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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission

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

notice. Pub. L. 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 January 4, 1999, through January 14, 1999.

The last biweekly notice was published on January 13, 1999.

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By February 26, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

scheduled in the proceeding, but such an amended

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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 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,

Units 1 and 2, Rock Island County, Illinois.

Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and

2, LaSalle County, Illinois.

Date of application for amendment request: December 17, 1998.

Description of amendment request: The amendments would revise the

respective facility Technical Specifications (TS) by adding a new

Limiting Conditions for Operations which provides an administrative

enhancement by allowing testing required to return equipment to service

to be conducted under administrative controls.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change has no impact on the design basis of the

plant. The change has no impact on the response of the plant during

normal or transient conditions. Incorporation of ISTS [improved

Standard Technical Specification] 3.0.5 provides the necessary

administrative controls that allow the return of equipment to

service to complete testing required to demonstrate operability.

Without this allowance, certain components could not be restored to

operable status and a plant shutdown would ensue. It is not the

intent of the TS to preclude the return to service of a component in

order to confirm its operability or the operability of other

equipment. This allowance is deemed to be a safer operation than

requiring a plant shutdown to complete necessary testing. This

allowance is considered acceptable because it: (1) is temporary; (2)

accompanied by appropriate administrative controls, and; (3)

provides a safety enhancement by restoring the plant status to, or

confirming the existing plant status is in, a condition that is

expected to provide for safe operation.

ISTS 3.0.5 was adopted to address the ambiguity that ACTION

requirements do not strictly allow the restoration of equipment to

its normal configuration to perform functional testing required to

demonstrate operability. The components involved will have completed

maintenance and or testing that will demonstrate, with reasonable

assurance, that the component can perform its intended safety

function.

Therefore, the proposed changes do 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 accident previously evaluated:

The proposed changes do not introduce new features or modify

plant structures, systems or components that may impact station

operations under normal or abnormal conditions. The proposed changes

will allow the necessary testing to ensure safety related equipment

will perform its design basis safety function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

the following reasons:

The proposed changes have no impact on any of the Safety Limits

provided in the Technical Specifications, nor does the change impact

the operation of structures, systems and components import to plant

safety. The purpose of the proposed change is to return equipment to

service, under administrative controls, to complete operability

testing. Therefore, allowing the return of equipment to service will

promote timely restoration of, or confirmation of, equipment

operability thereby increasing the margin of safety from that

existing with this equipment remaining out of service. Temporarily

returning inoperable equipment to service for the purpose of

confirming operability places the plant in a condition which has

been previously evaluated and determined to be acceptable for short

periods. Therefore, the proposed change does not involve a

significant reduction in safety.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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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

amendments requested involve no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021; for LaSalle, Jacobs Memorial Library, 815 North Orlando

Smith Avenue, Illinois Valley Community College, Oglesby, Illinois

61348-9692.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Project Director: Stuart A. Richards.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: December 24, 1998.

Description of amendment request: These amendment requests change

the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2)

Technical Specifications (TSs) to ensure that Emergency Diesel

Generator (EDG) requirements contained in Technical Specification 3/

4.8.1 for both units are consistent with assumptions contained in

design analyses and requirements of plant procedures. Revisions to TS

3/4.8.1 ``A.C. Sources,'' contained in this amendment provide more

conservative limiting conditions for operation (LCO) and surveillance

requirements that affect EDG fuel oil storage volume, EDG load

rejection and overspeed testing, and EDG operating frequency

requirements. The applicable bases for each unit are also refined, as

necessary, to strengthen the explanations regarding EDG fuel oil

storage systems and provide the EDG overspeed in terms of frequency

(Hertz) and speed (Revolutions Per Minute).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The addition of the term ``usable'' to LCO 3.8.1.1 and 3.8.1.2

for both Units will assure that the required quantity of fuel oil

will be available to operate the diesel during emergency conditions.

This revision including the discussion contained in the Technical

Specification Bases has no physical impact on the diesels or their

setpoints. These revisions also do not delete any function

previously provided by the diesels. There are no design bases

accidents for which failure of the diesel is considered an

initiating event. Therefore, the probability of an accident

previously evaluated in the safety analysis is not increased by this

change. The proposed changes do not involve an increase in the

consequences of an accident previously analyzed, as they make the

limiting condition for operation and associated bases more

conservative and involve no physical changes to the diesels.

The revised EDG single largest load rejection and overspeed

criteria do not involve an increase in the probability or the

consequences of accidents previously analyzed. The surveillance

tests impacted by the proposed revision are performed only during

shutdown when the opposite train EDG and its connected AC power

system are relied upon as the emergency AC power source. Further,

there are no design basis accidents for which changes to EDG load

rejection test acceptance criteria can be an initiating event. The

proposed changes affect the diesel testing requirements but do not

affect the operating or design parameters. The changes also do not

affect the diesels' ability to mitigate the consequences of an

accident. They serve to ensure the ability of the diesel to reject

the largest load. The overspeed criteria ensures that diesel

frequency does not exceed a certain value subsequent to a load

rejection. This criteria also ensures compliance with the guidance

of Safety Guide 9 for Unit 1 and Regulatory Guide 1.9 for Unit 2. It

does not involve an increase in the consequences of an accident

previously analyzed. The revision does not impact accidents

previously analyzed and would not, therefore, affect the

consequences of accidents previously analyzed.

Revising the EDG operating frequency as discussed in the

proposed amendment protects [engineered safety feature] ESF pumps

from runout conditions and motors from operating in an unanalyzed

condition. The narrower frequency limits are more restrictive and

have no adverse effect on the diesel generator operability. The

proposed revision to decrease the EDG operating frequency limit does

not involve an increase in the probability of an accident as

described in the [Updated Final Safety Analysis Report] UFSAR. There

are no design basis accidents for which failure of the diesel is

considered an initiating event. A narrower operating frequency does

not increase the probability of a design basis accident; it ensures

that equipment performs their intended function. This change is

intended to prevent the diesel from being loaded beyond analyzed

loading limits and protect ESF equipment. The more conservative

surveillance requirements being applied to operating limits will

provide greater assurance that the diesels will be operable and that

greater performance requirements are not imposed on ESF equipment.

This change, therefore, will not result in an increase in the

consequences of an accident previously described.

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

kind of accident from any accident previously evaluated?

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

different kind of accident from any accident previously evaluated.

They also will have no adverse impact on the design basis accidents

previously evaluated in the UFSAR. The revisions contained in the

proposed amendment are more restrictive to assure that diesel and

ESF equipment are available and fully operable to perform their

intended safety function following a design basis accident and a

loss of offsite power. The proposed changes do not involve physical

changes to plant equipment or the AC power system configuration. New

failure modes are not introduced as a result of the proposed

revisions. A revision of the diesel frequency will prevent motors

and pumps from being subjected to over-frequency conditions which

could reduce the life of the equipment. Increasing the load

rejection criteria for Unit 1 and including overspeed criteria for

both units revises surveillance test criteria for verifying load

rejection capability. This does not affect the probability of

malfunction of a diesel or its connected emergency AC power system.

Further, it does not create a new failure mode. Revising diesel fuel

oil storage requirements to include the term ``usable'' reduces the

potential for misinterpretation of this specification; it does not

create a new kind of accident from any accident previously

evaluated.

The revisions contained in this license amendment have the

effect of making the BVPS Technical Specifications more conservative

than previously. This license amendment request will not create the

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

previously evaluated.

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

of safety?

The margin of safety is not reduced as a result of the proposed

revisions. The margin of safety depends on the maintenance of

specific operating parameters within design limits. The margin of

safety derived from limiting condition for operation 3.8.1.1 and

3.8.1.2 for both Units is enhanced by adding ``usable'' in these

requirements. This revision reduces the possibility of

misinterpreting Technical Specification requirements. The addition

of diesel overspeed criteria (both units) and increasing load

rejection criteria for Unit 1 does not reduce the margin of safety.

Diesel reliability and performance during a loss of offsite power

and a design basis accident are enhanced by this more conservative

surveillance test requirement. Revision of diesel operating

frequency limits protects engineered safety features equipment from

overfrequency conditions; this would not be a significant reduction

in the margin of safety. Though the temporary Unit 1 EDG loading

limit of 2791.51 exceeds the Safety Guide 9 value of 2745, it still

is below the EDG 2000 hour rating limit of 2850 kW contained in

Surveillance Requirement 4.8.1.1.2.b.6. Further, the loading value

of 2791.51 kW does not exceed the design

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loading capability of the EDG. Based on engineering analyses, the

revisions contained in the proposed amendment will not significantly

reduce the margin of safety. Engineered safety features equipment

will continue to function, as assumed in the safety analysis, to

ensure that fuel, reactor coolant system and containment design

limits are not exceeded.

Therefore, this change will not involve a significant reduction

in a margin of safety due to the continued availability and

reliability of the A.C. electrical power sources.

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: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: S. Singh Bajwa.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: December 24, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specification (TS) requirements for the axial flux

difference [AFD] monitor, quadrant power tilt ratio [QPTR] monitor, rod

position deviation monitor, and rod insertion limit (RIL) monitor. The

changes would (1) relocate requirements for the AFD monitor and the

QPTR monitor to the Licensing Requirements Manual (LRM); (2) delete

requirements for the rod position deviation monitor and RIL monitor

from the TSs; (3) modify Unit 1 surveillance requirements (SR) 4.1.3.5

and 4.1.3.6 by incorporating the Unit 2 wording to provide

surveillances more consistent with the Limiting Condition for Operation

(LCO); (4) change Unit 1 SR 4.1.3.2.2, SR 4.1.3.5, SR 4.1.3.6 and Unit

2 SR 4.1.3.5 from 24 hour surveillance frequencies to 12 hour

frequencies; and (5) delete Unit 1 SR 4.1.3.2.3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment would modify applicable Technical

Specifications (TS) by deleting requirements associated with the rod

position deviation monitor and rod insertion limit (RIL) monitor and

relocating the requirements associated with the axial flux

difference (AFD) monitor and quadrant power tilt ratio (QPTR)

monitor from the following specifications and Bases:

Unit 1: 4.1.3.1.2, 3.1.3.2, 4.1.3.2.2, 4.1.3.2.3, 4.1.3.6, 4.2.1.1,

4.2.4;

Unit 2: 4.1.3.1.2, 4.1.3.2, 4.1.3.6, 4.2.1.1, 4.2.4.

The TS contains requirements where a reduced surveillance

interval is required in the event the monitors referenced in the

above specifications, surveillance requirements (SR) and associated

Bases are inoperable. Removing the requirements associated with

these monitors from the TS will not affect the ability of any system

to perform its design function.

Nuclear Electric Institute (NEI) Technical Specification Task

Force (TSTF) 110 Revision 2 provides the basis for these changes and

recommends relocating the requirements for these monitors to ``plant

administrative practices.'' The AFD monitor and the QPTR monitor

requirements will be relocated to the LRM and changes to these

requirements will be controlled in accordance with the 10 CFR 50.59

process which will require NRC approval if the change constitutes an

unreviewed safety question. However, based on the smaller change in

surveillance intervals, deletion and not relocation of the rod

position deviation monitor and the RIL monitor requirements can be

justified and is proposed.

Although these monitors are being removed from the TSs, they

will continue to be maintained as described in the [Updated Final

Safety Analysis Report] UFSAR (subject to revisions via the 10 CFR

50.59 process). Removing the rod deviation monitor requirements from

Unit 1 SR 4.1.3.2.3 makes the remaining portion of SR 4.1.3.2.3

redundant to SR 4.1.3.2.2.a; therefore, SR 4.1.3.2.3 has been

deleted. In addition, the 24-hour surveillance frequency in Unit 1

SR 4.1.3.2.2, 4.1.3.5 and 4.1.3.6 as well as in Unit 2 SR 4.1.3.5 is

being changed to 12 hours to assure the required parameters are

adequately monitored and to provide consistency between the units

and related requirements as well as the Improved Standard Technical

Specifications (ISTS).

Removing these monitors from the TS is consistent with the NRC

approved changes to the ISTS identified in TSTF-110, Revision 2.

Verification that plant conditions are within specified limits at

the frequency specified in the normal SR provides sufficient

information that allows the operator to detect a parameter that is

beginning to deviate from its expected limits. The specified

frequency takes into account other information (i.e., rod position

indication system, rod bottom alarm and excore neutron detectors)

that is continuously available to the operator in the control room,

so that during changes in plant conditions, deviation from the

limits can be readily detected.

The proposed changes do not affect the operation of the system

or the accident analyses and are consistent with the NRC approved

changes to the surveillances identified for the ISTS of NUREG-1431

identified in TSTF-110, Revision 2. These changes do not involve a

change to plant equipment and do not affect the performance of plant

equipment used to mitigate an accident. Although the deletion of

these monitor requirements from the TS results in elimination of the

reduced surveillance interval when the alarm is inoperable (for

those requirements not being relocated to the LRM) the change in

frequency is not significant considering the indications available

to the operator and the relatively slow changes in the parameters

being monitored during steady state operation. Therefore, based on

the above, these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Unit 1 SRs 4.1.3.5 and 4.1.3.6 have been additionally modified

by incorporating the Unit 2 wording which more closely provides a

surveillance appropriate for the LCO. The LCO requires the shutdown

rods/control banks to be within the insertion limits and the revised

SR requires a determination that each shutdown rod/control bank is

within the insertion limits on a 12-hour frequency. Therefore, the

revised SRs are consistent with the LCO requirements and more

clearly provide verification that the LCO is met. This change does

not affect the operation of the rod position indication system or

any other system and is consistent with the Unit 2 and ISTS wording.

This change will not affect the ability of any system to perform its

design function; therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

Changing the surveillance frequency from 24 to 12 hours is more

conservative and assures the affected parameters are adequately

monitored. In addition, the change removes monitors from the TSs and

provides consistency between the SRs, the units and the ISTS.

Changing the surveillance frequency, correcting the Unit 1 SRs and

removing reference to the identified monitors from the TS will not

cause a significant reduction in system reliability nor affect the

ability of any system to perform its design function. There are no

hardware changes associated with this license amendment nor are

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

system performs its safety function. No new accident scenarios,

transient precursors, failure mechanisms or limiting single failures

are introduced as a result of these changes. These changes do not

introduce any adverse effects or challenges to any safety-related

systems. No change is required to any system configurations, plant

equipment or analyses. Therefore, these changes will not create the

possibility of any new or different kind of accident from any

accident previously evaluated.

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

of safety?

[[Page 4156]]

The proposed changes do not affect the acceptance criteria for

any analyzed event nor impact any plant safety analyses since the

assumptions used will remain unchanged. The safety limits assumed in

the accident analyses and the design function of the equipment

required to mitigate the consequences of any postulated accidents

will not be changed since the proposed changes do not affect the

accident analyses assumptions or equipment required to mitigate

design basis accidents described in the UFSAR. Although the deletion

of these monitor requirements from the TSs results in elimination of

the reduced surveillance interval when the alarm is inoperable (for

those requirements not being relocated to the LRM) the effect is not

significant considering the indications available to the operator

and the relatively slow changes in the parameters being monitored

during steady state operation. The TSs continue to assure the

applicable operating parameters are maintained within the required

limits. Based on engineering judgement, incorporating these changes

will not involve a significant reduction in the margin of safety.

The margin of safety depends upon maintenance of specific

operating parameters within design limits. The TSs continue to

require that these limits be maintained and provide appropriate

remedial actions if a limit is exceeded. The maintenance of these

limits continues to be assured through performance of the normal

surveillance at the proposed frequency and the requirements for

increased monitoring that are relocated to the LRM. Additional

assurance that the required parameters are adequately monitored is

provided through other information readily available (i.e., rod

position indication system, rod bottom alarm and excore neutron

detectors) that allows the operator to detect a parameter that is

beginning to deviate from its expected limits and through the

proposed changes which reduce the normal surveillance interval from

24 hours to 12 hours to assure the affected parameters are

adequately monitored. Although these monitors are being removed from

the TSs, they will continue to be maintained as described in the

UFSAR (subject to revisions via the 10 CFR 50.59 process).

Therefore, the plant will be maintained within the analyzed limits

and the proposed changes will not involve a significant reduction in

a margin of safety.

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

this review, it appears that the three standards of 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: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: S. Singh Bajwa.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas.

Date of amendment request: November 24, 1998.

Description of amendment request: The proposed changes implement

the consolidated Entergy Operations Quality Assurance Plan Manual

approved by the NRC on November 6, 1998. The proposed changes also

clarify the responsibilities of the shift technical advisor position on

shift, simplify the contents of the monthly operating report

description in accordance with Generic Letter (GL) 97-02, complete the

relocation of fire protection requirements from the TS to the fire

protection program in accordance with GL 88-12, and replace position

titles with descriptions of functional responsibility in accordance

with GL 88-06.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed changes only affect the administrative controls

contained in Section 6.0 of the Arkansas Nuclear One--Unit 1 (ANO-1)

and Unit 2 (ANO-2) Technical Specifications (TSs). The proposed

changes either add additional administrative controls, reduce

regulatory duplication of requirements consistent with NUREG-1430

``Standard Technical Specifications--Babcock and Wilcox Plants''

dated April 1995, and NUREG-1432 ``Standard Technical

Specifications--Combustion Engineering Plants'' dated April 1995, or

revise or relocate administrative controls in accordance with NRC

guidance. The proposed changes do not affect the operation of any

structure, system, or component or the assumptions of any accident

analysis. The details relocated from the ANO-1 and ANO-2 TSs, and

changes to these details, are controlled under the ANO 10 CFR 50.59

or 10 CFR 50.54 processes as appropriate.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

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

Kind of Accident from any Previously Evaluated.

The proposed changes to the ANO-1 and ANO-2 Section 6.0

administrative controls do not involve a change in the plant design

or affect the configuration or operation of any structure, system,

or component.

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

or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin

of Safety.

The proposed changes to the ANO-1 and ANO-2 TSs affect only

administrative requirements and do not involve changes to safety

limits, limiting conditions for operation, or surveillance

requirements on equipment required to operate the station.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

this review, it appears that the three standards of 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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

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

NRC Project Director: John N. Hannon.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: November 30, 1998.

Description of amendment request: The proposed amendment would

change the CR-3 Improved Technical Specifications (ITS) Section 3.9.3,

Containment Penetrations. The proposed changes recognize the use of an

outage equipment hatch (OEH) during refueling operations. The proposed

changes would also allow both doors in the personnel air locks, and the

single door in the OEH, to be open during core alterations or movement

of irradiated fuel assemblies within containment provided certain

specified conditions are met.

The licensee stated that the ability to open these doors under

administrative controls would assist in the maintenance of cleanliness

and housekeeping, and would provide a safer work environment inside

containment. In addition, the licensee stated that evacuation of

personnel could be quickly achieved in the unlikely event of a fuel

handling accident or other radiological event inside containment,

reducing the potential for exposures.

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.

[[Page 4157]]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The proposed change would allow both doors in the personnel air

locks and the door in the outage equipment hatch (OEH) to remain

open during core alterations or the movement of irradiated fuel

inside containment. These doors are normally closed during this

period in order to prevent the escape of radioactive materials in

case of a fuel handling accident.

Operations involving the personnel air locks during refueling

operations cannot be an initiator of a fuel handling accident or

other radiological event inside containment. Similarly, operations

involving the OEH during refueling operations cannot be an initiator

of a fuel handling accident or other radiological event inside

containment. The personnel air locks and the OEH are remotely

located to the fuel handling equipment and cannot affect the

function of this equipment. The personnel air locks and the OEH are

not in the immediate vicinity of the reactor vessel and the

contained irradiated fuel, or any of the paths used for movement of

irradiated fuel. Additionally, allowing both doors in the personnel

air locks and the door in the OEH to be open during core alterations

or the movement of irradiated fuel inside containment cannot create

the possibility of a fuel handling accident or other radiological

event inside containment. Therefore, the probability of occurrence

of any accident previously evaluated is unaffected.

The approved fuel handling accident analysis does not take

credit for containment closure. This analysis results in a maximum

calculated offsite dose well within the limits of 10 CFR 100, and

the existing analysis as presented in the CR-3 Final Safety Analysis

Report does not require revision as a result of this proposed

change. By providing a designated individual readily available to

close at least one door in the personnel air locks and the door in

the OEH, containment closure is assured following any required

evacuation of containment terminating any release of radioactive

materials outside of the containment. Therefore, the consequences of

accidents will not be greater than that previously evaluated.

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

from previously evaluated accidents?

The operations involving the personnel air locks and the OEH

cannot be an initiator of any type of accident during refueling

operations. The personnel air locks and the OEH are passive

structural features designed to retain structural integrity under

the expected environmental conditions when installed. Operation of

the personnel air lock doors and the door in the OEH does not affect

any safety-related component or structure. Additionally, allowing

both doors in the personnel air locks and the door in the OEH to be

open during core alterations or the movement of irradiated fuel

inside containment cannot initiate any type of accident. Therefore,

the possibility of a new or different kind of accident occurring as

a result of this change is not created.

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

The margin of safety as defined by 10 CFR 100 has not been

reduced. The existing approved fuel handling accident analysis does

not credit containment closure, and remains bounding with both doors

in the personnel air locks and the door in the OEH open. Closing at

least one door in the personnel air locks and the door in the OEH

after evacuation of containment further reduces the offsite doses in

case of a fuel handling accident, and provides additional margin to

the calculated offsite doses. Therefore, the existing margin of

safety will not be reduced.

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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Project Director: Cecil O. Thomas

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook

Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: November 4, 1998.

Description of amendment request: The proposed change would revise

Technical Specifications Surveillance Requirement 4.5.2b.1 to delete

the prescribed method of venting the Emergency Core Cooling System

(ECCS) which would allow alternate methods to verify that the ECCS

piping is full of water. In addition, the associated Bases would be

expanded to reflect the intent of the surveillance requirement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change does not adversely affect accident

initiators or precursors nor alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. The proposed change does not alter or prevent the

ability of structures, systems and components (SSCs) to perform

their intended function to mitigate the consequences of an

initiating event within the acceptance limits assumed in the Updated

Final Safety Analysis Report (UFSAR).

Removal of the prescriptive requirements will not subject the

ECCS system to conditions adverse to nuclear safety. The proposed

change does not affect the source term, containment isolation or

radiological release assumptions used in evaluating the radiological

consequences of an accident previously evaluated in the Seabrook

Station UFSAR. The use of proven alternative techniques to verify

that the ECCS piping is full of water will continue to ensure that

the ECCS system is capable of performing its intended designed

safety function. Therefore, the proposed 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 accident previously evaluated.

The proposed change does not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated and maintained in a state of readiness. Existing

system and component redundancy is not being changed by the proposed

change. The proposed change has no adverse affect on component or

system interactions. The use of proven alternative techniques to

verify that the ECCS piping is full of water will continue to ensure

that the ECCS system is capable of performing its intended designed

safety function. Therefore, since there are no changes to the design

assumptions, conditions, configuration of the facility, or the

manner in which the plant is operated and maintained in a state of

readiness, the proposed change does not create the possibility of a

new or different kind of accident from any previously analyzed.

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

The proposed change does not adversely affect equipment design

or operation and there are no changes being made to the Technical

Specification required safety limits or safety system settings that

would adversely affect plant safety. The proposed change does not

change the intent of the surveillance requirement of ensuring that

the system will perform properly, injecting its full capacity into

the RCS upon demand without subjecting the system to hydraulic

transients, pump cavitation, and pumping of non-condensable gas

(e.g., air, nitrogen, or hydrogen) into the reactor vessel following

a safety injection (SI) signal or during shutdown cooling.

Thus, it is concluded that the ECCS will continue to be

available upon demand to mitigate the consequences of an accident

and, therefore, there is no 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.

[[Page 4158]]

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-245,

50-336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2,

and 3, New London County, Connecticut

Date of amendment request: December 22, 1998.

Description of amendment request: The proposed amendment would

replace specific titles in Section 6.0 of the Technical Specifications

of all three Millstone units with generic titles.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECO has reviewed the attached

proposed changes and ha[s] concluded that they do not involve a

Significant Hazard Consideration (SHC). The basis for this

conclusion is that the three criterion of 10 CFR 50.92 are not

compromised. The proposed change is not a[n] SHC because the

proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No design basis accidents are affected by these proposed

changes. The proposed changes are administrative in nature and are

being proposed to eliminate the need for a Technical Specification

change each time there is a change in the organization.

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

from any accident previously evaluated.

There are no changes in the way the plant is operated due to

these administrative changes. The potential for an unanalyzed

accident is not created. There is no impact on plant response, and

no new failure modes are introduced. The proposed administrative and

editorial changes have no impact on safety limits or design basis

accidents, and have no potential to create a new or unanalyzed

event.

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

These changes do not directly affect any protective boundaries

nor do they impact the safety limits for the protective boundaries.

These proposed changes are administrative and editorial in nature.

Therefore there is no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-336,

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of amendment request: November 13, 1998.

Description of amendment request: NNECO has determined that the

increase in radiological consequences, due to changes in the

assumptions used in the updated dose consequence analysis of the Steam

Generator Tube Rupture (SGTR) event in the Millstone Unit No. 2 Final

Safety Analysis Report (FSAR), involves an unreviewed safety question

(USQ). The changes include a change in High Pressure Safety Injection

(HPSI) pump runout flowrate, a change in Auxiliary Feedwater Pump (AFW)

flowrate, a change in the iodine partition factor for the air ejector,

inclusion of the potential of flashing of the primary-to-secondary

leakage, and a change in the atmospheric release point assumed

following actuation of the Enclosure Building Filtration Actuation

Signal (EBFAS). Therefore, per 10CFR50.59(c), NNECO requested that the

NRC review and approve the changes to the FSAR through an amendment to

Operating License DPR-65, pursuant to 10CFR50.90.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10CFR50.92, NNECO has reviewed the proposed

changes and has concluded that they do not involve a Significant

Hazards Consideration (SHC). The basis for this conclusion is that

the three criteria of 10CFR50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The FSAR changes reflect changes in the updated SGTR analysis.

The analysis was updated because of changes in the assumptions used

in the dose consequence analysis of the SGTR event in Millstone Unit

No. 2 FSAR. These changes include a change in the iodine partition

factor for the air ejector, inclusion of the potential of flashing

of the primary-to-secondary leakage, and a change in the atmospheric

release point assumed following actuation of the EBFAS. In addition,

the operator actions associated with Reactor Coolant System (RCS)

cooldown that are specified in the Emergency Operating Procedures

have been incorporated, mass releases assuming an RCS cooldown to

Shutdown Cooling Entry conditions have been used in the dose

consequence analysis, thyroid doses were calculated using ICRP-30

dose conversion factors, Iodine releases account for potential

flashing of the primary-to-secondary leakage, and the Reactor

Coolant pumps are assumed to be tripped following actuation of a

safety injection actuation signal. The revised HPSI flowrate is

higher than that used in the previous analysis. Higher HPSI

flowrates would increase the primary-to-secondary break flow and,

thereby, increase the dose consequences. A more conservative iodine

partition factor for the air ejector has been used along with more

limiting atmospheric dispersion coefficients as a result of manual

realignment of the air ejector discharge path to the atmosphere.

These changes in radiological assumptions are the major reason for

the increase in calculated dose. The revised AFW flowrate is lower

than that used in the previous analysis. Lower AFW flowrate would

tend to increase the steaming required and, thereby, increase the

dose consequences. The probability that an accident could occur due

to these changes is not increased since changing the analysis and

its description can not cause a steam generator tube rupture.

Therefore, these changes will not significantly increase the

probability of an accident previously evaluated.

The dose consequences for the updated SGTR analysis are higher

than the dose consequences for the previous analysis. However, the

dose consequences are within the acceptance criteria of SRP

[Standard Review Plan] 15.6.3 and GDC [General Design Criterion] 19.

Therefore, these changes will not significantly increase the

consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The FSAR changes reflect changes in the updated SGTR analysis.

The updated analysis does not introduce any new or unanalyzed

failure modes of equipment or systems, and does not change the

configuration of the plant. While the updated analysis incorporates

operator actions that are in accordance with the Emergency Operating

Procedures, it does not alter the way any structure, system, or

component functions, and does not alter the manner in which the

plant is operated. Therefore, there are no new or different types of

failures of systems or equipment important to safety

[[Page 4159]]

which could cause a new or different type of accident from any

accident previously evaluated.

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

The FSAR changes reflect changes in the updated SGTR analysis.

The updated analysis shows that the dose consequence acceptance

criteria are met. The updated analysis incorporates operator actions

that are in accordance with the Emergency Operating Procedures, and

credits equipment consistent with its capabilities. Therefore, the

updated analysis does not reduce the margin of safety. The FSAR

changes do not alter the acceptance limits of the safety parameters

of the accident analyses stated in the FSAR. Therefore, these

changes do not significantly reduce the margin of safety.

The NRC has provided guidance concerning the application of

standards in 10CFR50.92 by providing certain examples (March 6,

1986, 51 FR 7751) of amendments that are considered not likely to

involve an SHC. The changes proposed herein are covered by example

(vi) in that the consequences for the updated SGTR analysis are

higher than dose consequences for the previous analysis. However,

the dose consequences are within the acceptance criteria of SRP

15.6.3 and GDC 19.

As described above, this License Amendment Request does not

involve a significant increase in the probability of an accident

previously evaluated, does not involve a significant increase in the

consequences of an accident previously evaluated, does not create

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

accident previously evaluated, and does not result in a significant

reduction in a margin of safety. Therefore, NNECO has concluded that

the proposed changes do not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: November 20, 1998.

Description of amendment request: This amendment request updates

the Emergency Diesel Generator (EDG) day tank volume Surveillance

Requirement (SR) 3.8.1.4.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated. The proposed

amendment changes EDG day tank volume requirements to reflect the

[Susquehanna Steam Electric Station] SSES design.

The safety function of the EDG day tanks is to supply the EDG's

with enough fuel to ensure the availability of necessary power to

[engineered safety feature] ESF systems so that fuel, reactor

coolant and containment system design limits are not exceeded. The

proposed change increases the minimum diesel fuel oil day tank

volume for Unit 1 and Unit 2 SR 3.8.1.4 from 325 gallons to 420

gallons for EDG A-D and 425 gallons for EDG E.

This volume corresponds to the tank volume at which automatic

refill occurs. This volume provides for 55 minutes of EDG A-D and 62

minutes for EDG E operation at continuous rated load conditions.

Currently, the bases for SR 3.8.1.4 identifies that

``administrative controls ensure a useable volume of the fuel oil in

the day tank adequate for approximately 60 minutes of DG operation

plus 10% at the continuous rated load.'' These administrative

controls ensure compliance with the Regulatory Guide 1.137

requirements. Regulatory Guide 1.137 revision 1 endorses American

National Standards Institute (ANSI) N195-1976. The ANSI N195-1976

requires each diesel to be equipped with a day tank whose capacity

is sufficient to maintain at least 60 minutes of operation. This

capacity is to be based on the fuel consumption at a load of 100% of

the continuous rating of the diesel plus a minimum margin of 10%.

These administrative controls on day tank level ensure that the

required initial fuel oil supply is available to meet the intent of

the Standard as it applies to the Technical Specification

surveillance. This Technical Specification change eliminates these

unnecessary controls needed to conform to the ANSI standard.

An assessment of the proposed change based on the guidance

provided in Regulatory Guide 1.174, July 1998, ``An Approach for

Using Probabilistic Risk Assessment in Risk-Informed Decisions on

Plant Specific Changes to the Licensing Basis'' concludes that the

increase in risk is insignificant. It is therefore concluded that

the proposed changes to SSES Unit 1 and Unit 2 Technical

Specification SR 3.8.1.4 day tank volume requirements ensures the

volume is adequate to support the EDG's post accident design basis

safety function to ensure the availability of necessary power to ESF

systems so that fuel, reactor coolant system, and containment design

limits are not exceeded.

Based upon the above, PP&L concludes that the proposed action

does not involve an increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The change to the day tank required minimum volume does not change

any plant systems, structures, or components, nor does the change

affect any existing or create any new or different kind of accident.

An assessment of the proposed change based on the guidance

provided in Regulatory Guide 1.174, July 1998, ``An Approach for

Using Probabilistic Risk Assessment in Risk-Informed Decisions on

Plant Specific Changes to the Licensing Basis'' concludes that the

increase in risk is insignificant. Based on this, it is concluded

that the proposed changes to SSES Unit 1 and Unit 2 Technical

Specification SR 3.8.1.4 day tank volume requirements ensures the

volume is adequate to support the EDG's post accident design basis

safety function to ensure the availability of necessary power to ESF

systems so that fuel, reactor coolant system, and containment design

limits are not exceeded.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

An assessment of the proposed change based on the guidance

provided in Regulatory Guide 1.174, July 1998, ``An Approach for

Using Probabilistic Risk Assessment in Risk-Informed Decisions on

Plant Specific Changes to the Licensing Basis'' concludes that the

increase in risk is insignificant.

It is concluded that the proposed changes to SSES Unit 1 and

Unit 2 Technical Specification SR 3.8.1.4 day tank volume

requirements ensures the volume is adequate to support the EDG's

post accident design basis safety function to ensure the

availability of necessary power to ESF systems so that fuel, reactor

coolant system, and containment design limits are not exceeded.

Based on this, the proposed changes do not involve a 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.

[[Page 4160]]

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: S. Singh Bajwa.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of amendment request: November 23, 1998.

Description of amendment request: These amendments would modify the

Susquehanna Steam Electric Station, Units 1 and 2, Technical

Specifications (TS) limiting condition for operation (LCO) 3.8.3 and

surveillance requirement (SR) 3.8.3.1 to increase the minimum fuel oil

storage tank (FOST) volume ranges. The Bases would be modified to

reflect that the proposed volumes equal the 7-day fuel oil consumption

at the continuous emergency diesel generator (EDG) ratings, which are

greater than design basis analysis (DBA) loads, plus the unusable

volume in the storage tanks.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated. The proposed

amendment increases FOST volume requirements so to increase the

margin of safety thus providing further assurance that the EDG FOST

volume is adequate to support the EDG's post accident design basis

safety function.

The safety function of the EDG FOST is to supply the emergency

diesel generators with enough fuel to ensure the availability of

necessary power to ESF systems so that fuel, reactor coolant and

containment system design limits are not exceeded. The current

Technical specification FOST specified volume is based on the EDG

post DBA load profile. The proposed FOST volume is based on EDG

continuos [sic] [continuous] rated load rating which is greater than

the post DBA load profile providing margin and further assurance

that the EDG FOST will support the EDG safety function. The proposed

required FOST volumes are calculated in accordance with ANSI N195-

1976.

Based upon the above, PP&L concludes that the proposed action

does not involve an increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The FOST required minimum values do not change any plant systems,

structures, or components, nor do they change any existing or create

any new or different kind of accident. The proposed amendment

changes FOST volume requirements so to increase the margin of safety

thus providing further assurance that the EDG FOST volume is

adequate to support the EDG's post accident design basis safety

function. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change increases the margin of safety since the

proposed FOST values are based on the EDG continuos [sic]

[continuous] rated load ratings which bound the post DBA load

profile.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

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 amendment request: November 6, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications for the Nuclear Instrumentation

System [NIS] Power Range daily surveillance requirement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed surveillance change involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The proposed surveillance change does not significantly increase

the probability or consequences of an accident previously evaluated

in the FSAR [Final Safety Analysis Report]. This modification does

not directly initiate an accident. The consequences of accidents

previously evaluated in the FSAR are not adversely affected by this

proposed change because the change to the NIS Power Range channel

adjustment requirement ensures the conservative response of the

channel even at part power levels.

2. Does the proposed surveillance change create the possibility

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

evaluated?

The proposed surveillance change does not create the possibility

of a new or different kind of accident than any accident already

evaluated in the FSAR. No new accident scenarios, failure

mechanisms, or limiting single failures are introduced as a result

of the proposed change. The proposed Technical Specifications change

does not challenge the performance or integrity of any safety-

related systems. Therefore, the possibility of a new or different

kind of accident is not created.

3. Does the proposed surveillance change involve a significant

reduction in a margin of safety?

The proposed surveillance change does not involve a significant

reduction in a margin of safety. The proposed change does require a

revision to the criterion for implementation of Power Range channel

adjustment based on secondary power calorimetric calculation;

however, the change does not eliminate any RTS [Reactor Trip

Setpoint] surveillances or alter the frequency of surveillances

required by the Technical Specifications. The revision to the

criterion for implementation of the daily surveillance will have a

conservative effect on the performance of the NIS Power Range

channel, particularly at part power after normalization at 100% RTP

[Rated Thermal Power] conditions. The nominal trip setpoints

specified by the Technical Specifications and the safety analysis

limits assumed in the transient and accident analysis are unchanged.

The margin of safety associated with the acceptance criteria for any

accident is unchanged. Therefore, the proposed change will not

significantly reduce the margin of safety as defined in the

Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

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

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

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

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

Alabama.

[[Page 4161]]

NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch

Nuclear Plant, Units 1 and 2, Appling County, Georgia.

Date of amendment request: December 4, 1998.

Description of amendment request: The proposed amendments would

make two changes to the Technical Specifications (TSs). Change 1 would

delete the footnote in Hatch Unit 1 TS Section 2.1.1.2 that ties the

Safety Limit Minimum Critical Power Ratio to Cycle 18. Change 2 would

delete TS Section 5.6.5.b.2 for Units 1 and 2, and incorporate TS

Section 5.6.5.b.2 into TS Section 5.6.5.b.1 for both units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

Basis for Proposed Change 1

The change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The footnote in Section 2.1.1.2 of the Hatch-1 Technical

Specifications restricts the applicability of the Safety Limit for

MCPR [minimum critical power ratio] (SLMCPR) [safety limit minimum

critical power ratio] to Cycle 18 only. By applying the same NRC-

approved methods used to calculate the Cycle 18 SLMCPR it has been

determined that the current value is bounding for Cycle 19 as well.

However, because of the footnote, it [cannot] be applied to Cycle 19

without a Technical Specifications amendment. In order to eliminate

future Technical Specifications revisions that do not change the

SLMCPRs values, SNC [Southern Nuclear Operating Company, Inc.]

proposes to delete the footnote which ties those values to a

specific operating cycle. Removing the footnote does not change the

method of calculating SLMCPR for other cycles, nor does it eliminate

the requirement to revise the Technical Specifications if a

different value is used for future cycles. Deletion of the cycle-

specific footnote does not change the operation of any plant

structure, system or component; therefore, it has no affect on the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any previously evaluated.

Deleting the cycle-specific footnote in Section 2.1.1.2 of the

Technical Specifications does not result in any new methods of

operating the facility and does not involve any facility

modifications. No new initiating events or transients result from

this change.

Therefore, this proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The purpose of the SLMCPR in the Technical Specifications is to

ensure at least 99.9% of the fuel pins in the core are expected to

avoid transition boiling during the worst anticipated operational

occurrence (AOO) throughout an operating cycle. The footnote in

Section 2.1.1.2 of the Hatch-1 Technical Specifications is intended

to ensure the correct SLMCPR is used each cycle. Prior to the Spring

of 1996, the Safety Limits had been calculated for each fuel type,

independently of operating cycle. As long as the limiting fuel type

in the core did not change from cycle to cycle, the Safety Limit did

not change. It was discovered in 1996, however, that generic SLMCPRs

based on fuel type alone may not be bounding for all cycles for all

reactors. In response to this discovery GE committed to evaluating

SLMCPRs based on cycle-unique information as a more accurate method

of ensuring 99.9% of the fuel pins in the core are expected to avoid

transition boiling during AOOs. The new methodology, which is now

applied each cycle, is based on NRC-approved methods and

incorporates implementing procedures that model cycle-specific

parameters. This methodology was used to calculate the Cycle 18

value that is currently in the Technical Specifications. The same

procedure was also employed to determine that the Hatch-1 Cycle 19

SLMCPR and it was determined the Cycle 19 value is bounded by the

Cycle 18 value. Thus, except for the footnote in Section 2.1.1.2,

there is no need to revise the Hatch-1 Technical Specifications in

order to ensure the correct SLMCPR is implemented for Cycle 19. As a

way of avoiding similar changes in the future, SNC proposes that the

footnote be deleted. Since NRC-approved methodology will still be

used to determine the cycle-specific SLMCPRs to ensure that [ ]

99.9% of the fuel rods are expected to avoid transition boiling

during AOOs, there will be no reduction of margin of safety as a

result of this change.

Basis for Proposed Change 2

The change does not involve a significant hazards consideration

for the following reasons:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Section 5.6.5.b.2) no longer describes NRC-approved methods for

analyzing fuel in the Unit 1 and Unit 2 reactors because the ANF

[advanced nuclear fuel] LUAs [lead use assemblies] have been

permanently discharged. Deleting Section 5.6.5.b.2) from the

Administrative Controls portion of the Technical Specifications does

not change the operation of any structure, system, or component in

the facility. Therefore, this amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any previously evaluated.

Deleting Section 5.6.5.b.2), which describes the use of ANF

methods for analyzing LUAs, from the Technical Specifications does

not result in any new methods of operating the facility and does not

involve any facility modifications. No new initiating events or

transients result from this change. Therefore, this proposed change

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

accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

ANF LUAs are no longer used as fuel in the Plant Hatch reactors,

therefore, ANF NRC-approved methods described in Technical

Specifications Section 5.6.5.b.2) are not used to determine power

distribution limits which appear in the COLR [Core Operating Limit

Report]. GE's [General Electric's] reload licensing methodology

described in Section 5.6.5.b.1) will be incorporated into Section

5.6.5.b. and will continue to be used to analyze the GE fuel in both

units. Therefore, this change does not involve a significant

reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

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

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

NRC Project Director: Herbert N. Berkow.

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

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

Date of amendment request: November 4, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) Sections 4.6.A.1.b and Basis

3.16 for Units 1 and 2 to revise the start/load time testing and

ratings for emergency diesel generators (EDGs). The changes will bring

the TS into conformance with the Updated Final Safety Analysis Report.

[[Page 4162]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Operation of the Surry Units 1 and 2 in accordance

with the proposed Technical Specification change does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

The currently specified ``less than 30 seconds'' time to be

replaced has no specific safety significance or design basis

regarding EDG starting. The proposed time change to ``less than or

equal to 10 seconds'' is more conservative and in agreement with

current accident analysis and surveillance testing. These changes do

not, in any way, affect the as-built conditions of the plant and do

not affect the initiators of analyzed events or the assumed

mitigation of accident or transient events. Analyzed events are

initiated by the failure of plant structures, systems, or

components. The proposed changes do not impact the condition or

performance of these structures, systems or components. Consequences

of analyzed events are the result of the plant being operated within

assumed parameters at the onset of any event, and the successful

functioning of at least one train or division of the equipment

credited with mitigating the event. There is no impact on the

capability of the credited equipment to perform, nor is there any

change in the likelihood that credited equipment will fail to

perform. As a result, there is no significant increase in the

probability or consequences of any accident previously evaluated and

Criterion 1 is, thereby, satisfied.

Criterion 2--The proposed Technical Specifications change does

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

from any accident previously evaluated.

The proposed change does not involve a physical alteration of

the plant, or a change in the methods used to operate the plant or

to respond to plant transients. No new or different equipment is

being installed and no installed equipment is being removed or

operated in a different manner. There is no alteration to the

parameters within which the plant is normally operated or in the

setpoints, which initiate protective or mitigative actions.

Consequently, no new failure modes are introduced and the proposed

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

accident from any previously evaluated and Criterion 2 is, thereby

satisfied.

Criterion 3--The proposed Technical Specifications change does

not involve a significant reduction in a margin of safety.

Margin of safety is established through the design of the plant

structures, systems and components, the parameters within which the

plant is operated, and the establishment of the setpoints for the

actuation of equipment relied upon to respond to an event. The

replacement of the ``less than 30 seconds'' requirement for loading

the EDGs with the more stringent ``less than or equal to 10

seconds'' requirement makes no change to the condition or

performance of equipment or system used in accident mitigation or

assumed for any accident analysis that could reduce a margin of

safety as described in the basis for any TS. Therefore, the proposed

changes do not involve a significant reduction in any margin of

safety described in the bases for the Technical Specifications and

Criterion 3 is, thereby, satisfied.

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: Donald P. Irwin, Esq., Hunton and Williams,

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

23219.

NRC Project Director: Herbert N. Berkow.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: September 28, 1998 (TSCR 208).

Description of amendment request: The proposed amendments will

clarify the notation definition of ``R'' in the Technical

Specifications (TS) and add a new frequency of ``A.'' The revision of

``R'' would specify the refueling frequency as 18 months and ``A''

would be defined as an annual or 12-month frequency.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant [PBNP] in

accordance with the proposed amendments will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

These changes do not involve a significant increase in the

probability of an accident previously evaluated because no such

accidents are affected by the proposed revisions to clarify that the

provisions of TS 15.4.0.2 apply to notation ``R'' in TS Table

15.4.1-1. The proposed TS changes do not introduce any new accident

initiators since no accidents previously evaluated have as their

initiators anything related to the change in the frequency of

surveillance testing.

The increased time potential between surveillance frequencies

does not significantly increase the probability [of] failure of the

instrumentation contained in TS Table 15.4.1-1. As noted above,

instrument drift studies concluded that the magnitude of the

instrument drift (for instrumentation affected by drift) that could

occur over a 22.5-month interval was bounded by the uncertainty

allowances used in determining safety system setpoints, and the

review of historical calibration data concluded that the as-found

and as-left data has not exceeded acceptable limits for the

calibration intervals reviewed, except on rare occasions.

In addition, initiating conditions and assumptions are unchanged

and remain as previously analyzed for accidents in the PBNP Final

Safety Analysis Report. The proposed TS changes do not involve any

physical changes to systems or components, nor do they alter the

typical manner in which the systems or components are operated.

Therefore, these changes do not increase the probability of

previously evaluated accidents.

These changes do not involve a significant increase in the

consequences of an accident previously evaluated because the source

term, containment isolation or radiological releases are not being

changed by these proposed revisions. Existing system and component

redundancy and operation is not being changed by these proposed

changes. The assumptions used in evaluating the radiological

consequences in the PBNP Final Safety Analysis Report are not

invalidated; therefore, these changes do not affect the consequences

of previously evaluated accidents.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

These changes do not introduce nor increase the number of

failure mechanisms of a new or different type than those previously

evaluated since there are no physical changes being made to the

facility. The surveillance test requirements and the way they are

performed will remain unchanged. The design and design basis of the

facility remain unchanged. The plant safety analyses remain

unchanged. Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated is not introduced.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in

a margin of safety.

The proposed changes do not involve a significant reduction in

the margin of safety because existing component redundancy is not

being changed by these proposed changes. There are no new or

significant changes to the initial conditions contributing to

accident severity or consequences, and safety margins established

through the design and facility license including the Technical

Specifications remain unchanged. Therefore, there are no significant

reductions in a margin of safety introduced by [these] proposed

amendment[s].

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

this review, it appears that the three

[[Page 4163]]

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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: October 5, 1998 (TSCR 200).

Description of amendment request: The proposed change modifies

Technical Specifications Section 15.4.1, ``Operational Safety Review,''

by removing the requirement to check environmental monitors on a

monthly basis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant [PBNP] in

accordance with the proposed amendments does not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change eliminates a surveillance requirement for

environmental monitors. The environmental monitors referred to by

this surveillance were eliminated from the Radiological

Environmental Monitoring Program and from the Technical

Specifications by previous amendments. Therefore, this change is

administrative in nature in that it corrects a previous

administrative oversight. The requirement is not related to any

accident initiator or accident mitigation structures, systems or

components for any previously evaluated accident. Therefore, no

increase in the probability or consequences of a previously

evaluated accident can result.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment[s] does not create a new or different kind of

accident from any accident previously evaluated.

The amendments remove a surveillance requirement from the

Technical Specifications related to environmental monitors. The

environmental monitors were removed from the environmental

monitoring program by previously approved amendments. The

surveillance requirement is not related to an existing design

feature of PBNP. Therefore, elimination of the surveillance

requirement cannot create a new or different kind of accident from

any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment[s] does not result in a significant reduction

in a margin of safety.

Margins of safety are defined by the safety limits and design

limits for PBNP. The surveillance is not related to, nor does it

affect, these limits. Monitoring of the environment continues under

an approved Radiological Environmental Monitoring Program which

ensures that any changes in radiation levels in the environs is

detected, thus ensuring the impact of PBNP operation on the

environment is minimized. Therefore, the proposed change cannot

result in a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

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.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1 (WBN), Rhea County, Tennessee

Date of application for amendment: December 22 and revision dated

December 23, 1998.

Brief description of amendment: In order to prevent a potential

shutdown due to sporadic grounds encountered on an annunciator circuit

used to confirm operability of an ice condenser inlet door position

monitoring system, the proposed amendment would provide a temporary,

optional method of satisfying the requirements for the channel check

until the next operating Mode, planned in late February 1999, for the

next refueling outage. Date of publication of individual notice in the

Federal Register: December 31, 1998 (63 FR 72339).

Expiration date of individual notice: February 1, 1999.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

Street, NW., Washington, DC, and at the

[[Page 4164]]

local public document rooms for the particular facilities involved.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: November 11, 1998.

Brief description of amendments: The amendments revise Technical

Specification Surveillance Requirements (SRs) 3.6.11.6 AND 3.6.11.7,

regarding the Containment Pressure Control System (CPCS), of the units'

joint Technical Specifications. The revision brings the SRs into

conformity with the current design of the CPCS.

Date of issuance: January 14, 1999.

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

concurrently with implementation of Amendment Nos. 173 (Unit 1) and 165

(Unit 2).

Amendment Nos.: 174--Unit 1; Unit 2--166.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66591). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

2, Pope County, Arkansas

Date of amendment request: September 17, 1998.

Brief description of amendment: The amendment incorporates the use

of a range rather then a specific setpoint for the automatic removal of

the operating bypasses for the core power calculator (CPC) generated

trips and the high logarithmic power level trip to accommodate the

design of the plant protection system (PPS) which uses a single

bistable to control both of these functions.

Date of issuance: December 31, 1998.

Effective date: December 31, 1998.

Amendment No.: 196.

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56247).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: June 29, 1998.

Brief description of amendment: The amendment modifies the TS

surveillance requirements for SR 4.8.2.3.b.2, SR 4.8.2.3.c.4 and the

Bases for TS 3.8.2.3 Action b. The licensee is planning to modify the

120 volt vital alternating current (ac) electrical distribution system

by installing new inverters during the 2R13 refueling outage. Normally,

the present inverters for ANO-2 are ac powered and automatically shift

to direct current (dc) power on a loss of the ac source. The new

inverters will be powered from the 125 dc system at all times.

Date of issuance: January 13, 1999.

Effective date: January 13, 1999, with implementation following

completion of the required modifications but prior to restart from the

2R13 outage.

Amendment No.: 198.

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56244).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

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

Date of application for amendment: October 29, 1998.

Brief description of amendment: The amendment revised the

terminology used in the St. Lucie Plant Technical Specifications (TS)

relative to the implementation and automatic removal of certain

protection system trip bypasses to ensure that the meaning of explicit

terms used in the TS are consistent with the intent of the stated

requirements.

Date of Issuance: January 5, 1999.

Effective Date: As of date of issuance and shall be implemented

within 30 days of receipt.

Amendment No.: 159.

Facility Operating License No. DPR-67: Amendment revised the TS.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66594) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 5, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: August 4, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) relating to the condensate storage tank (CST)

relating to the required minimum water volume and also adds a new TS

which establishes requirements for the atmospheric steam dump valves

(ASDVs) to assure their operability. The applicable TS Bases for the

CST is updated to reflect the proposed changes and a new TS Bases

section is added to discuss the new TS for the ASDVs.

Date of issuance: December 31, 1998.

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

60 days from the date of issuance.

Amendment No.: 223.

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

Technical Specifications.

Date of initial notice in Federal Register: August 26, 1998 (63 FR

45526).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

Unit No. 1, Washington County, Nebraska

Date of amendment request: March 26, 1997, as supplemented by

letters dated March 18, 1998, and November 17, 1998.

Brief description of amendment: The amendment revises Technical

[[Page 4165]]

Specifications (TS) 2.1.6 and its associated Basis to restrict the

number of inoperable main steam safety valves when the reactor is

critical.

Date of issuance: December 31, 1998.

Effective date: December 31, 1998.

Amendment No.: 189.

Facility Operating License No. DPR-40: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38137). The March 18, 1998, and November 17, 1998, supplemental letters

provided additional clarifying information and did not change the

original no significant hazards consideration determination. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated December 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

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

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

Date of application for amendment: April 14, 1997, as supplemented

October 17, 1997, March 20, 1998, May 18, 1998, and August 17, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications to allow for a Safety Review Committee review of plant

performance as opposed to an audit of plant performance and replaces

the position title of Vice President Regulatory Affairs and Special

Projects with Director Regulatory Affairs and Special Projects.

Date of issuance: December 30, 1998.

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

30 days.

Amendment No.: 186.

Facility Operating License No. DPR-64: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 27, 1997 (62 FR

45460).

The October 17, 1997, March 20, 1998, May 18, 1998, and August 17,

1998, letters provided clarifying information that did not change the

proposed no significant hazards consideration determination. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated December 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: March 22, 1996, as revised and

supplemented on February 6, 1998, April 17, 1998, and October 30, 1998.

Brief description of amendment: The amendment provides function-

specific actions and allowed outage times for certain instrumentation,

and relocates some instrumentation requirements to licensee-controlled

documents.

Date of issuance: January 12, 1999.

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

60 days.

Amendment No.: 250.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 8, 1996 (61 FR

20855).

The revision and supplemental information provided on February 6,

1998, April 17, 1998, and October 30, 1998, provided clarifying

information that did not change the initial proposed no significant

hazards consideration. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated January 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of application for amendments: June 30, 1997.

Brief description of amendments: The amendments delete License

Condition 2.C(19)b for San Onofre Nuclear Generating Station (SONGS)

Unit 2 and revises TSs 3.3.1, 3.3.2, 3.3.5, 3.3.10, 3.3.11, 3.4.7,

3.4.12.1, 3.7.5, 5.5.2.10 and 5.5.2.11 for both SONGS units. These

changes reinstate provisions of the SONGS Units 2 and 3 TS previously

revised as part of NRC Amendment Nos. 127 and 116, respectively, make

corrections to the TS, or remove information inadvertently added to the

TS that are not applicable to the SONGS units design.

Date of issuance: December 22, 1998.

Effective date: December 22, 1998, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--147; Unit 3--139.

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

revised Facility Operating License No. NPF-10 and the technical

specifications for both licenses.

Date of initial notice in Federal Register: March 11, 1998 (63 FR

11921). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713.

Dated at Rockville, Maryland, this 20th day of January 1999.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-1705 Filed 1-26-99; 8:45 am]

BILLING CODE 7590-01-P