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

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

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

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

publish notice of any

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amendments issued, or proposed to be issued, under a new provision of

section 189 of the Act. This provision grants the Commission the

authority to issue and make immediately effective any amendment to an

operating license upon a determination by the Commission that such

amendment involves no significant hazards consideration,

notwithstanding the pendency before the Commission of a request for a

hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from March 2, 1998, through March 13, 1998. The

last biweekly notice was published on March 11, 1998 (63 FR 11913).

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 April 24, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the

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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.

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick

Steam Electric Plant, Unit 1, Brunswick County, North Carolina

Date of amendment request: February 23, 1998.

Description of amendment request: The amendment request proposes

changes to the Brunswick Steam Electric Plant Unit 1 Technical

Specifications (TS) in support of Cycle 12 operation, including a

change to the Minimum Critical Power Ratio safety limit (safety limit

MCPR) to a value equivalent to the generic safety limit MCPR for

General Electric type GE-13 fuel. The request would additionally remove

a footnote limiting the stated value for the safety limit MCPR to a

specific fuel cycle and reference to an NRC safety evaluation

documenting acceptance of methods used for determining the current

cycle safety limit MCPR. The amendment request is provided both in the

format of the current TS as well as improved Standard Technical

Specifications (iSTS). The Brunswick licensee applied for conversion to

ISTS on November 1, 1996, as supplemented on October 13, 1997, and

February 26, 1998, and that application is currently undergoing NRC

staff review. For iSTS, the licensee has proposed two safety limits

MCPR, one pertaining to two-recirculation loop operation and the other

to single-recirculation loop operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed license amendment establishes a revised safety

limit MCPR value of 1.09 [two-recirculation loop and 1.10 for

single-recirculation loop operation] for use during Unit 1 Cycle 12

operation. General Electric (GE) has determined that both generic

and plant-specific evaluations [two-loop operation] yield the same

calculated safety limit MCPR value. Additionally, a document

referenced by the Technical Specification 6.9.3.2 of methodologies

used in determining core operating limits is being removed.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits have been established, consistent with NRC[-]

approved methods, to ensure that fuel performance during normal,

transient, and accident conditions is acceptable.

The probability of an evaluated accident is not increased by

revising the safety limit MCPR value to 1.09 [two-loop/1.10 single-

loop]. The change does not require any physical plant modifications

or physically affect any plant components. Therefore, no individual

precursors of an accident are affected.

The proposed license amendment establishes a revised safety

limit MCPR that ensures the fuel is protected during normal

operation and during any plant transients or anticipated operational

occurrences. Specifically, the reload analysis demonstrates that a

safety limit MCPR value of 1.09 [two-loop/1.10 single-loop] ensures

that less than 0.1 percent of the fuel rods will experience boiling

transition during any plant operation if the limit is not violated.

The methods for calculating the safety limit MCPR have been

approved by the NRC and are described in GE's reload licensing

methodology topical report NEDE-24011, ``General Electric Standard

Application for Reactor Fuel (GESTAR II).'' Based on (1) the

determination of the new safety limit MCPR value using conservative

approved methods, and (2) the operability of plant systems designed

to mitigate the consequences of accidents not having been changed;

the consequences of an accident previously evaluated have not been

increased.

Additionally, removal of the footnote on the safety limit MCPR

value in Technical Specification 2.1.2 and removal of reference

``c'' from the document list in Technical Specification 6.9.3.2 will

not increase the probability or consequences of accidents previously

evaluated. The footnote on the safety limit MCPR value in Technical

Specification 2.1.2 and reference ``c'' in Technical Specification

6.9.3.2 were associated with the safety limit MCPR value of 1.10 for

Unit 1 Cycle 11 operation. Since the current safety limit MCPR value

of 1.10 applies only to Unit 1 Cycle 11 operation, the footnote on

the safety limit MCPR value in Technical Specification 2.1.2 and the

reference ``c'' in Technical Specification 6.9.3.2 are no longer

needed and should be deleted. Thus, removal of the footnote on the

safety limit MCPR value in Technical Specification 2.1.2 and removal

of reference ``c'' from Technical Specification 6.9.3.2 is an

administrative change that has no effect on the probability or

consequences of accidents previously evaluated.

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

kind of accident from any accident previously evaluated?

This proposed license amendment involves a revision of the

safety limit MCPR from 1.10 to 1.09 [two-loop/1.10 single-loop]

based on the results of both cycle-specific and generic analyses,

removal of the footnote on the safety limit MCPR value in Technical

Specification 2.1.2, and the removal of a document reference listed

in Technical Specification 6.9.3.2 describing the methods used only

during Unit 1 Cycle 11 to determine core operating limits. Creation

of the possibility of a new or different kind of accident would

require the creation of one or more new precursors of that accident.

New accident precursors may be created by modifications of the plant

configuration, including changes in allowable modes of operation.

This proposed license amendment does not involve any modifications

of the plant configuration or changes in the allowable modes of

operation. Therefore, no new precursors of an accident are created

and no new or different kinds of accidents are created.

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

of safety?

As previously stated, the methods for calculating the safety

limit MCPR have been previously approved by the NRC and are

described in GE's reload licensing methodology topical report NEDE-

24011. Use of these methods ensures that the resulting safety limit

MCPR satisfies the fuel design safety criteria that less than 0.1

percent of the fuel rods experience boiling transition if the safety

limit is not violated. Based on the assurance that the fuel design

safety criteria will be met, the proposed license amendment does not

involve a significant reduction in a margin of safety.

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

this review, it appears that the three

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standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff

proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Pao-Tsin Kuo (Acting).

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of amendment request: May 16, 1997.

Description of amendment request: The proposed changes would

replace the existing Technical Specification (TS) 4.6.2.3 a.2 cooling

water flow rate of 1425 gpm with a new value of 1300 gpm.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Cooling water flow to the Containment Fan Coolers is provided by

the Emergency Service Water (ESW) System, and Emergency Service

Water is not an initiating system in any FSAR [Final Safety Analysis

Report] Chapter 15 analyses. Revising the minimum cooling water flow

to the Containment Fan Coolers will not increase the probability of

initiating any previously evaluated accident, because Containment

Fan Cooler performance and integrity will not be adversely affected.

The heat removal capacity of the Containment Fan Coolers will be

maintained consistent with the assumptions used in the existing HNP

[Harris Nuclear Plant] containment analyses, and, therefore,

containment integrity should not be challenged.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

(2) The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed amendment will not create any new accident

scenarios, because the change does not introduce any new single

failures, adverse equipment or material interactions, or release

paths.

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 amendment does not involve a significant

reduction in the margin of safety.

Although the proposed amendment replaces the TS 4.6.2.3 a.2

cooling water flow rate of 1425 gpm with a lower flow rate of 1300

gpm, a cooling water flow rate of greater than or equal to 1300 gpm

maintains adequate heat removal capacity as required by existing HNP

containment analyses. The Bases for TS 4.6.2.3 a.2 is to ensure that

adequate heat removal capacity is available, when the Containment

Fan Coolers are operated in conjunction with the Containment Spray

Systems, during post-LOCA [Loss-of-Coolant Accident] conditions to

prevent the pressure inside containment from exceeding its design

rating.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Pao-Tsin Kuo (Acting).

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit No. 2, Shippingport, Pennsylvania

Date of amendment request: October 22, 1997

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TSs) by reducing the reactor

coolant system (RCS) specific activity limits in accordance with

Generic Letter 95-05. The definition of DOSE EQUIVALENT I-131 would be

replaced with the Improved Standard TS definition wording in the first

sentence and an equation added based on dose conversion factors derived

from International Commission on Radiation Protection (ICRP) ICRP-30.

TS 3.4.8, Specific Activity, would be revised by reducing the DOSE

EQUIVALENT I-131 limit from 1.0 [micro] Ci[curies]/gram to 0.35

[micro]Ci[curies]/gram. Item 4.a in TS Table 4.4-12, Primary Coolant

Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the

Bases for TS 3/4.4.8 would be modified to reflect the reduced DOSE

EQUIVALENT I-131 limit.

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 change reduces the reactor coolant system (RCS)

specific activity limits of Specification 3.4.8 from 1.0 [micro]Ci/

gram to 0.35 [micro]Ci/gram and lowers the graph in Figure 3.4-1 by

39 [micro]Ci/gram following the guidance provided in Generic Letter

(GL) 95-05. This reduces the RCS acvitity allowed to leak to the

secondary side when the plant is operating so that additional margin

is available to support a higher allowable accident-induced leakage

value as justified by analysis.

The proposed changes to Specification 3.4.8 and the definition

of DOSE EQUIVALENT I-131 ensure these requirements are consistent

with the latest analyses.

These changes implement the more restrictive RCS activity limits

in accordance with applicable analyses and GL 95-05 to ensure the

regulations are satisfied. Therefore, these changes do 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?

The proposed change does not alter the configuration of the

plant or affect the operation with the reduced specific activity

limit. By reducing the specific activity limit, the limit would be

reached sooner to initiate evaluation of the out of limit condition.

The proposed changes will not result in any additional challenges to

the main steam system or the reactor coolant system pressure

boundary. Consequently, no new failure modes are introduced as a

result of the proposed changes. As a result, the main steam line

break, steam generator tube rupture and loss of coolant accident

analyses remain bounding. Therefore, the proposed change 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 proposed change reduces the RCS specific activity limit to

0.35 [micro]Ci/gram along with lowering the Figure 3.4-1 limits by

39 [micro]Ci/gram. Reduction of the RCS specific activity limits

allows an increase in the limit for the projected SG [steam

generator] leakage following SG tube inspection and repair in

accordance with the voltage-based SG tube alternate repair criteria

(ARC). This follows the guidance provided in GL 95-05 and

effectively takes margin available in the specific activity limits

and applies it to the projected SG leakage for the ARC. This has

been determined to be an acceptable means for accepting higher

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projected leakage rates while still meeting the applicable limits of

10 CFR [Part] 100 and GDC [General Design Criterion] 19 with respect

to offsite and control room doses.

The capability for monitoring the specific activity and

complying with the required actions remains unchanged. In addition,

there is no resultant change in dose consequences. Therefore, the

proposed change does not involve a significant reduction in a margin

of safety.

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

this review, it appears that the three standards of 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: John F. Stolz.

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 amendment request: March 3, 1998.

Description of amendment request: The licensee proposed to revise

Section 6.2.3.2 of the units' Technical Specifications. Currently, this

section prescribes that the Catawba Safety Review Group (SRG) be

composed of at least five individuals and at least three of these shall

have a bachelor's degree in engineering or related science and at least

2 years professional level experience in his/her field, at least 1 year

of which experience shall be in the nuclear field. The licensee

proposed to revise this section to provide the option of replacing one

of the three degreed individuals with one with at least 15 years of

professional level experience in his/her field, at least 10 years of

which experience shall be in the nuclear field, at least 3 years of

which nuclear experience shall be supervisory/managerial experience in

engineering, and shall hold or have held a Senior Reactor Operator

license. The licensee also proposed to editorially revise this section.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below.

1. Would the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed amendment would only change administrative

requirements related to personnel qualifications for one of the five

SRG [Safety Review Group] positions. The SRG is an oversight group,

and the individual who meets the new qualification requirements

would be expected to perform at the same level of quality as an

individual who meets the current qualification requirements.

Changing qualification requirements for an individual who primarily

performs an oversight function will not have any direct effect on

the design or operation of any plant structures, systems, or

components. No previously analyzed accidents were initiated by the

functions of the SRG, and the SRG was not a factor in the

consequences of previously analyzed accidents. Therefore, the

proposed change would have no impact on the consequences or

probabilities of any previously evaluated accidents.

2. Would the change create the possibility of a new or

difference kind of accident from any accident previously evaluated?

No. The proposed change would not lead to any hardware or

operating procedure change. Hence, no new equipment failure modes or

accidents from those previously evaluated will be created.

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

of safety?

No. Margin of safety is associated with confidence in the design

and operation of the plant. The proposed change to the Technical

Specifications does not involve any change to plant design or

operation. Thus, the margin of safety previously analyzed and

evaluated is maintained.

Based on this analysis, 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Project Director: Herbert N. Berkow.

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

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

Date of amendment request: February 7, 1997.

Description of amendment request: The proposed amendment, if

approved, would revise Technical Specification (TS) as delineated

below:

1. 4160 Volt Tie From Unit 2.

TS sections 3.7.2.b & d to delete reference to the optional use of

the 4160 volt tie from the unit 2 transformer.

2. Emergency Load Sequence and Power Transfer.

a. The testing required by Section 4.5.1.1.b of the TS would be

considered satisfactory if the pumps have started and valves have

completed travel. The need to evidence the successful starting of pumps

and fans and the complete travel of valves by observation of control

board component operating lights will be deleted. Neither would a

second means of verification, such as: the station computer or control

board indicating lights initiated by separate limit switch contacts be

required.

b. Section 4.5.1.2.b would be revised in the same manner as

4.5.1.1.b above.

3. Reactor Building Cooling and Isolation System.

a. Section 4.5.3.1.a.1 of the TS would be revised to delete the

need to simultaneously test start a spray pump using a Reactor Building

30-psi high pressure test signal while testing the emergency loading

sequence.

The proposed change also eliminates the need to evidence the

successful starting of the spray pumps by observation of the control

board indicating lights or the use of the station computer for Sections

4.5.3.1.a.1 and 4.5.3.1.b.2.

4. Instrument Surveillance Requirements.

Table 4.1-1 of the TS would be revised to delete the strong motion

accelerometer and its quarterly battery check surveillance requirement.

5. Air Intake Tunnel (AIT) Fire Protection Systems.

Section 5.5 of the TS would be deleted. The description of the

equipment contained in Section 5.5 would be transferred to the Final

Safety Analysis Report (FSAR).

6. Hydrogen Recombiner System.

The Bases for Section 4.4.4 TS would be changed to reflect a

reduction in the time interval for operation of the hydrogen recombiner

following a loss of cooling accident (LOCA) from 9.8 to 9 days.

7. Various editorial and typographical errors would be corrected.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated. The revised TS eliminate overly prescriptive

requirements for evidencing component performance, the requirement

for redundant diesel block loading tests,

[[Page 14487]]

instrumentation from SR [surveillance requirement] tables having no

associated LCO [limiting condition for operation], AIT fire

protection systems descriptive text, and correct previous

typographical errors. Several of the proposed revisions involve

changes which are consistent with NUREG-1430, the Revised Standard

Technical Specifications (RSTS) for B&W plants. The reliability of

systems and components depended upon to prevent or mitigate the

consequences of accidents previously evaluated is not degraded by

the proposed changes because assurance of system and equipment

availability is maintained by surveillance testing program

requirements.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated. The revised

surveillance requirements create no new failure modes. Verification

of equipment operation continues to be required by plant procedures.

Elimination of the AIT fire protection system descriptive text from

the TSs would not create a new or different kind of accident since

the change has no effect on surveillance methodology and frequency

requirements. They are maintained in the Fire Protection Program.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety because no operating limits are affected.

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: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Cecil O. Thomas, Director.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: March 3, 1998.

Description of amendment request: The proposed revision to the

Millstone Unit 3 licensing basis would eliminate the requirement to

have the recirculation spray system directly inject into the reactor

coolant system following a design basis accident.

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:

Northeast Nuclear Energy Company (NNECO) has reviewed the

proposed revision in accordance with 10CFR50.92 and has concluded

that the revision does not involve a significant hazards

consideration (SHC). The basis for this conclusion is that the three

criteria of 10CFR50.92(c) are not satisfied. The proposed revision

does not involve an SHC because the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The change to the Emergency Operating Procedures (EOP) to

eliminate the use of Recirculation Spray System (RSS) direct

injection during cold and hot leg recirculation does not effect the

probability of any accident. The elimination of the requirement to

have RSS directly [inject] into the reactor coolant system did not

increase the consequences of the previously evaluated accidents.

These consequences were evaluated based on very conservative

assumptions concerning the containment pressure after the design

basis Loss of Coolant Accident (LOCA), containment integrated

leakage rates, and the fraction of the sprayed volume. None of these

assumptions were affected by the elimination of the direct cold-leg

injection.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The modification to the RSS did not create the possibility of a

new or different accident from those previously analyzed. The change

involved elimination of the direct injection flow path from the

design basis of the system but did not involve physical

modifications to the system itself. The operability of the affected

valves within the direct injection alignments remained unchanged and

these paths were still available to the operators for contingencies

beyond the design basis. The EOPs provided clear and explicit

guidance to that effect.

Therefore, the proposed revision does not create the possibility

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

evaluated.

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

In considering the impact on the margin of safety as defined in

the bases of the Technical Specifications, the impact of the change

on the design basis analysis of the fission product barriers must be

evaluated.

The minimum Emergency Core Cooling System flow requirement for

long-term core cooling is that the modified alignment deliver

sufficient flow to satisfy the inventory lost to the boil off in the

vessel due to the decay heat and the extended boiling from hot metal

in the downcomer and the lower plenum. The analysis determined that

these requirements were being met.

The elimination of the direct injection resulted in a flow

reduction through the RSS heat exchanger, from approximately 4000

gpm [gallons per minute] to 1200 gpm, thus reducing the rate of the

heat transfer from the containment to the service water system. The

design basis of the containment heat removal systems (circa 1986) is

that the containment pressure will decrease to subatmospheric within

one hour after the Design Basis Accident to compensate for the

reduction in heat removal from the containment, a smaller allowable

RSS pump degradation was assumed in the revised containment

analysis. The original RSS pump performance curve was based on a 10

percent reduction in developed head from the design curve. For the

modification, a 5 percent reduction was used. The results of the

analysis show that with these changes the design basis of

maintaining subatmospheric containment pressure was met.

Based on the above, elimination of the direct injection did not

reduce the margin of safety because there was no violation of the

acceptance limits and no weakening of the protective boundaries.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

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 amendment requests: November 2, 1995, as supplemented by

letter dated January 9, 1998. The January 9, 1998, submittal supersedes

the staff's proposed no significant hazards consideration determination

evaluation for the requested changes that was published on April 10,

1996 (61 FR 15995).

Description of amendment requests: In the November 2, 1995, letter,

the

[[Page 14488]]

licensee proposed to revise Technical Specification (TS) 3.8.1, ``AC

Sources--Operating,'' to extend the offsite circuit completion time and

to extend the allowed outage time for an emergency diesel generator.

The January 9, 1998, letter modifies the original request to (1)

further extend the offsite completion time and allowed outage time for

an emergency diesel generator, and (2) add a new TS 5.5.2.14,

``Configuration Risk Management Program,'' that ensures a

proceduralized probabilistic risk assessment-informed process is in

place that assesses the overall impact of plant maintenance on plant

risk.

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 Emergency Diesel Generators (EDGs) are backup alternating

current power sources design to power essential safety systems in

the event of a loss of offsite power. EDGs are not accident

initiators in any accident previously evaluated. Therefore, this

change does not involve an increase in the probability of an

accident previously evaluated.

The EDGs provide backup power to components that mitigate the

consequences of accidents. The proposed changes to the Completion

Times do not affect any of the assumptions used in the deterministic

safety analysis.

To fully evaluate the effect of the EDG Completion Time

extension, Probabilistic Safety Analysis (PSA) methods were

utilized. The results of these analyses show no significant increase

in the core damage frequency. As a result, there would be no

significant increase in the consequences of accidents previously

evaluated.

The Configuration Risk Management Program is an Administrative

Program that assesses risk based on plant status. Adding the

requirement to implement this program for Technical Specification

3.8.1 does not affect the probability or the consequences of an

accident.

Therefore, this change does not involve a significant increase

in the probability or consequences of any 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 proposed change does not alter the design, configuration,

or method of operation of the plant. Therefore, this change does not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the Limiting Conditions for

Operation or their Bases that are used in the deterministic analyses

to establish the margin of safety. PSA evaluations were used to

evaluate these changes and these evaluations determined that the

changes are either risk neutral or risk beneficial.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

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 amendment requests: December 19, 1997.

Description of amendment requests: The licensee proposed to revise

Technical Specification (TS) 3.4.9, ``Pressurizer,'' to reduce the

allowable pressurizer water volume for pressurizer operability.

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 limiting events impacted by this Technical Specification

change have been reanalyzed. These events are the Chemical and

Volume Control System (CVCS) Malfunction and CVCS Malfunction With a

Concurrent Single Failure of an Active Component, Inadvertent

Operation of the Emergency Core Cooling System (ECCS) During Power

Operation (Including Single Failure of an Active Component), and

Feedwater System Pipe Breaks. The probability of these events is not

changed by the restriction of the pressurizer level to 57%. An

operator action time of 15 minutes has been identified for the CVCS

malfunction and inadvertent ECCS operation events. Based on the

availability of operator alarms and indications and operator

Simulator training, 15 minute operator action is sufficient to

recognize and mitigate the inadvertent CVCS or ECCS operation.

Therefore, this change will not involve an increase in the

probability or consequences of any previously evaluated accident.

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

or different kind of accident from any accident previously

evaluated.

This amendment request does not involve any change to plant

equipment or operation. All the events identified in Chapter 15 of

the Updated Final Safety Analysis Report (UFSAR) were evaluated to

determine the impact of the change in pressurizer level. In addition

to the normally analyzed Inadvertent Operation of the ECCS During

Power Operation event a concurrent single failure of an active

component was considered in this evaluation. The analysis of this

event with single failure of an active component produced

consequences that are bounded by the CVCS malfunction with single

failure of an active component. No new or different kind of accident

will be created as a result of this Technical Specification change.

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

different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

This amendment request does not change the manner in which

safety limits, limiting safety settings, or limiting conditions for

operation are determined. There are no changes to the acceptance

criteria for these events as a result of the proposed reduction in

the maximum pressurizer water level. This change does not reduce a

margin of safety since it lowers allowed pressurizer operational

level to 57%. An operator action time of 15 minutes has been

identified for the CVCS malfunction and inadvertent ECCS operation

events. Based on the availability of operator alarms and

indications, and demonstrated operator response in Simulator

training, 15 minute operator action has been demonstrated to be

adequate to recognize and mitigate the inadvertent CVCS or ECCS

operation. Therefore, this proposed change does not involve a

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

[[Page 14489]]

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 amendment requests: January 2, 1998.

Description of amendment requests: The licensee proposed to revise

Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW)

System,'' to indicate the turbine driven AFW pump is operable when

running in the manual mode to support plant startups, shutdowns, and

testing.

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.

Probabilistic analyses have been performed in support of

declaring P140 operable when the pump is manually actuated and

operating.

The results show that, considering P-140 to be in test for an

entire year, the core damage risk of a Main Steam Line Break/

Feedwater Line Break (MSLB/FWLB) slightly increases (4.3E-8/yr)

while the risk due to other initiating events decreases (3E-7/yr).

The net core damage impact of P-140 in test for an entire year is a

Core Damage Frequency (CDF) decrease of 2E-7/yr. Having P140

operating instead of being in standby increases its reliability.

This increased reliability reduces the risk due to other initiating

events, such as loss of main feedwater, medium and small Loss of

Coolant Accidents (LOCAs), Steam Generator Tube Rupture (SGTR), and

Loss of Offsite Power (LOP), which require Auxiliary Feedwater (AFW)

and which occur with much greater frequency than MSLB/FWLB. With the

overall CDF reduction a result of considering P140 being in a test

configuration for an entire year, the actual cumulative risk

incurred is the weighted fraction that P140 is in the test

configuration over a year period. Based on past experience, the pump

is running in manual approximately 500 minutes/year, which results

in an annual net cumulative CDF reduction on the order of 2E-10/yr

due to running P140 in the manual mode.

Therefore, the operation of the facility in accordance with this

proposed change does not involve a significant 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 change does not involve a plant hardware modification or

allow the operation of any plant equipment in any way other than

originally designed. This change only affects the administrative

tracking of the turbine-driven AFW pump when the steam driven AFW

pump is operating in the manual mode.

Therefore, the operation of the facility in accordance with this

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Pump history shows the pump is run approximately 500 minutes per

year. In all cases except for the one postulated scenario of the

Main Steam Isolation Signal followed by an Emergency Feedwater

Actuation Signal the turbine-driven AFW pump is not susceptible to

being tripped. Also, this postulated scenario does not affect the

capability of the motor-driven AFW pumps.

Even though there is a small increase in the CDF from the AFW

steam driven pump operating in manual mode based on the possibility

of a MSLB/FWLB, also considering other initiating events results in

an annual net cumulative CDF reduction on the order of 2E-10/yr due

to P140 running in the manual mode.

Therefore, the operation of the facility in accordance with this

proposed change does not involve a significant reduction in the

margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

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

Yankee Nuclear

Date of amendment request: August 20, 1997, as supplemented by

letters dated September 18, 1997 and October 31, 1997.

Description of amendment request: The proposed change would revise

the Vermont Yankee Technical Specifications Section 6.0,

``Administrative Controls,'' to add and revise reference to NRC-

approved methodologies which will be used to generate the cycle-

specific thermal operating limits in the Vermont Yankee Core Operating

Limits Report.

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 will not involve any significant increase

in the probability or consequences of an accident previously

evaluated.

The change updates the Technical Specifications to include an

NRC approved method reference to allow calculation of thermal limits

with a revised method. It does not affect plant operation and will

not weaken or degrade the facility.

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

or different kind of accident since the change is administrative. No

physical alterations of the plant, setpoint changes, or operating

conditions are proposed.

3. The proposed change will not involve a significant reduction

in a margin of safety. The change involves an update to the

Administrative Controls in Section 6.0 of the Technical

Specifications by adding a reference to NRC approved methods. This

administrative change does not alter plant safety margins.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W., Washington, DC 20037-1128.

NRC Project Director: Cecil O. Thomas, Director.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 4, 1998.

Description of amendment request: The amendment would revise

Technical Specification 3.2.4, quadrant power tilt ratio (QPTR), and

associated Bases, to clarify the required actions for the limiting

condition for operation (LCO) and other changes consistent with the

technical specification conversion application submitted by letter

dated May 15, 1997.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

[[Page 14490]]

1. Requirements for Determining QPTR

The Action to calculate QPTR once per hour until THERMAL POWER

was reduced to less than 50% RATED THERMAL POWER (RTP) when QPTR

exceeds the LCO requirements would be deleted and replaced by a new

requirement to determine QPTR at least once per 12 hours.

The proposed change involves only the compensatory measures to

be taken should the QPTR be outside its limit. The frequency with

which QPTR is calculated is not assumed in the initiating events for

any accident previously evaluated. In addition, the change does not

involve any new operating activities or hardware change. Therefore,

the proposed change would not significantly increase the probability

of an accident previously evaluated.

Once THERMAL POWER has been reduced appropriately in proportion

to the amount that QPTR exceeds 1.00, any additional change would be

sufficiently slow that a 12-hour interval for recalculating QPTR

will provide an adequate level of protection. Therefore, the

proposed change will not significantly increase the consequences of

any accident previously evaluated.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip

Setpoints

The proposed change to allow 72 hours for resetting the Power

Range Neutron Flux-High trip setpoints involves only the

compensatory measures to be taken should the QPTR be outside its

limit. These compensatory measures are not assumed in the initiating

events for any accident previously evaluated. The proposed actions

recognize that the required reduction in power (3% for each 1% of

indicated QPTR in excess of 1.00) provide adequate margin for fuel

design limits so that consequences of assumed accidents would not be

significantly affected. Therefore, the proposed change will not

adversely affect the probability or consequences of any accident

previously evaluated. Further, by permitting more time to perform

resetting the trip setpoints, the chances of a transient may be

reduced.

3. Delete(tion) of the Actions (a.3., a.4.) for verifying QPTR to be

restored within 24 hours and for identifying and correcting the cause

of the out-of-limit condition prior to increasing THERMAL POWER

The proposed changes would delete current Actions a.3. and a.4.

and add new Actions for QPTR out of limit including requirements for

measuring FQ(Z) and F N delta H prior to and

following a return to power and performing safety analyses to verify

safety requirements are met prior to increasing power above the

limits of Action a.1. The proposed changes involve only the

compensatory measures to be taken should the QPTR be outside its

limit. These compensatory measures are not assumed in the initiating

events for any accident previously evaluated. Therefore, the

proposed change will not affect the probability or consequences of

any accident previously evaluated.

4. Deletion of the Actions for QPTR in excess of 1.09

The proposed change would delete the required Actions for QPTR

in excess of 1.09 and Actions for QPTR in excess of 1.02 are

followed for all instances where QPTR exceeds 1.02. The proposed

change involves only the compensatory measures to be taken should

the QPTR be outside its limit. These compensatory measures are not

assumed in the initiating events for any accident previously

evaluated. The proposed actions recognize that the required

reduction in power (3% for each 1% of indicated QPTR in excess of

1.00) provide adequate margin for fuel design limits so that

consequences of assumed accidents would not be significantly

affected. Therefore, the proposed change will not affect the

probability or consequences of any accident previously analyzed.

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

or different kind of accident from any accident previously

evaluated.

1. Requirements for Determining QPTR

The proposed change for calculating QPTR once every 12 hours

does not involve a physical alteration to the plant or change the

method by which any safety-related system performs its function. The

manner in which the plant would be operated would not be altered.

Therefore, the proposed change will not create the possibility of a

new or different kind of accident from any previously evaluated.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip

Setpoints

The proposed change to allow 72 hours for resetting the Power

Range Neutron Flux-High trip setpoints does not involve a permanent

physical alteration to the plant; no new or different kinds of

equipment will be installed. The change would not alter the manner

in which the plant would be operated only the timing of actions that

provide potential mitigation of accidents. Thus, the change would

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

from any accident previously evaluated.

3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored

within 24 hours and for identifying and correcting the cause of the

out-of-limit condition prior to increasing THERMAL POWER

The proposed changes would delete current Actions a.3, and a.4.

and add new Actions for QPTR out-of-limit including requirements for

measuring FQ(Z) and F N delta H prior to and

following a return to power and performing safety analyses to verify

safety requirements are met prior to increasing power above the

limits of Action a.1. The proposed changes do not involve a physical

alteration to the plant; no new or different kinds of equipment

would be installed. The changes would not alter the manner in which

the plant would be operated only the timing of actions that provide

potential mitigation of accidents. Thus, the changes would not

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

any accident previously evaluated.

4. Deletion of the Actions for QPTR in excess of 1.09

The proposed change would delete the required Actions for QPTR

in excess of 1.09 and Actions for QPTR in excess of 1.02 are

followed for all instances where QPTR exceeds 1.02. The proposed

change does not involve a physical alteration to the plant or

changes in the way in which the plant is operated. The proposed

change involves only the compensatory measures to be taken should

QPTR be outside its limit. The assumptions of the accident analyses

are unaffected by the proposed change. No new permutations or event

initiators are introduced by the proposed alternate methods of

dealing with QPTRs in excess of 1.09. Therefore, there is no

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

1. Requirements for Determining QPTR

The proposed change for calculating QPTR once every 12 hours

does not change any accident analysis assumptions, initial

conditions or results. The proposed change will continue to ensure

that the plant is maintained in a safe condition while QPTR is in

excess of its limit. Additionally, calculating QPTR once per 12

hours as opposed to every hour while QPTR is in excess of its limit

would avoid the diversion of personnel resources from corrective

actions with regard to meeting the LCO. Therefore, the proposed

change will not involve a significant reduction in any margin of

safety.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip

Setpoints

The proposed change to allow 72 hours for resetting the Power

Range Neutron Flux-High trip setpoints will continue to ensure that

the plant is maintained in a safe condition within the envelope of

the safety analyses while QPTR is in excess of its limit. The

proposed actions recognize that the required reduction in power (3%

for each 1% of indicated QPTR in excess of 1.00) provide adequate

margin for fuel design limits so that consequences of assumed

accidents would not be significantly affected. Therefore, the

proposed change will not involve a significant reduction in any

margin of safety.

3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored

within 24 hours and for identifying and correcting the cause of the

out-of-limit condition prior to increasing THERMAL POWER

The proposed changes would delete current Actions a.3. and a.4

and add new Actions for QPTR out-of-limit including requirements for

measuring FQ(Z) and F N delta H prior to and

following a return to power and performing safety analyses to verify

safety requirements are met prior to increasing power above the

limits of Action a.1. The proposed changes will continue to ensure

that the plant is maintained in a safe condition within the envelope

of the safety analysis while QPTR is in excess of its limit.

Therefore, the proposed changes will not involve a significant

reduction in any margin of safety.

4. Deletion of the Actions for QPTR in excess of 1.09

The proposed change would delete the required Actions for QPTR

in excess of 1.09

[[Page 14491]]

and Action for QPTR in excess of 1.02 are followed for all instances

where QPTR exceeds 1.02. The proposed change will continue to ensure

that the plant is maintained in a safe condition within the envelope

of the safety analyses while QPTR is in excess of its limit. While

different actions are taken in response to a QPTR in excess of 1.09,

the proposed change will assure that accident analyses assumptions

continue to be met. Therefore, the proposed changes will not involve

a significant reduction in any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

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

NRC Project Director: William H. Bateman.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 4, 1998.

Description of amendment request: The amendment would revise the

technical specifications to (1) create separate functional units for

the analog and digital portions of the engineered safety features

actuation system (ESFAS) function associated with starting the turbine-

driven auxiliary feedwater pump on a loss of offsite power, and (2) add

a table notation to clarify that the testing of the time delay relays

for the 4 kV undervoltage, loss of voltage and grid degraded voltage

portion of the ESFAS is performed as part of the channel calibration.

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.

Overall protection system performance will remain within the

bounds of the previously performed accident analyses since no

hardware changes are proposed. The recognition that different

OPERABILITY and surveillance requirements apply to analog vs.

digital circuitry does not impact any previously analyzed accidents.

The clarification that testing of the time delay relays is performed

as part of the CHANNEL CALIBRATION does not impact any previously

analyzed events. The proposed change will not affect any of the

analysis assumptions for any of the accidents previously evaluated.

The proposed change does not alter the current method or procedures

for meeting the surveillance requirements in Table 4.3-2. The

proposed change will not affect the probability of any event

initiators nor will the proposed change affect the ability of any

safety-related equipment to perform its intended function. There

will be no degradation in the performance of nor an increase in the

number of challenges imposed on safety-related equipment assumed to

function during an accident situation. Therefore, the proposed

change does not involve a significant 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.

There are no hardware changes nor are there any changes in the

method by which any safety-related plant system performs its safety

function. The separation of analog and digital portions of

Functional Unit 6.f or the clarification of testing of the time

delay relays will not impact the normal method of plant operation.

The OPERABILITY requirements, ACTION Statement, and surveillance

requirements for the analog portion, new Functional Unit 6.f.1), are

identical to those of Functional Unit 8.a, while the requirements

for the digital portion, new Functional Unit 6.f.2), are consistent

with the current technical specifications, other than the new ACTION

Statement 30 provisions that defer to the TDAFW pump Specification

3.7.1.2 requirements and the performance of a TADOT during

appropriate plant conditions. These changes do not change any ESFAS

design standard and are appropriate for digital functions such as

this.

Testing of the time delay relays has been performed as part of

the 18 month CHANNEL CALIBRATION. The tolerancesfor the time delay

relays are sufficient to account for relay drift encountered during

the 18 month surveillance testing. The calculated tolerances for the

time delay setpoints have been evaluated to insure that safety-

related systems, subsystems and components would not be adversely

affect[ed] by the drift within the permissible tolerance band.

No new accident scenarios, transient precursors, failure

mechanisms, or limiting single failures are introduced as a result

of this change. Therefore, the proposed change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not affect the acceptance criteria for

any analyzed event. There will be no effect on the manner in which

safety limits or limiting safety system settings are determined nor

will there be any effect on those plant systems necessary to assure

the accomplishment of protection functions. There will be no impact

on any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

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

NRC Project Director: William H. Bateman.

Previously Published Notices of Consideration of Issuance of

Amendments toFacility 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.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: January 26, 1998.

Brief description of amendment request: The proposed amendment

would change the SSES Technical Specifications facility staff

requirements to allow an individual who does not hold a current senior

reactor operator (SRO) license to hold the position of Manager-Nuclear

Operations (MNO) and require an individual serving in the capacity of

the Operations Supervisor-Nuclear to hold a current SRO license

[[Page 14492]]

and report directly to the MNO and be responsible for directing the

licensed activities of licensed operators.

Date of publication of individual notice in Federal Register:

February 24, 1998 (63 FR 9270).

Expiration date of individual notice: March 26, 1998.

Local Public Document Room location: Osterhout Free Library,

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

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: February 25, 1998, TXX-98050.

Description of amendment request: The proposed amendment would be a

temporary change to the Technical Specifications to remove the

requirement to demonstrate the load shedding feature of MCC XEB4-3 as

part of Surveillance Requirements (SRs) 4.8.1.1.2f.4)a) and

4.8.1.1.2f.6)a) until the plant startup subsequent to the next

refueling outage or until an outage of greater than 24 hours in

duration for each respective unit. This temporary change is requested

as a result of the failure to confirm the load shedding feature of MCC

XEB4-3 during the last performance of these SRs for the Unit 1 and Unit

2 train B diesel generators (DGs).

Date of individual notice in the Federal Register: March 9, 1998,

(63 FR 11458).

Expiration date of individual notice: April 8, 1998.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: March 18, 1997, as supplemented

by letters dated July 28, 1997, and September 9, 1997.

Brief description of amendments: The amendments revise the

operating licenses to reflect approval of Amendment 42 to the Palo

Verde Nuclear Generating Station Physical Security Plan. The amendments

revise the methods used to search materials, packages, and personnel

prior to their entry into the protected area, as described in the

security plan.

Date of issuance: March 4, 1998.

Effective date: March 4, 1998.

Amendment No.: Unit 1-115; Unit 2-108; Unit 3-87.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the operating licenses.

Date of initial notice in Federal Register: October 8, 1997 (62 FR

52580).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: December 17, 1997, as

supplemented by letters dated February 6, 1998 and March 12, 1998.

Brief description of amendment: The proposed change would revise

Technical Specifications Section 5.6.5, ``Core Operating Limits

Report.'' The revisions add reference to an additional approved

methodology for correlating departure from nucleate boiling (DNB)

ratios. The added methodology is the Siemens Power Corporation Topical

Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling

Correlation for High Thermal Performance Fuel.''

Date of issuance: March 16, 1998.

Effective date: March 16, 1998.

Amendment No. 178.

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4309). The February 6 and March 12, 1998 submittals provided clarifying

information that did not affect the initial determination of no

significant hazards considerations. The Commission's related evaluation

of the amendment is contained in a Safety Evaluation dated March 16,

1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

Date of application for amendments: October 27, 1997.

Brief description of amendments: The amendments would change the

Dresden and Quad Cities Technical Specifications (TS) to clarify the

applicability, action and surveillance requirements for the Standby

Liquid Control System (SLCS). The changes would make the current TS

requirements for the SLCS consistent with the Improved Standard

Technical Specifications (ISTS) contained in NUREG-1433, ``Standard

Technical Specifications General Electric Plants, BWR/4.''

Date of issuance: March 6, 1998.

[[Page 14493]]

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 167, 162, and 180, 178.

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 14, 1998 (63 FR

2277).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 6, 1998.

No significant hazards consideration comments received: No.

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.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: October 2, 1996, as supplemented

July 31, 1997.

Brief description of amendment: The amendment revises Figures

3.1.A-1, 3.1.A-2 and 3.1.A-3, Section 3.1.B and its Bases, Figures

3.1.B-1 and 3.1.B-2, and the Bases of Section 4.3 and Figure 4.3-1 of

the Technical Specifications to incorporate the revised Indian Point

Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.

Date of issuance: February 27, 1998.

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

30 days.

Amendment No.: 195.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1996 (61

FR 58901).

The July 31, 1997, letter 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 February 27, 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.

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: December 17, 1997Brief

description of amendments: The amendments revise Section 6.9.1.9 of the

Technical Specifications to reference updated or recently approved

topical reports, which contain methodologies used to calculate cycle-

specific limits contained in the Core Operating Limits Report.

Date of issuance: March 2, 1998.

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

30 days.

Amendment Nos.: Unit 1-163; Unit 2-155.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4310).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 2, 1998.

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

No. 2, Pope County, Arkansas

Date of application for amendment: July 21, 1997, as supplemented

February 18, 1998.

Brief description of amendment: Technical Specification Change

Request concerning Emergency Feedwater Surveillance Testing. This

request is to make several changes to the ANO-2 Technical

Specifications including extension of the emergency feedwater (EFW)

pump surveillance testing frequency, a reduction in the minimum steam

generator pressure required to perform the surveillance testing on the

turbine-driven EFW pump, and a modification to the EFW pump testing

requirements.

Date of issuance: March 12, 1998.

Effective date: March 12, 1998.

Amendment No.: 188.

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

Technical Specifications/license.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43367).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 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: September 23, 1997, as

supplemented by letters dated February 27 and March 4, 1998.

Brief description of amendment: The amendment changes the Reactor

Protective System (RPS) and Engineering Safety Actuation System (ESFAS)

trip set point and allowable values for steam generator low pressure.

The amendment also relocates the RPS and ESFAS response time tables

from the Technical Specifications to the Safety Analysis Report as

described in NRC Generic Letter 93-08, ``Relocation of Technical

Specification Tables of Instrument Response Time Limits,'' dated

December 29, 1993.

Date of issuance: March 12, 1998.

Effective date: March 12, 1998.

Amendment No.: 189.

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

Technical Specifications/license.

Date of initial notice in Federal Register: January 28, 1998, (63

FR 4311).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 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: September 23, 1997, as

supplemented by letters dated February 27 and March 4, 1998.

Brief description of amendment: The amendment reduces the minimum

required reactor coolant system flow rate in TS 3.2.5 until the ANO-2

steam generators are replaced. The reduced reactor coolant system flow

requirement will account for plugging of up to approximately 30 percent

of the tubes in the existing steam generators at ANO-2.

Date of issuance: March 12, 1998.

Effective date: March 12, 1998.

Amendment No.: 190.

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

Technical Specifications/license.

Date of initial notice in Federal Register: January 28, 1998, (63

FR 4312).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 1998.

[[Page 14494]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: December 5, 1997, as

supplemented December 11, 1997, January 9, February 12 and 19, 1998.

Brief description of amendment: To revise the Final Safety Analysis

Report (FSAR) and the Improved Technical Specification Bases to reflect

the modified reactor building fan recirculation system fan cooler

starting logic.

Date of issuance: March 9, 1998.

Effective date: March 9, 1998.

Amendment No.: 165.

Facility Operating License No. DPR-31: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 15, 1998 (63 FR

2423). The supplemental letters dated December 11, 1997, January 9,

February 12 and 19, 1998, 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 March 9, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida.

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: Frederick J. Hebdon.

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

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of application for amendments: May 14, 1997, as supplemented

by letter dated October 9, 1997 (published in Federal Register as May

15, 1997).

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance frequencies

from at least once every 18 months to at least once per refueling

interval (nominally 24 months) including (1) reactor coolant system

total flow rate, (2) instrumentation for radiation monitoring, (3)

instrumentation and controls for remote shutdown, (4) instrumentation

for accident monitoring, and (5) several miscellaneous TS.

Date of issuance: February 27, 1998.

Effective date: February 27, 1998, to be implemented within 90 days

of the date of issuance.

Amendment Nos.: Unit 1-123; Unit 2-121.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 30, 1997 (62 FR

40855).

The October 9, 1997, supplemental letter provided additional

clarifying information and did not change the staff's initial no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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: January 2, 1997, as supplemented

November 13, 1997.

Brief description of amendment: The amendment changes the Technical

Specifications by extending the surveillance interval for the

functional testing of certain Inservice Inspection American Society of

Mechanical Engineers Code Class 1, 2, and 3 pumps and valves from once

a month to once a quarter.

Date of issuance: March 2, 1998.

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

30 days.

Amendment No.: 178.

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

Technical Specifications.

Date of initial notice in Federal Register: March 26, 1997 (62 FR

14468).

The November 13, 1997, submittal contained clarifying information

that did not change the staff's proposed finding of no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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: December 14, 1995, as

supplemented September 26, 1997.

Brief description of amendment: The amendment changes the James A.

FitzPatrick Technical Specifications (TSs) to incorporate the inservice

testing requirements of Section XI of the American Society of

Mechanical Engineers Boiler and Pressure Vessel Code. The amendment

supplements Amendment No. 241, dated December 2, 1997, by issuing seven

TS pages inadvertently omitted from Amendment No. 241.

Date of issuance: February 27, 1998.

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

30 days.

Amendment No.: 242.

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

Technical Specifications.

Date of initial notice in Federal Register: January 22, 1996 (61 FR

1635).

The September 26, 1997, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 27, 1998.

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.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of application for amendments: December 15, 1997.

Brief description of amendments: The amendments revise the

Technical Specifications (TSs) to adopt Option B, of 10 CFR Part 50,

Appendix J, ``Primary Reactor Containment Leakage Testing for Water-

Cooled Power Reactors,'' to implement a performance-based approach for

Type B and C testing. Additionally, the wording in the TSs would be

modified for the previous adoption of Option B on Type A testing and a

section added on the primary

[[Page 14495]]

containment leakage rate testing program.

Date of issuance: February 27, 1998.

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

within 60 days.

Amendment Nos: 207 and 188.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 14, 1998 (63 FR

2281).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 27, 1998.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: December 17, 1997.

Brief description of amendments: The amendments extended the

surveillance interval of the containment spray nozzle air flow test to

ten years from five years.

Date of issuance: March 11, 1998.

Effective date: March 11, 1998.

Amendment Nos.: Unit 1--Amendment No. 94; Unit 2--Amendment No. 81.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4325).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, OES Nuclear, Inc.,

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

Perry Nuclear Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: December 23, 1997.

Brief description of amendment: This amendment revised Technical

Specification 3.8.1, ``A.C. Sources--Operating,'' consistent with the

recommendations in NRC Generic Letter 94-01, ``Removal of Accelerated

Testing and Special Reporting Requirements for Emergency Diesel

Generators.''

Date of issuance: March 12, 1998.

Effective date: March 12, 1998.

Amendment No.: 92.

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4326).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: July 11, 1997, as supplemented

November 21, December 22, 1997, and February 6, 1998.

Brief description of amendment: The amendment revised Technical

Specifications 3.7/4.7 and their associated Bases to incorporate Option

B of Appendix J to 10 CFR 50, and editorial changes to TS Table 4.7.2

Date of Issuance: February 26, 1998.

Effective date: February 26, 1998, with full implementation within

30 days.

Amendment No.: 152.

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: (62 FR 45465). The

November 21, December 22, 1997, and February 6, 1998, letters did not

change the initial proposed no significant hazards determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated February 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: November 20, 1997.

Brief description of amendment: The amendment revised Technical

Specification (TS) 3.10 and its associated Bases to eliminate the use

of battery charger AB for meeting the requirement of the TS.

Date of issuance: March 5, 1998.

Effective Date: This license amendment is effective as of its date

of issuance, to be implemented within 30 days.

Amendment No.: 153

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68319).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 5, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 22, 1997, as supplemented

by letter dated September 18 and October 31, 1997.

Brief description of amendment: The amendment revises the Technical

Specifications to address the new low pressure CO2 suppression system

for the East and West Switchgear Rooms and more clearly describes the

separation of the two rooms.

Date of Issuance: March 6, 1998.

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

within 30 days.

Amendment No.: 154.

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 8, 1997 (62 FR

52590). Information provided by letter dated October 31, 1997, did not

affect the original no significant hazards consideration.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated March 6, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following

[[Page 14496]]

amendments. The Commission has determined for each of these amendments

that the application for the amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective 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

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

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

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By April 24, 1998, the

licensee may file a request for a hearing with respect to issuance of

the amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

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

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

[[Page 14497]]

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. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 the

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

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: February 5, 1998, as

supplemented February 12, March 3 and 5, 1998.

Brief description of amendments: The amendments revised the

surveillance requirements in Technical Specification (TS) 4.6.1.2

(Requirement a). The change to the referenced TS adds a footnote

stating that the requirement for Type A testing will not apply to

certain instrument line penetrations.

Date of issuance: March 10, 1998.

Effective date: Both units, as of the date of issuance.

Amendment Nos.: 173 and 146.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No. On February 5, 1998, the staff issued a Notice of

Enforcement Discretion, which was immediately effective and remained in

effect until this amendment was issued.

The Commission's related evaluation of the amendments, finding of

emergency circumstances, consultation with the State of Pennsylvania,

and final no significant hazards consideration determination are

contained in a Safety Evaluation dated March 10, 1998.

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: John F. Stolz.

Dated at Rockville, Maryland, this 18th day of March 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

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

Nuclear Reactor Regulation.

[FR Doc. 98-7652 Filed 3-24-98; 8:45 am]

BILLING CODE 7590-01-P