Amendment No.: 162

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

Technical Specifications.

Date of initial notice in Federal Register: November 24, 1993 (58

FR 62150)The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

North Carolina

Date of application for amendment: August 27, 1993, as supplemented

November 10, 1993 and February 1, 1994

Brief description of amendment: The amendment revises the Technical

Specifications and allows elimination of the existing reactor coolant's

resistance temperature detector (RTD) bypass manifold system and the

substitution of a new design with RTDs mounted in thermowells that

extend directly into the flow stream of the reactor coolant system.

Date of issuance: February 18, 1994

Effective date: February 18, 1994

Amendment No. 43

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

Technical Specifications.

Date of initial notice in Federal Register: September 15, 1993 (58

FR 48379)The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 18, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

North Carolina

Date of application for amendment: July 16, 1993, as supplemented

February 17, 1994

Brief description of amendment: The amendment revises the Technical

Specifications (TS) 3/4.2.1, Axial Flux Difference, 3/4.2.2, Heat Flux

Hot Channel Factor (FQ), deletes surveillance requirement 4.2.2.2

and 4.2.2.3, and changes 6.9.1.6, Core Operating Limit Report, and

associated Bases related to the transition from nuclear fuel supplied

by Westinghouse to nuclear fuel supplied by Siemens Power Corporation

(SPC) beginning with Cycle No. 6.

Date of issuance: March 1, 1994

Effective date: March 1, 1994

Amendment No. 44

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

Technical Specifications.

Date of initial notice in Federal Register: August 4, 1993 (58 FR

41500)The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 1, 1994.No significant hazards

consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

North Carolina

Date of application for amendment: May 15, 1993, as supplemented

February 17 and February 25, 1994

Brief description of amendment: The amendment modifies the SHNPP

Operating License to provide for a one-time exemption from compliance

with License Condition 2.C.(8) which requires periodic emergency diesel

generator engine teardowns for component inspections.

Date of issuance: March 3, 1994

Effective date: March 3, 1994

Amendment No. 45

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

Operating License Condition 2.C.(8) as defined in Attachment 1 to

Operating License NPF-63.

Date of initial notice in Federal Register: June 9, 1993 (58 FR

32378) The February 17, 1994, letter provided supplemental information

and the February 25, 1994, letter modified the May 15, 1993, letter and

requested a one-time exemption. Neither supplemental letter changed the

initial proposed determination of no significant hazards consideration

as published in the Federal Register.The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

March 3, 1994.No significant hazards consideration comments received:

No

Local Public Document Room location: Cameron Village Regional

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

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

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

North Carolina

Date of application for amendment: July 16, 1993, as supplemented

February 17, 1994

Brief description of amendment: The amendment revises the SHNPP

Technical Specification to incorporate changes to reactor core safety

limits, reactor trip system instrumentation setpoints, power

distribution limits, and shutdown boron concentration control in

support of the transition from nuclear fuel supplied by Westinghouse

Electric Corporation to nuclear fuel supplied by Siemens Power

Corporation and a reactor core safety average temperature reduction

effort.

Date of issuance: March 3, 1994

Effective date: March 3, 1994

Amendment No. 46

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

Technical Specifications.

Date of initial notice in Federal Register: September 29, 1993 (58

FR 50966) The February 17, 1994, letter provided clarifying information

that did not change the initial determination of no significant hazards

consideration as published in the FEDERAL REGISTER.The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated March 3, 1994. No significant hazards consideration comments

received: No

Local Public Document Room location: Cameron Village Regional

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

Commonwealth Edison Company, 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 13, 1993, as

supplemented by letters dated September 15, 1993, September 16, 1993,

December 17, 1993, January 19, 1994, February 11, 1994, and February

24, 1994.

Brief description of amendments: These amendments revise Technical

Specification 3/4.4.5, ``Steam Generators,'' to allow sleeving of

defective steam generator tubes as an alternative to tube plugging. Two

different methods of sleeving are approved for Byron and Braidwood

stations: Westinghouse laser-welded tube sleeving and Babcock and

Wilcox (B&W) kinetic-welded tube sleeving.

Date of issuance: March 4, 1994

Effective date: March 4, 1994

Amendment Nos.: 47, 47, 59, and 59

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: October 27, 1993 (58 FR

57846) The additional information contained in the supplemental letters

dated September 15, 1993, September 16, 1993, December 17, 1993,

January 19, 1994, February 11, 1994, and February 24, 1994, served to

clarify the amendments, were within the scope of the initial notice,

and did not affect the Commission's proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated March 4, 1994.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. 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: September 2, 1993, as

supplemented by letters dated January 7, 1994, and February 10, 1994.

Brief description of amendments: The amendments revise the Byron

and Braidwood Technical Specifications (TS) to allow replacement of the

125 volt DC Gould batteries with new 125 volt DC AT&T batteries and

rephrase the specification for their design duty cycle. In addition,

the amendments revise the crosstie loading limitations and crosstie

breaker limitations. The associated TS Bases are also revised to

include the purpose for the crosstie limitations and a discussion of

the design duty cycle requirements.

Date of issuance: March 4, 1994

Effective date: March 4, 1994

Amendment Nos.: 59, 59, 47, and 47

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: February 2, 1994 (59 FR

4936) The February 10, 1994, supplemental submittal did not change the

original no significant hazards consideration determination. The

Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated March 4, 1994. 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. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: December 15, 1989, as

supplemented September 16, 1992.

Brief description of amendments: The amendments revise Technical

Specifications (TSs) 3.0.4, 4.0.3, and 4.0.4 to incorporate guidance

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

the Standard Technical Specifications (STS) on the Applicability of

Limiting Conditions for Operation and Surveillance Requirements.''

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