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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

notice. Pub. L. 97-

[[Page 17022]]

415 revised section 189 of the Atomic Energy Act of 1954, as amended

(the Act), to require the Commission to publish notice of any

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

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

authority to issue and make immediately effective any amendment to an

operating license upon a determination by the Commission that such

amendment involves no significant hazards consideration,

notwithstanding the pendency before the Commission of a request for a

hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from March 13, 1999, through March 26, 1999. The

last biweekly notice was published on March 24, 1999 (64 FR 14278).

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 23, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

scheduled in the proceeding, but such an amended 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

[[Page 17023]]

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: February 26, 1999.

Description of amendments request: The proposed amendment would

revise Technical Specification (TS) 3.5.3, ``Emergency Core Cooling

System--Operating,'' to extend the completion time for one inoperable

low pressure safety injection subsystem from 72 hours to 7 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed amendment will extend the Completion Time

for one inoperable low pressure safety injection (LPSI) subsystem in

Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems

(ECCE)[S]--Operating, from 72 hours to 7 days. The LPSI subsystem is

part of the ECCS train and part of the shutdown cooling subsystem.

The LPSI components are not accident initiators in any accident

previously evaluated. Therefore, this change does not involve a

significant increase in the probability of an accident previously

evaluated.

The LPSI system is primarily designed to mitigate the

consequences of a large break loss of coolant accident (LOCA). These

proposed changes do not affect any of the assumptions used in the

deterministic LOCA analysis.

In order to evaluate the LPSI Completion Time extension with

respect to the ECCS, 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 an accident

previously evaluated. These analyses are detailed in CE NPSD-995,

Combustion Engineering Owners Group ``Joint Applications Report for

Low Pressure Safety Injection System AOT Extension,'' May 1995, as

supplemented by updated PVNGS data provided in the attachment to

this enclosure.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed amendment will extend the Completion Time for one

inoperable low pressure safety injection (LPSI) subsystem in

Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems

(ECCE)[S]--Operating, from 72 hours to 7 days. The proposed change

does not change 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 change does not involve a significant reduction in

a margin of safety. The proposed amendment will extend the

Completion Time for one inoperable low pressure safety injection

(LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency

Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days.

The proposed change does not affect the limiting conditions for

operation or their bases used in the deterministic analyses to

establish the margin of safety. PSA evaluations were used to

evaluate these changes. These evaluations demonstrate that the

changes will be risk neutral or risk beneficial for PVNGS. These

evaluations are detailed in CE NPSD-995, as supplemented by updated

data provided in the attachment to this enclosure.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

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

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

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

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

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: January 22, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) Sections 3.7.D.1.g, 6.2.2.h and

6.3.1. Specifically, (1) Section 3.7.D.1.g would be revised to correct

an editorial error; (2) Section 6.2.2.h would be revised to change the

senior reactor operator license requirement for the Operations Manager;

and (3) Section 6.3.1 would modify the qualification requirement for

the Operations Manager.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change [to Section 3.7.D.1.g] is administrative in

nature. It involves making an editorial change to provide the

correct functional description of the breakers. This change does not

affect possible initiating events for accidents previously evaluated

or alter the configurations or operation of the facility. The

Limiting Safety Systems Settings and Safety Limits specified in the

current Technical Specifications

[[Page 17024]]

remain unchanged. Therefore, the proposed change to the subject

Technical Specification would not increase the probability or

consequences of an accident previously evaluated.

The proposed change [to Section 6.2.2.h] is administrative in

nature. The individual who provides the day to day direction of the

activities of the operating shift will still possess an SRO [Senior

Reactor Operator] license and this proposed change is consistent

with the statement in NUREG-1431, Section 5.2.2.f. This change does

not affect possible initiating events for accidents previously

evaluated or alter the configuration or operation of the facility.

The Limiting Safety Systems Settings and Safety Limits specified in

the current Technical Specifications remain unchanged. Therefore,

the proposed change to the subject Technical Specification would not

increase the probability or consequences of an accident previously

evaluated.

The proposed change [to Section 6.3.1] is administrative in

nature. The individual who provides the day to day direction of the

activities of the operating shift will still possess an SRO license

and this proposed change is consistent with the statement in NUREG-

1431, Section 5.2.2.f. This change does not affect possible

initiating events for accidents previously evaluated or alter the

configuration or operation of the facility. The Limiting Safety

Systems Settings and Safety Limits specified in the current

Technical Specifications remain unchanged. Therefore, the proposed

change to the subject Technical Specification would not increase the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

As stated above, the proposed change [to Section 3.7.D.1.g] is

administrative in nature. The safety analysis of the facility

remains complete and accurate. There are no physical changes to the

facility and the plant conditions for which the design basis

accidents have been evaluated are still valid. The operating

procedures and emergency procedures are unaffected. Consequently, no

new failure modes are introduced as a result of the proposed change.

Therefore, the proposed change will not initiate any new or

different kind of accident.

The proposed change [to Section 6.2.2.h] is administrative in

nature. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility and the

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedures and emergency

procedures are unaffected. Consequently, no new failure modes are

introduced as a result of the proposed changes. Therefore, the

proposed change will not initiate any new or different kind of

accident.

The proposed change [to Section 6.3.1] is administrative in

nature. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility and the

plant conditions for which the design basis accidents have been

evaluated are still valid. The operating procedures and emergency

procedures are unaffected. Consequently, no new failure modes are

introduced as a result of the proposed changes. Therefore, the

proposed change will not initiate any new or different kind of

accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change [to Section 3.7.D.1.g] is administrative in

nature. Since there are no changes to the operation of the facility

or physical design the Updated Final Safety Analysis Report (UFSAR)

design basis, accident assumptions, or Technical Specification Bases

are not affected. Therefore, the proposed changes will not result in

a reduction in the margin of safety.

The proposed change [to Section 6.2.2.h] is administrative in

nature. Since there are no changes to the operation of the facility

or physical design the Updated Final Safety Analysis Report (UFSAR)

design basis, accident assumptions, or Technical Specification Bases

are not affected. Therefore, the proposed changes will not result in

a reduction in the margin of safety.

The proposed change [to Section 6.3.1] is administrative in

nature. Since there are no changes to the operation of the facility

or physical design the Updated Final Safety Analysis Report (UFSAR)

design basis, accident assumptions, or Technical Specification Bases

are not affected. Therefore, the proposed changes will not result in

a reduction in the margin of safety.

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

this review, it appears that the three standards of 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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Director.

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

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: January 22, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) Section 4.3. Specifically, the

revision would permit the reactor coolant system (RCS) leak test to be

performed at normal operating pressure after it has been closed

following normal opening in lieu of a hydrostatic test being performed

at 2335 psig.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The change proposes a system leakage test for

the RCS that is comparable to the hydrostatic test that it replaces,

as acknowledged by the NRC approval of ASME Code Case N-498,

``Alternative Rules for 10-Year Hydrostatic Pressure Testing for

Class 1 and 2 Systems Section XI, Division 1,'' and the ASME

[American Society for Mechanical Engineers] Boiler and Pressure

Vessel Code, Section XI. [. . .] The proposed change to substitute a

system leak test at normal operating pressure in lieu of the

hydrostatic test at 2335 psig will minimize challenge to plant

safety and demonstrate leak tightness of the RCS. Therefore, the

proposed change would 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.

The proposed license amendment does not create the possibility

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

evaluated. The proposed changes do not involve the addition of any

new or different type of equipment, nor do they involve the

operation of equipment required for safe operation of the facility

in a manner different from those addressed in the Updated Final

Safety Analysis Report. [. . .] Based on industry experience, it is

expected that any leaks would be discovered by the leak test at

normal operating pressure.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed license amendment does not involve a significant

reduction in a margin of safety. The proposed changes do not

adversely affect performance of any safety related system or

component, instrument operation, or safety system setpoints and do

not result in increased severity of any of the accidents considered

in the safety analysis. Although the current basis states that if

the system does not leak at 2335 psig (operating pressure + 100

psig) it will be leak tight during normal operation, industry

experience demonstrates that leaks are not discovered as a result of

hydrostatic test pressure propagating a preexisting flaw through

wall. In most cases, leaks are discovered when the system is at

normal operating pressure. Also, testing will continue to be

performed as required by the ASME Boiler and Pressure Vessel Code

Section XI.

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.

[[Page 17025]]

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Director.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: March 8, 1999.

Description of amendment request: The proposed amendments would

delete certain requirements from Technical Specification (TS) Section

6.0 ``Administrative Controls'' that are adequately controlled by

existing regulations, other than 10 CFR 50.36 and the TS. The

amendments also relocate selected requirements from TS Section 6.0 to

the licensee's controlled documents such as the Turkey Point Units 3

and 4 Updated Final Safety Analysis Report (UFSAR). The amendments also

clarify certain provisions of TS Section 6.0. The proposed changes are

to relocate, revise, delete, or clarify the following provisions of the

TS:

------------------------------------------------------------------------

Existing TS section Subject Proposed change

------------------------------------------------------------------------

6.2.2.f................ Administrative Controls Partly delete, partly

on Working Hours of relocate within TS.

Plant Staff.

Table 6.2-1............ Minimum Shift Crew Clarify.

Composition.

6.2.3.................. Shift Technical Advisor Clarify.

6.4.................... Training............... Delete.

6.5.................... Review and Audit....... Relocate to UFSAR.

6.6.................... Reportable Event Action Partly delete, partly

relocate to UFSAR.

6.8.2.................. Review and Approval of Relocate to UFSAR.

Procedures.

6.8.3.................. Temporary Changes to Relocate to UFSAR.

Procedures.

6.8.4.b................ In-Plant Radiation Relocate to UFSAR.

Monitoring.

6.8.4.g................ Radiological Relocate to UFSAR.

Environmental

Monitoring Program.

6.10................... Record Retention....... Relocate to UFSAR.

6.11................... Radiation Protection Relocate to UFSAR.

Program.

6.12................... High Radiation Area.... Clarify.

6.13................... Process Control Program Relocate to UFSAR.

(PCP).

6.14................... Offsite Dose Revise to reflect

Calculation Manual changes to 6.5 &

(ODCM). 6.10.

------------------------------------------------------------------------

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 plant in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed changes are administrative in nature. These

proposed changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because they do not affect assumptions contained in plant safety

analyses, the physical design and/or operation of the plant, nor do

they affect Technical Specifications that preserve safety analysis

assumptions. None of the proposed changes involve a physical

modification to the plant, a new mode of operation or a change to

the UFSAR transient analyses. No Limiting Condition for Operation,

ACTION statement or Surveillance Requirement is affected by any of

the proposed changes. Also, these proposed changes, in themselves,

do not reduce the level of qualification or training such that

personnel requirements would be decreased. Further, the Proposed

changes do not alter the design, function, or operation of any plant

component. Therefore, the proposed changes do not affect the

probability or consequences of accidents previously evaluated.

2. Operation of the plant in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The changes being proposed are administrative in nature and do not

affect assumptions contained in plant safety analyses, the physical

design and/or modes of plant operation defined in the plant

operating license, or Technical Specifications that preserve safety

analysis assumptions. The proposed changes do not introduce a new

mode of plant operation or surveillance requirement, nor involve a

physical modification to the plant. The proposed changes are

administrative in nature. The changes propose to revise, delete, or

relocate the stated administrative control provisions from the TS to

the UFSAR whereby adequate control of information is maintained.

Furthermore, the proposed changes do not alter the design, function,

or operation of any plant components. Therefore, the proposed

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

accident from any accident previously evaluated.

3. Operation of the plant in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes do not involve a significant reduction in a

margin of safety because they are administrative in nature. The

operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. None of the proposed changes involve a physical

modification to the plant, a new mode of operation or a change to

the UFSAR transient analyses. No Limiting Condition for Operation,

ACTION statement, or Surveillance Requirement is affected.

Additionally, the proposed changes do not alter the scope of

equipment currently required to be OPERABLE or subject to

surveillance testing, nor does the proposed change affect any

instrument setpoints or equipment safety functions. Therefore, the

change does not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

NRC Project Director: Cecil O. Thomas.

[[Page 17026]]

GPU Nuclear, Inc. etal., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey

Date of amendment request: December 23, 1998.

Description of amendment request: The proposed Technical

Specification (TS) change request will change the surveillance

frequency for verifying the operability of motor-operated isolation

valves and condensate makeup valves in the Isolation Condenser TS

4.8.A.1 and Bases page from once per month to once per 3 months.

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 TS change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed surveillance interval change does not alter the

actual surveillance requirements, nor does it alter the limits and

restrictions on plant operations. The reliability of systems and

components relied upon to prevent or mitigate the consequences of

accidents previously evaluated is not degraded by the proposed

change to the surveillance interval. Assurance of system and

equipment availability is maintained. The proposed change does not

alter any system or equipment configuration.

Based on the above, the proposed change does not significantly

increase the probability or consequences of a[n] accident previously

evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed surveillance interval change does not alter the

actual surveillance requirements, nor does it alter the limits and

restrictions on plant operations. Assurance of system and equipment

availability is maintained. The proposed change does not alter any

system or equipment configuration nor does it introduce any new

mechanisms which could contribute to the creation of a new or

different kind of accident than previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed change extends the surveillance interval for

verifying the operability of Isolation Condenser motor-operated

isolation valves and condensate makeup valves from once per month to

once per three months. The proposed change does not alter the actual

surveillance requirements, the limits and restrictions on plant

operations nor the design, function or manner of operation of any

structures, systems or components. System availability and

reliability are maintained. Accordingly, the proposed TS 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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

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

NRC Project Director: Elinor G. Adensam.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey

Date of amendment request: February 12, 1999.

Description of amendment request: The proposed Technical

Specification (TS) change will delete the organizational chart and the

related organizational references from the Appendix B Environmental TS

and revise the appearance and format of the Environmental TS.

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 because deletion of the organization charts and other

organizational references in the [Environmental Technical

Specifications] ETS does not affect plant operation. GPU Nuclear

will continue to inform the NRC of organizational changes through

other required controls.

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

or different type of accident than previously evaluated because the

proposed change is administrative in nature, and no physical

alteration of plant configuration, changes to setpoints or operating

parameters are proposed.

3. The proposed change does not involve a significant reduction

in the margin of safety because it does not alter the design,

function or manner of operation of any structures, systems or

components. Organizational structure or its representation does not

directly impact the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

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

NRC Project Director: Elinor G. Adensam.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa.

Date of amendment request: February 18, 1999.

Description of amendment request: The proposed amendment would

revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' by

deleting the manual initiation function of the high pressure coolant

injection (HPCI) system and reactor core isolation cooling (RCIC)

system isolation. A related condition as well as corresponding

surveillance requirements and bases would also be deleted. Thus, the

change would (1) revise Table 3.3.6.1-1 by removing items 3j. and 4.j.;

(2) revise Note 2 to Surveillances to Licensing Condition for Operation

(LCO) 3.3.6.1 by deleting information regarding items 3 j. and 4.j.;

and (3) revise LCO 3.3.6.1 by removing Condition G and Surveillance

Requirement 3.3.6.1.10.

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:

After reviewing this proposed amendment, we [the licensee] have

concluded:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The Manual Initiation Function for HPCI and

RCIC Isolation is not considered to be an initiator for any accident

previously evaluated in the UFSAR. Therefore, this change does not

involve a significant increase in the probability of any previously

evaluated accidents. The Manual Initiation push button channels

introduce signals into HPCI and RCIC System isolation logics that

are redundant to the automatic protective instrumentation and

provide manual isolation capability only if a system initiation

signal is present. Technical Specification Section 3.3.6.1 Condition

G requires isolation of the System flowpath, which renders the

System inoperable and reduces the availability of the System due to

the failure of a manually initiated isolation, an isolation which is

not assumed in any transient or accident analysis in the UFSAR.

Removal of the Manual Initiation Function

[[Page 17027]]

for HPCI and RCIC from the Primary Containment Isolation

Instrumentation Section of Technical Specifications does not affect

the automatic protective instrumentation and the automatic isolation

capability. Therefore, this change does not significantly increase

the consequences of a previously analyzed accident.

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

new or different kind of accident from any accident previously

evaluated. The proposed change introduces no new mode of plant

operation and does not involve physical modification to the plant.

Therefore, it does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. The proposed amendment will not involve a significant

reduction in a margin of safety. The proposed change deletes the

Manual Initiation Function from Technical Specifications, but no

significant reduction in a margin of safety is involved. Technical

Specification Section 3.3.6.1 Condition G requires isolation of the

System flowpath, which renders the System inoperable and reduces the

availability of the System due to the failure of a manually

initiated isolation, an isolation that is not assumed in any

transient or accident analysis in the UFSAR. Removal of the Manual

Initiation Function for HPCI and RCIC from the Primary Containment

Isolation Instrumentation Section of Technical Specifications does

not affect the automatic protective instrumentation and the

isolation capability. This change is acceptable based on the fact

that the Manual Initiation Function is not assumed in any accident

or transient analysis in the UFSAR.

Based upon the above, we [licensee] have determined that the

proposed amendment will not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, SE., Cedar Rapids, IA 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: T.J. Kim, Acting.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 1, 1999.

Description of amendment request: The proposed amendment would

change the Cooper Nuclear Station (CNS) Technical Specifications (TSs)

to revise the calibration frequency of the reactor recirculation flow

transmitters from once every 184 days to once every 18 months. This

calibration is required as part of TS Surveillance Requirement (SR)

3.3.1.1.10.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Changing the calibration frequency of the recirculation

loop flow transmitters from 184 days to 18 months may increase the

amount of drift experienced by the transmitters. However, CNS

calculation (NEDC 98-024 [forwarded by letter dated March 10, 1999])

takes into account the 18 month calibration intervals. This

calculation, performed in accordance with the General Electric (GE)

setpoint methodology for CNS, demonstrates that the expected drift

is not significant, and is consistent with past operating

experience. Changing the calibration frequency of the flow

transmitters does not change any of the precursors assumed in the

accident analysis. Therefore, changing the calibration frequency for

flow transmitters from 184 days to 18 months does not involve a

significant increase in the probability of an accident previously

evaluated in the USAR [Updated Safety Analysis Report].

The proposed change will not create the possibility of a new or

different kind of accident than evaluated in the USAR. The proposed

change does not result in any physical change to plant structures,

systems, or components. The proposed change does not alter the form,

fit, or function of any equipment or components credited in the

accident analyses described in the USAR. Therefore, changing the

test frequency does not create the possibility of a new or different

kind of accident.

The proposed change will not involve a significant reduction in

a margin of safety. This conclusion is based on the fact that the

proposed change is consistent with the drift assumptions used in CNS

approved calculation (NEDC 98-024). The calibration frequency of 18

months is consistent with the operating practices prior to

conversion to Improved Technical Specifications, and is consistent

with past operating practice at CNS. Therefore, changing the

calibration frequency from 184 days to 18 months 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, Nebraska 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, Nebraska 68602-0499.

NRC Project Director: George Dick, Acting.

Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-336

and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New

London County, Connecticut

Date of amendment request: March 5, 1999.

Description of amendment request: The proposed amendment would

relocate certain Technical Specification (TS) Section 6.0

administrative controls to the NRC-approved Northeast Utilities Quality

Assurance Program (NUQAP) Topical Report. Specifically, Sections 6.2.3

(Unit 3 only), 6.5, 6.6 (partial), 6.7 (partial), and 6.10. The

proposed amendment would also delete parts of Section 6.6 and 6.7

because their requirements are duplicated in existing regulations or

elsewhere in the TS. In addition, the proposed amendment would modify

the table of contents and other TS sections to incorporate the

aforementioned changes (e.g., correct references).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

proposed changes and has concluded that they do not involve a

Significant Hazards Consideration (SHC). The basis for this

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

compromised. The proposed changes are not a SHC because the proposed

change will not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

No design basis accidents are affected by these proposed

changes. The proposed changes relocate portions of the Technical

Specifications to the NUQAP Topical Report or remove duplicate

sections and are being proposed to eliminate the need for a T.S.

change each time there is a related change in the administrative

controls for the site.

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

from any accident previously evaluated.

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

these revisions. The potential for an unanalyzed accident is not

created. There is no impact on plant response, and no new failure

modes are introduced. The proposed deletions and

[[Page 17028]]

editorial changes have no impact on safety limits or design basis

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

event.

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

These changes do not directly affect any protective boundaries

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

These proposed changes relocate portions of the administrative

controls to the NUQAP Topical Report or are editorial in nature.

Therefore, there is no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: Elinor G. Adensam.

PECO Energy Company, Docket No. 50-353, Limerick Generating Station,

Unit 2, Montgomery County, Pennsylvania

Date of amendment request: March 11, 1999.

Description of amendment request: The proposed revision to the

Technical Specifications (TSs) involves a change to TS Section 2.1 and

its associated TS Bases to revise the minimum critical power ratio

(MCPR) Safety Limits for Cycle 6.

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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The revised MCPR Safety Limits for LGS Unit 2 Technical

Specifications, and their use to determine cycle-specific thermal

limits, have been calculated using NRC-approved methods (i. e,

GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific

inputs. The use of these methods assures that the SLMCPR [safety

limit minimum critical power ratio] value is within the existing

design and licensing basis, and cannot increase the probability or

severity of an accident.

The basis for the MCPR Safety Limit calculation is to ensure

that greater than 99.9 percent of all fuel rods in the core avoid

transition boiling if the limit is not violated. The MCPR Safety

Limit preserves the existing margin to transition boiling and fuel

damage in the event of a postulated accident. The probability of

fuel damage is not increased.

Therefore, the proposed TS changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The MCPR Safety Limit is a Technical Specification numerical

value designed to ensure that fuel damage from transition boiling

does not occur as a result of the limiting postulated accident. The

MCPR Safety Limit is not an accident initiator; therefore, it cannot

create the possibility of any new type of accident. The new MCPR

Safety Limits are calculated using NRC-approved methods (i.e.,

GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific

inputs.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in the margin of safety.

The margin of safety as defined in the TS Bases will remain the

same. The new MCPR Safety Limits are calculated using NRC-approved

methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the

current fuel design and licensing criteria, and are based on LGS,

Unit 2, Cycle 6 specific inputs. The MCPR Safety Limit remains high

enough to ensure that greater than 99.9 percent of all fuel rods in

the core will avoid transition boiling if the limit is not violated,

thereby preserving the fuel cladding integrity.

Therefore, the proposed TS changes do not involve a reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: Elinor G. Adensam.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit

1, Luzerne County, Pennsylvania

Date of amendment request: March 12, 1999.

Description of amendment request: The amendment would modify the

Susquehanna Steam Electric Station, Unit 1, Technical Specifications

Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The

change updates the allowable values for both the Core Spray (CS) and

Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome

Pressure--Low'' functions for initiation and injection permissive.

Specifically, the allowable values are being changed from a specified

minimum pressure to a specified allowable pressure band. This more

restrictive allowable value range will prevent CS and LPCI system

overpressurization while still permitting injection to prevent fuel

clad temperature limits from being exceeded.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated. The proposed

amendment changes the ``Reactor Steam Dome Pressure-Low'' Allowable

Values so to provide further assurance that the Core Spray and RHR

systems will perform their LOCA [Loss-of-coolant accident] design

basis function.

The functional design basis of the Core Spray and LPCI is to

inject water into the reactor vessel to cool the core during a LOCA

by opening the Core Spray and LPCI injection valves when reactor

pressure drops below the reactor vessel low pressure permissive. The

upper analytical limit for the permissive is the Core Spray and LPCI

systems' maximum design pressure, and the lower analytical limit is

the lowest pressure which allows injection to prevent exceeding the

fuel cladding temperature limit. The new allowable values were

selected to lie within the upper and lower limits to ensure there

will be no change in the required logic or functions of the Core

Spray and LPCI systems. These new values do not affect the LOCA or

its ``limiting fault'' frequency of occurrence and do not introduce

any new accidents or malfunctions of equipment important to safety.

Since they do not affect the LOCA, they do not change the

probability of occurrence of the LOCA. The new allowable values do

not change the logic or function of the reactor vessel low pressure

permissive. These new values simply provide the basis for which the

associated pressure

[[Page 17029]]

instruments are to be set to ensure proper operation of Core Spray

and LPCI within the design pressures as described above. Therefore,

the change in allowable values does not increase the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety.

Based upon the analysis presented above, PP&L [PP&L, Inc.]

concludes that the proposed action does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The new allowable values do not change any plant systems,

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

any new Core Spray and LPCI logic or functions. The new allowable

values were selected to ensure the required operation of the Core

Spray and LPCI systems within the design pressures described above.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The change does not involve a reduction in the margin of safety.

Technical Specification Bases Section B3.3.5.1 9 (ECCS

Instrumentation) identifies that the low reactor steam dome pressure

signals are used as permissives for operation of the low pressure

ECCS subsystems. The new allowable values were selected so to not

impact the logic, redundancy, operability or surveillance

requirements for these subsystems. The new allowable values maintain

the margin requirements that the Core Spray and LPCI system

pressures such that they do not exceed their system maximum design

pressures and that system pressures are high enough to ensure that

the ECCS injection prevents the fuel peak cladding temperature from

exceeding the limits of 10CFR50.46.

The margin of safety is unaffected by the proposed changes.

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 proposed to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

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

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General

Counsel, PP&L, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101-

1179.

NRC Project Director: Elinor G. Adensam.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit

1, Luzerne County, Pennsylvania

Date of amendment request: March 12, 1999.

Description of amendment request: This proposed amendment would

revise the minimum critical power ratio safety limit in Technical

Specification (TS) Section 2.1.1.2. Also, the proposed amendment would

modify the references in TS Section 5.6.5 in order to include only

those references that directly support the generation of the Core

Operating Limit and to remove the reference for the Lead Use

Assemblies, which will be discharged during the next Unit 1 refueling

outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The applicable sections of the FSAR are Chapters 4.4 and 15.

FSAR Chapter 4.4 describes the MCPR Safety Limit, and Chapter 15

describes the transient and accident analyses. The reference to be

added to Section 5.6.5 of the Unit 1 Technical Specifications

describes a NRC approved critical power correlation for

ATRIUMTM-10 fuel. This correlation is appropriate for use

in conservative methodologies for generating MCPR Safety Limits and

MCPR Operating Limits to assure safe operation of Unit 1 with

ATRIUMTM-10 fuel. A discussion of the impact of the

proposed Technical Specification change is provided below.

The proposed change in critical power correlation does not

physically affect the plant or its systems. Thus, it does not

increase the probability of an accident previously evaluated.

A Unit 1 Cycle 12 MCPR Safety Limit analysis was performed for

PP&L by SPC. This analysis used NRC approved methods described in

ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods

will be used each cycle to calculate the Unit 1 Safety Limits. For

Unit 1 Cycle 12, the critical power performance of the 9 x 9-2 and

ATRIUMTM-10 fuel was determined using the NRC approved

ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR

calculations statistically combine uncertainties on feedwater flow,

feedwater temperature, core flow, core pressure, core power

distribution, and uncertainties in the Critical Power Correlation.

The SPC analysis used cycle specific power distributions and

calculated MCPR values such that at least 99.9% of the fuel rods are

expected to avoid boiling transition during normal operation or

anticipated operational occurrences. The resulting two-loop and

single-loop MCPR Safety Limits are included in the proposed

Technical Specification change. Thus, the cladding integrity and its

ability to contain fission products are not adversely affected.

Analyses of the Single Loop Pump Seizure accident with the NRC

approved ANFB-10 correlation for ATRIUMTM-10 fuel

(Reference 1) will be performed to demonstrate that the NRC

acceptance criterion (i.e., small fraction of 10CFR100 dose limits)

is met. Analyses will also be performed to validate the conclusion

that two-loop transients are more severe than those events analyzed

in single-loop operation.

Changes to Section 2.1.1.2 reflect the change from a flow

dependent MCPR Safety Limit to a single value MCPR Safety Limit for

two-loop operation and single-loop operation.

Changes to Reference 5.6.5 delete the methodology used for

critical power analyses for ATRIUMTM-10 fuel and add the

NRC approved ANFB-10 methodology to the list of approved

methodologies. Other changes in Reference 5.6.5 are administrative

in nature because they delete references not directly related to the

generation of Core Operating Limits. No new analysis approaches are

used due to these changes.

Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion

of the ANFB-10 critical power correlation. The range of the

applicability of the ANFB-10 is valid for pressures > 571 psia and

bundle mass fluxes > 0.115 x 10\6\ lb/hr-ft \2\. These values

assure that a valid CPR calculation will result at or above 25% of

rated core thermal power, that is, reactor steam dome pressure

785 psig and core flow 10 Mlbm/hr.

Changes to BASES Sections 3.2.1, 3.2.2, 3.2.3, and 3.2.4 reflect

the removal of Reference 7 for the ABB LUAs, since the four LUAs

will be discharged from Unit 1 during the Unit 1 11th Refueling and

Inspection Outage.

The consequences of transients and accidents will remain within

the criteria approved by the NRC. The methodology used to perform

the analyses has been previously approved by the NRC. Thus, analysis

results using the new methodology will continue to provide assurance

that the reactor will perform its design safety function during

normal operation and design basis events. Therefore, the proposed

action does not involve an increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes to the Unit 1 Technical Specifications

(MCPR Safety Limits, removal of methodology references not directly

supporting the generation of Core Operating Limits, removal of the

two references describing previously approved methodology for

applying ANFB to ATRIUMTM-10 fuel, removal of the ABB LUA

reference, and inclusion of the ANFB-10 correlation reference) do

not require any physical plant modifications, physically affect any

plant components, or entail changes in plant operation. Removal of

the Unit 1 Cycle 11 footnote allows Unit 1 Cycle 12 and future cycle

operation with NRC

[[Page 17030]]

approved methodology. Thus, the proposed change does not create the

possibility of a previously unevaluated operator error or a new

single failure. The consequences of transients and accidents will

remain within the criteria approved by the NRC. Therefore, the

proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The applicable Technical Specification Sections include 2.1.1.2

and 5.6.5.

The changes to the Unit 1 Technical Specifications discussed in

Item 1 above do not require any physical plant modifications,

physically affect any plant components, or entail changes in plant

operation. Therefore, the proposed change will not jeopardize or

degrade the function or operation of any plant system or component

governed by Technical Specifications. The consequences of transients

and accidents will remain within the criteria approved by the NRC.

The proposed MCPR Safety Limits and use of the ANFB-10 critical

power correlation described in the reference added to Section 5.6.5

do not involve a significant reduction in the margin of safety as

currently defined in the Bases of the applicable Technical

Specification sections.

Therefore, the 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

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

Attorney for licensee: Bryan A. Snapp, Esquire, PP&L, Inc., 2 North

Ninth St., Allentown, PA 18101.

NRC Project Director: Elinor G. Adensam.

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: March 2, 1999.

Description of amendment request: The proposed amendment would

clarify the use of a ``check valve with flow through the valve

secured'' as a means to isolate an affected containment penetration

(i.e., a penetration with an inoperable penetration barrier) in

Technical Specification 3.6.3 Action b.

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 does not involve an increase in the

probability or consequences of an accident previously evaluated. The

proposed change does not involve any hardware changes. The proposed

change will clarify Technical Specification 3.6.3 Action b to allow

the use of a check valve with the flow through the valve secured as

a means to isolate an inoperable containment penetration. This

change is consistent with the changes identified in NUREG-1431,

``Improved Standard Technical Specifications for Westinghouse

Plants'', Specification 3.6.3 (Containment Isolation Valves), which

identifies check valves with flow through the valve secured as a

type of deactivated automatic valve, and with 10 CFR 50 Appendix A

General Design Criteria 55 and 56, which include the use of check

valves as ``automatic isolation valves''. The proposed change will

not affect the containment isolation valve OPERABILITY requirements

or associated isolation time limits established in the

Specifications. Therefore the proposed change will not affect any

safety margin or safety limit applicable to the facility. Therefore

no increase in the probability or consequences of any accident

previously evaluated will occur.

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

kind of accident from any accident previously evaluated?

The proposed change will clarify Technical Specification 3.6.3

Action b to allow the use of a check valve with the flow through the

valve secured as a means to isolate an inoperable containment

penetration. The proposed change will not involve any physical

change to plant systems, structures, or components (SSC). This

change is consistent with the changes identified in NUREG-1431,

``Improved Standard Technical Specifications for Westinghouse

Plants'', Specification 3.6.3 (Containment Isolation Valves), which

identifies check valves with flow through the valve secured as a

type of deactivated automatic valve, and with 10 CFR 50 Appendix A

General Design Criteria 55 and 56, which include the use of check

valves as ``automatic isolation valves''. The proposed change only

provides clarification to the existing Specification 3.6.3, and will

not affect the established containment isolation valve OPERABILITY

requirements or associated isolation time limits. Since the proposed

change does not impact operation of the facility as presently

approved, no possibility exists for a new or different kind of

accident from those previously evaluated.

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

of safety?

The proposed change will clarify Technical Specification 3.6.3

Action b to allow the use of a check valve with the flow through the

valve secured as a means to isolate an inoperable containment

penetration. This change is consistent with the changes identified

in NUREG-1431, ``Improved Standard Technical Specifications for

Westinghouse Plants'', Specification 3.6.3 (Containment Isolation

Valves), which identifies check valves with flow through the valve

secured as a type of deactivated automatic valve, and with 10 CFR 50

Appendix A General Design Criteria 55 and 56, which include the use

of check valves as ``automatic isolation valves''. The proposed

change only provides clarification to the existing Specification

3.6.3, and will not affect the established containment isolation

valve OPERABILITY requirements or associated isolation time limits.

The proposed change does not involve a significant reduction in a

margin of safety because the ability to isolate containment in the

event of a release of radioactive material to the containment

atmosphere or pressurization of the containment will be maintained.

The margin of safety is defined by the established containment

isolation valve OPERABILITY requirements and associated isolation

time limits. The proposed change does not alter these operating

restrictions and the margin of safety which assures the ability to

isolate containment is not affected.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: George Dick, Acting.

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: March 9, 1999.

Description of amendment request: The amendment request proposes

that reference to the Independent Safety Engineering Group be removed

from Technical Specification requirements, with supporting changes to

the Operations Quality Assurance Plan.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed amendment is a programmatic and administrative

change that

[[Page 17031]]

does not physically alter safety-related systems, nor does it affect

the way in which safety-related systems perform their functions. The

functions assigned to the Independent Safety Engineering Group are

addressed by other organizations. Because the design of the facility

and system operating parameters are not being changed, the proposed

amendment does not involve an increase in the probability or

consequences of any accident previously evaluated.

The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed amendment is a programmatic and administrative

change that does not physically alter safety-related systems, nor

does it affect the way in which safety-related systems perform their

functions. The functions assigned to the Independent Safety

Engineering Group are addressed by other organizations. Because the

design of the facility and system operating parameters are not being

changed, the proposed amendment does not create the possibility of a

new or different kind of accident previously evaluated.

The proposed change does not involve a significant reduction in

a margin of safety.

The proposed amendment is a programmatic and administrative

change that provides assurance that plant operations continue to be

conducted in a safe manner. The functions assigned to the

Independent Safety Engineering Group are addressed by other

organizations. As stated above the proposed amendment does not

physically alter safety-related systems, nor does it affect the way

in which safety-related systems perform their functions. Because the

design of the facility and system operating parameters are not being

changed, the proposed amendment 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: George F. Dick, Acting.

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: March 15, 1999 (Supplement to October

29, 1998).

Description of amendment request: The proposed amendments were

submitted by application dated October 29, 1998, to relocate Technical

Specification (TS) 3/4.7.9 requirements for snubbers to the Technical

Requirements Manual. The Commission issued a Notice of Consideration of

Issuance of Amendments regarding its proposed no significant hazards

consideration determination that was published in the Federal Register

on December 16, 1998 (63 FR 69346).

Subsequently, by letter dated March 15, 1999, supplemental

information was submitted to include TS 6.10.3.l to be relocated to the

Technical Requirements Manual. This information is being noticed to

provide for public comment on the issue of no significant hazards

consideration.

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 supplement to the amendment request relocates the record

keeping requirements of Technical Specification 6.10.3.l to the

Technical Requirements Manual. The change does not involve a

physical alteration of the plant (no new or different type of

equipment will be installed) or make changes in the methods

governing normal plant operation. The change will not impose

different requirements, and adequate control of information will be

maintained. This change will not alter assumptions made in the

safety analysis and licensing basis.

The Technical Requirements Manual is incorporated in the South

Texas Project Updated Final Safety Analysis Report and will be

maintained pursuant to 10 CFR 50.59. In addition, snubber

operability is addressed in existing surveillance procedures that

are also controlled by 10 CFR 50.59 and subject to the change

control provisions imposed by plant administrative procedures, which

endorse applicable regulations and standards.

Therefore, the supplement to the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The supplement to the amendment request relocates the record

keeping requirements of Technical Specification 6.10.3.l to the

Technical Requirements Manual. The change does not involve a

physical alteration of the plant (no new or different type of

equipment will be installed) or make changes in the methods

governing normal plant operation. The change will not impose

different requirements, and adequate control of information will be

maintained. This change will not alter assumptions made in the

safety analysis and licensing basis.

The Technical Requirements Manual is incorporated in the South

Texas Project Updated Final Safety Analysis Report and will be

maintained pursuant to 10 CFR 50.59. In addition, snubber

operability is addressed in existing surveillance procedures that

are also controlled by 10CFR50.59 and subject to the change control

provisions imposed by plant administrative procedures, which endorse

applicable regulations and standards.

Therefore, the change does 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 supplement to the amendment request relocates the record

keeping requirements of Technical Specification 6.10.3.l to the

Technical Requirements Manual. The relocated requirements remain the

same as the existing Technical Specifications. The change will not

reduce a margin of safety because it has no impact on any safety

analysis assumptions. Future changes to the relocated requirements

will be evaluated per the requirements of 10CFR50.59.

Therefore, the supplement will not result in 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: George F. Dick, Acting.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: February 15, 1999.

Description of amendment request: The proposed amendment would

revise requirements of Technical Specifications Section 6,

``Administrative Controls,'' related to (1) plant manager's

responsibilities, (2) plant staff titles and organization, (3) offsite

and onsite review committee (4) reportable events, and (5) actions

required in event of a safety limit violation.

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

[[Page 17032]]

consideration, which is presented below:

The proposed amendment will not change the intent of the TS or

decrease WPSC's management support or involvement in activities at

the Kewaunee Plant. Furthermore, it will not result in a decrease in

the engineering or technical support supplied by the plant staff or

the corporate support staff. The proposed changes are administrative

in nature. They primarily involve the relocation of existing

requirements to owner controlled documents; therefore, there are no

significant hazards associated with this change. As an

administrative change this will not result in a significant increase

in the probability of occurrence or consequences of an accident. As

an administrative change this will not create the possibility of a

new or different kind of accident from any previously analyzed. This

administrative change relocates existing requirements, and

therefore, will not involve a significant decrease in the margin of

safety.

In addition, the staff analyzed the proposed changes in accordance

with the provisions of 10 CFR 50.92. The proposed change will not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

The analyses for the previously evaluated accidents are presented

in Chapter 14 of the Updated Safety Analysis Report. There are 19

postulated accidents addressed therein. The proposed amendment would

not affect the safety analysis assumptions or analytical models used

for any of these analyses. Also, the calculated dose consequences for

analyzed accidents would be unaffected. Therefore the proposed changes

do 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 previously evaluated.

The proposed accident does not involve any physical change to the

design of the physicality, or operation of the facility outside the

bounds of the existing analyses. Thus, there is no possibility of

creating a new or different kind of accident.

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

The proposed changes do not involve any physical changes to any of

the fission product barriers or to the design or operation of any

safety systems. Also, no safety limits, limiting safety systems

settings, limiting conditions for operation or testing requirements

would be affected. Therefore, the proposed changes do not involve a

significant reduction in the margin of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P.O. Box 1497, Madison, WI 53701-1497.

NRC Project Director: Cynthia A. Carpenter.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power

Station, Franklin County, Massachusetts

Date of amendment request: March 17, 1999.

Description of amendment request: Licensee submitted a License

Amendment request to delete administrative Technical Specification (TS)

requirements related to overtime restrictions. The licensee stated it

will provide appropriate constraints on excessive overtime in its

Administrative Procedures.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes are administrative in nature and simply

eliminate outdated requirements from the YNPS Technical

Specifications. As such the changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The administrative

nature of the changes will not affect safety-related systems or

components or their mode of operation and therefore, will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

any previously evaluated. The proposed changes do not modify any

plant systems or components and, therefore, do not create the

possibility of a new or different accident from any previously

evaluated.

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

changes are administrative in nature involving the deletion of

outdated requirements in the technical specifications; therefore,

there will be no reduction in the margin of safety.

Based on the considerations noted above, it is concluded that the

proposed changes will not endanger the public health and 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: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: Seymour H. Weiss.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power

Station, Franklin County, Massachusetts

Date of amendment request: March 17, 1999.

Description of amendment request: Licensee submitted a License

Amendment request to transfer Technical Specification Sections 6.7--

Procedures and Programs and 6.9--Record Retention to the Yankee

Decommissioning Quality Assurance Program (YDQAP).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes are administrative in nature.

Administrative requirements in Sections 6.7 and 6.9 of the YNPS

Technical Specifications are to be transferred to the YDQAP which is

the current location of related administrative requirements. As such

the changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The administrative

nature of the changes will not affect safety-related systems or

components or their mode of operation and therefore, will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

any previously evaluated. The proposed changes do not modify any

plant systems or components and, therefore, will not create the

possibility of a new or different accident from any previously

evaluated.

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

changes are administrative in nature involving the relocation of

administrative requirements from one licensing document to another

licensing document currently containing related requirements;

therefore, there will be no significant reduction in the margin of

safety.

Based on the considerations noted above, it is concluded that the

proposed changes will not endanger the public health and safety.

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

this

[[Page 17033]]

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: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: Seymour H. Weiss.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power

Station, Franklin County, Massachusetts

Date of amendment request: March 17, 1999.

Description of amendment request: Licensee submitted a License

Amendment request to consolidate management positions and to transfer

Technical Specification review and audit functions to the Yankee

Decommissioning Quality Assurance Program (YDQAP).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes are administrative in nature and reflect a

streamlining of the YAEC/YNPS management structure and procedures

consistent with the on-going requirement to complete the remaining

scope of YNPS decommissioning safely and efficiently. As such the

changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The administrative

nature of the changes will not affect safety-related systems or

components or their mode of operation and therefore, will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

any previously evaluated. The proposed changes do not modify any

plant systems or components and, therefore, will not create the

possibility of a new or different accident from any previously

evaluated.

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

Elimination of the Manager of Operations position and the Plant

Superintendent position will not eliminate any of the

responsibilities or functions currently assigned to these positions.

These responsibilities or functions will be reassigned to an

appropriately qualified YAEC/YNPS manager, i.e., the Decommissioning

Manager. This change and replacement of the PORC and the NSARC

review and audit functions with an independent safety review and an

IRAC are consistent with the significant reduction in the scope and

the complexity of activities at YNPS as the facility moves into the

later stages of the decommissioning effort; therefore, there will be

no significant reduction in the margin of safety.

Based on the considerations noted above, it is concluded that the

proposed changes will not endanger the public health and 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: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: Seymour H. Weiss.

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

No. 2, Pope County, Arkansas

Date of application for amendment: February 24, 1999.

Brief description of amendment: The amendment would revise

Technical Specification Table 3.3-1, ``Reactor Protective

Instrumentation,'' Action 2, for Arkansas Nuclear One, Unit No. 2. The

proposed change would add a footnote to Action 2 that would allow

startup and operation with the functional units associated with the

Channel ``D'' ex-core nuclear instrumentation to be maintained in the

bypassed or tripped condition following the restart from Refueling

Outage 2R13. This footnote is intended to support normal plant

operations until such time that the Channel ``D'' ex-core detector

assembly can be restored to an operable status.

Date of publication of individual notice in Federal Register: March

8, 1999 (64 FR 11067).

Expiration date of individual notice: April 7, 1999.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

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.

[[Page 17034]]

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz

Mill, Pennsylvania

Date of application for amendment: September 28, 1998 supplemented

on November 17, 1998.

Brief description of amendment: This amendment changes the license

to reflect the new legal name of the licensee for the Westinghouse Test

Reactor to CBS Corporation.

Date of issuance: March 25, 1999.

Effective Date: March 25, 1999.

Amendment No: 9.

Facility License No. TR-2: This amendment changes the license.

Date of initial notice in Federal Register: December 16, 1998, (63

FR 69334).

The Commission has issued a Safety Evaluation for this amendment

dated March 25, 1999.

No significant hazards consideration comments received: No.

Local Public Document: N/A.

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

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

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

1 and 2, Rock Island County, Illinois

Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of application for amendments: December 17, 1998.

Brief description of amendments: The amendments revised the

respective facility Technical Specifications (TS) by adding a new

Limiting Condition for Operations that provided an administrative

enhancement by allowing testing required to return equipment to service

to be conducted under administrative controls.

Date of issuance: March 16, 1999.

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 172, 167; 184, 181; 132, 117.

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

NPF-11 and NPF-18.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4153) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 16, 1999.

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; for LaSalle, the Jacobs Memorial Library, 815 North

Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: August 14, 1998, as

supplemented on October 13 and December 23, 1998.

Brief description of amendments: The amendments revised the

Technical Specifications (TSs) to reflect the use of Siemens Power

Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments

incorporate the following into the TSs: (1) new methodologies that will

enhance operational flexibility and reduce the likelihood of future

plant derates; (2) administrative changes that adopt Improved Standard

Technical Specification (iSTS) language where appropriate; and (3)

changes to the Minimum Critical Power Ratio.

Date of issuance: March 16, 1999.

Effective date: Immediately, to be implemented prior to startup of

Cycle 9 for Unit 1 and prior to startup of Cycle 8 for Unit 2.

Amendment Nos.: 131, 116.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the TSs.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59588). The December 23, 1998, submittal provided additional clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

March 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Commonwealth Edison Company, 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: August 14, 1998, as

supplemented by letters dated October 13, 1998, and December 23, 1998.

Brief description of amendments: The amendments changed the Quad

Cities Technical Specifications (TS) to reflect the use of Siemens

Power Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments

incorporate the following into the TS: (a) new methodologies that will

enhance operational flexibility and reduce the likelihood of future

plant derates; (b) administrative changes that eliminate the cycle-

specific implementation of ATRIUM-9B fuel and adopt Improved Standard

Technical Specification language where appropriate; and (c) changes to

the Minimum Critical Power Ratio (MCPR).

The amendment for Unit 1 also reflects the removal of Unit 1

specific pages incorporated into Unit 1 TS by Amendment No. 182 and are

no longer applicable. The August 14, 1998, application superseded an

August 29, 1997, application in its entirety (63 FR 2274).

Date of issuance: March 17, 1999.

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 185 & 182.

Facility Operating License Nos. DPR-29, DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48258) and November 4, 1998 (63 FR 59588). The October 13, 1998,

submittal changed a reference to a recently NRC-approved additive

constant uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF-

1125 (P)(A), supplement 1, Appendix E) from Appendix D which provided

an interim value for ACU. This change was noticed on November 4, 1998

(63 FR 48258). The December 23, 1998, submittal provided additional

clarifying information that did not change the initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated March 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Commonwealth Edison Company, 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: November 30, 1998.

Brief description of amendments: The amendments changed the

technical specifications (TSs) by decreasing the Allowed Outage Time

(AOT) from 67 days to 14 days for the Safe Shutdown Makeup Pump (SSMP).

Date of issuance: March 26, 1999.

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 186 & 183.

[[Page 17035]]

Facility Operating License Nos. DPR-29 and DPR-30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 13, 1999 (64 FR

2246).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina.

Date of application of amendments: September 30, 1998.

Brief description of amendments: The amendments increase the

maximum fuel rod internal pressure in the spent fuel pool from 1200

pounds per square inch gauge (psig) to 1300 psig by changing the

Updated Final Analysis Report (UFSAR) reference to the computer code

used to determine the fuel rod internal pressure (TACO3 computer code

would be added) in UFSAR Chapter 15. In addition, the amendments

justify not increasing the overall effective decontamination factor for

iodine as a consequence of a fuel handling accident and change the

terminology used in the UFSAR from ``fuel assembly gap gas pressure''

to ``fuel rod internal pressure.''

Date of Issuance: March 26, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1-301; Unit 2-301; Unit 3-301.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:

Amendments authorized change(s) to the FSAR.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59590).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

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

No. 2, Pope County, Arkansas.

Date of application for amendment: February 25, 1999.

Brief description of amendment: This amendment revises Technical

Specification (TS) Table 3.3-1, ``Reactor Protective Instrumentation,''

Action 2, for Arkansas Nuclear One, Unit No. 2 (ANO-2). This change

adds a footnote to Action 2 that allows startup and operation with the

functional units associated with the Channel ``D'' ex-core nuclear

instrumentation to be maintained in the bypassed or tripped condition

following the restart from Refueling Outage 2R13. This footnote is

intended to support normal plant operations until such time that the

Channel ``D'' ex-core detector assembly can be restored to an operable

status. This footnote will be in effect for a time period not to extend

beyond Mid-Cycle Outage 2P99, which is the next planned entry into cold

shutdown conditions for ANO-2. A Notice of Enforcement Discretion

(NOED) related to TS Table 3.3-1, Action 2, was issued verbally on

February 23, 1999. The NOED is documented in a letter dated February

25, 1999.

Date of issuance: March 23, 1999.

Effective date: As of the date of issuance.

Amendment No.: 202.

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

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration (NSHC): Yes (64 FR 11067 dated March 8, 1999). The notice

provided an opportunity to submit comments on the Commission's proposed

NSHC determination. No comments have been received. The notice also

provided for an opportunity to request a hearing by April 7, 1999, but

indicated that if the Commission makes a final NSHC determination, any

such hearing would take place after issuance of the amendment.

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

exigent circumstances, and final NSHC determination are contained in a

Safety Evaluation dated March 23, 1999.

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

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

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: November 2, 1995, and as

supplemented by submittal dated January 7, 1999.

Brief description of amendment: This amendment revises technical

specification requirements for handling irradiated fuel in the Primary

Containment and the Fuel Handling Building, and selected specifications

associated with performing core alterations.

Date of issuance: March 11, 1999.

Effective date: March 11, 1999.

Amendment No.: 102.

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

Technical Specifications.

Date of initial notice in Federal Register: December 6, 1995 (60 FR

62497).

The supplemental information contained clarifying information and

did not change the initial no significant hazards consideration

determination and did not change the scope of the original application.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 11, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: August 27, 1996, as supplemented

by submittals dated April 9, 1997, July 22, 1998, December 3, 1998, and

January 18, 1999.

Brief description of amendment: This amendment revised Technical

Specification 3.6.1.3, ``Primary Containment Isolation Valves

(PCIVs),'' and 3.6.1.9, ``Main Steam Isolation Valve (MSIV) Leakage

Control System (LCS).'' The amendment reflects implementation of the

revised accident source term in NUREG-1465, ``Accident Source Terms for

Light-Water Nuclear Power Plants'' and permits the licensee to

eliminate the MSIV LCS and increase the allowable leak rates of the

MSIVs.

Date of issuance: March 26, 1999:

Effective date: March 26, 1999.

Amendment No.: 103.

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

Technical Specifications.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53958).

The supplemental information contained clarifying information and

did not change the initial no significant hazards consideration

determination and did not expand the scope of the original application.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

[[Page 17036]]

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: October 27, 1998.

Brief description of amendment: This amendment revised the minimum

critical power ratio (MCPR) safety limit contained in TS 2.1.1.2. In

addition, the amendment removes a note to TS 2.1.1.2 and a footnote to

TS 5.6.5.b that references MCPR safety limit values as cycle specific.

Date of issuance: March 26, 1999:

Effective date: March 26, 1999.

Amendment No.: 104.

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

Technical Specifications.

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

66603).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio.

Date of application for amendment: September 9, 1998, as

supplemented by submittals dated January 6, March 4, and March 18,

1999.

Brief description of amendment: This amendment revises the design

and licensing basis of containment isolation valves in the feedwater

system. The amendment revises (1) Surveillance Requirement 3.6.1.3.11

of Technical Specification (TS) 3.6.1.3, ``Primary Containment

Isolation Valves (PCIVs)'' to exclude the feedwater check valves from

the hydrostatic test program, (2) TS 5.5.2, ``Primary Coolant Sources

Outside Containment,'' to stipulate that water leakage past the

feedwater motor-operated containment isolation valves and the reactor

water cleanup system return to feedwater line is added to the program,

and (3) TS 5.5.12, ``Primary Containment Leakage Rate Testing

Program,'' to state that the feedwater check valves will be tested in

accordance with the Inservice Testing Program (TS 5.5.6).

Date of issuance: March 26, 1999.

Effective date: March 26, 1999.

Amendment No.: 105.

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

Technical Specifications.

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

56262).

The supplemental information contained clarifying information and

did not change the initial no significant hazards consideration

determination and did not expand the scope of the original Federal

Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

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

Date of application for amendments: August 24, 1998.

Brief description of amendments: These amendments change the St.

Lucie Technical Specifications (TSs) by both removing obsolete license

conditions and revising the TSs. The amendments change the TSs to

modify the St. Lucie Unit 1 TSs to add components, not previously

described in the TSs, to the list of components that comprise an

operable control room emergency ventilation system, to modify the Unit

1 and Unit 2 TSs surveillance requirements to clarify component

operations, not previously described, that must be verified in response

to a containment sump recirculation actuation signal, to delete from

the facility operating license No. NPF-16 for Unit 2, license condition

2.C.19 to reflect the completion of the Unit 1 spent fuel pool re-rack

and delete license condition 2.I to reflect the resolution of

litigation and to modify license condition 2.B.5 to restore the

original syntax of the license condition and license condition 2.F to

update the references to current license conditions.

Date of Issuance: March 17, 1999.

Effective Date: These amendments shall be implemented within 30

days of receipt.

Amendment Nos.: 160 and 99.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the TSs.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50937).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

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

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey.

Date of application for amendment: September 3, 1998.

Brief description of amendment: The amendment revises Technical

Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a Condensate

Storage Tank water level of greater than 35 feet.

Date of Issuance: March 17, 1999.

Effective date: March 17, 1999, to be implemented within 30 days

Amendment No.: 204.

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6698).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated March 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois.

Date of application for amendment: January 20, 1999, as

supplemented February 4, 8, and 25, and March 5, 1999.

Brief description of amendment: The amendment changes the

undervoltage relay setpoints.

Date of issuance: March 26, 1999.

Effective date: March 26, 1999.

Amendment No.: 122.

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1999 (64 FR

4474).

The four supplemental submittals provided additional information

and did not change the requested amendment or affect the proposed no

significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

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

[[Page 17037]]

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of application for amendment: April 13, 1998, as supplemented

November 5, 1998.

Brief description of amendment: The proposed amendment would revise

the Appendix A Technical Specifications to base the Limiting Condition

for Operation for the fuel storage pool water level on a revised

analysis of the fuel handling accident and a new analysis for

radiological shielding during movement of irradiated fuel.

Date of issuance: March 16, 1999.

Effective date: March 16, 1999 (and shall be implemented no later

than 30 days).

Amendment No.: 162.

Facility Operating License No. DPR-36: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 20, 1998 (63 FR

27763). The November 5, 1998, submittal provided additional clarifying

information and did not change the initial proposed no significant

hazards determination and did not expand the scope of the original

application.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: December 30, 1998.

Brief description of amendment: The amendment changes Technical

Specification (TS) Tables 3.6.14-2 and 4.6.14-2 regarding the noble gas

activity monitor channel operability requirement and daily sensor check

surveillance requirement to be consistent with the conditions specified

in TS 3.1.3.a for operability of the emergency cooling system. Also,

this amendment corrects a clerical error in TS 4.6.15.d.

Date of issuance: March 16, 1999.

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

30 days.

Amendment No.: 165.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6699).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of application for amendment: November 19, 1998.

Brief description of amendment: This amendment changes surveillance

frequencies in Technical Specifications 4.8.4.4a and 4.8.4.5a to

require testing of the Electrical Protection Assemblies once every 6

months with the plant on-line rather than shut down.

Date of issuance: March 18, 1999.

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

30 days.

Amendment No.: 86.

Facility Operating License No. NPF-69: Amendment revises the

Technical Specifications

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71970).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 18, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

Date of amendment request: May 20, 1998, as supplemented by letter

dated January 28, 1999.

Description of amendment request: Revise Technical Specifications

Table 3.3-4 and associated bases to depict a change to the refueling

water storage tank low-low level setpoint

Date of issuance: March 12, 1999.

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

within 60 days.

Amendment No.: 60.

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43205).

The supplemental letter provided clarifying information and did not

change the staff's proposed no significant hazards determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

Date of application for amendment: September 9, 1998, as

supplemented February 19 and 26, 1999.

Brief description of amendment: The amendment resolves several

previously identified technical specifications (TSs) compliance issues.

Specifically, the amendment: (1) changed TS definitions 1.24, ``Core

Operating Limits Report,'' 1.27, ``Engineering Safety Feature Response

Time,'' and 1.31, ``Radiological Effluent Monitoring and Offsite Dose

Calculation Manual (REMODCM)''; (2) changed TS 3.0.2, ``Limiting

Condition for Operation,'' by adding a new TS 3.0.6 to the Limiting

Condition for Operation TS section; (3) changed TS 4.0.5,

``Surveillance Requirements''; (4) changed the mode applicability of TS

3.2.3, ``Total Unrodded Integrated Radial Peaking--FrT'';

(5) changed TS 3.3.2.1, ``Engineered Safety Features Actuation System

Instrumentation,'' by modifying TS Table 4.3-2 Table Notation (1) which

it references; and (6) changed TS 3.4.1.1, ``Reactor Coolant System--

Reactor Coolant System Vents.'' The associated TS Bases sections were

also changed.

Date of issuance: March 11, 1999.

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

60 days from the date of issuance.

Amendment No.: 230.

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

Technical Specifications.

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

56251).

The supplemental letters provided clarifying information that did

not change the original proposed no significant hazards consideration

determination or expand the scope of the original Federal Register

notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 11, 1999.

[[Page 17038]]

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Date of application for amendment: July 17, 1998, as supplemented

November 10, 1998, and February 11, 1999.

Brief description of amendment: The amendment revises certain

diesel generator (DG) action statements and surveillance requirements

to improve overall DG reliability and availability.

Date of issuance: March 12, 1999.

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

60 days from the date of issuance.

Amendment No.: 231.

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

Technical Specifications.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43207).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: November 25, 1997, as

supplemented September 25 and November 11, 1998, and January 28, 1999.

Brief description of amendment: The amendment revises the Technical

Specifications for the condensate storage tank (CST) low level suction

transfer setpoint for the high pressure coolant injection (HPCI) and

reactor core isolation cooling (RCIC) systems to allow removing one CST

from service for maintenance.

Date of issuance: March 19, 1999.

Effective date: March 19, 1999, with full implementation within 30

days.

Amendment No.: 105.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 18, 1998 (63

FR 69344) The November 25, 1997, letter and September 25 and November

11, 1998, supplements were referenced in the original Federal Register

notice. The January 28, 1999, supplement provided an updated Technical

Specification page following the incorporation of Amendment 103, issued

December 23, 1998. This information was within the scope of the

original Federal Register notice and did not change the staff's initial

proposed no significant hazards considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: November 25, 1998.

Brief description of amendments: The amendments revise Technical

Specifications 3.2 and Table 3.5-2B to allow limited inoperability of

boric acid storage tank level channels and transfer logic channels to

provide for required testing and maintenance of the associated

components.

Date of issuance: March 17, 1999.

Effective date: March 17, 1999, with full implementation within 30

days

Amendment Nos.: 143 and 134.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 16, 1998 (63

FR 69345).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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: July 30, 1997, as supplemented

by letter dated December 23, 1998.

Brief description of amendments: The amendments revise the combined

Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP)

Unit Nos. 1 and 2 by adding a Limiting Condition for Operation, trip

setpoints, and surveillance requirements for a residual heat removal

pump trip on refueling water storage tank level-low.

Date of issuance: March 26, 1999.

Effective date: March 26, 1999, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 1--130; Unit 2--128.

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

revised the Technical Specifications.

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

FR 68312).

The December 31, 1997 supplemental letter provided additional

clarifying information, did not expand the scope of the application as

originally noted, and did not change the staff's proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

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.

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: August 26, 1997, as

supplemented by letters dated October 14 and November 13, 1997, and

January 29, 1998.

Brief description of amendments: The amendments approve a

modification to the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2

auxiliary saltwater (ASW) system to bypass approximately 800 feet of

Unit 1 and 200 feet of Unit 2 Class 1 ASW pipe, a portion of which is

buried below sea level in the tidal zone outside the intake structure.

Date of issuance: March 26, 1999.

Effective date: March 26, 1999, and shall be implemented in the

next

[[Page 17039]]

periodic update to the FSAR Update in accordance with 10 CFR 50.71(e).

Amendment Nos.: Unit 1--131; Unit 2--129.

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

revised the Final Safety Analysis Report Update.

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

FR 48677).

The October 14 and November 13, 1997, and January 29, 1998,

supplemental letters provided additional clarifying information, did

not expand the scope of the application as originally noticed, and did

not change the staff's original proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 26, 1999.

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.

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: September 25, 1996, as

supplemented on October 29, 1997, March 16, 1998, and February 8, 1999.

Brief description of amendments: The amendments revise the

Technical Specifications by revising the voltage and frequency

acceptance criteria and the start-timing methodology for the emergency

diesel generator surveillance testing.

Date of issuance: March 23, 1999.

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment Nos: 218 and 200.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

5039).

The October 29, 1997, March 16, 1998, and February 9, 1999, letters

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 23, 1999.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

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: September 17, 1998.

Brief description of amendments: The amendments revise Technical

Specification 3/4.8.2, ``Electrical Power Sources--Shutdown,'' for the

AC distribution system and the 125-volt and 28-volt DC distribution

systems. Specifically, the amendments change the Applicability and

Action Statements, if less than the complement of equipment and buses

are operable, to eliminate the need to establish containment integrity

and to add the action to suspend core alterations, positive reactivity

additions, and movement of irradiated fuel assemblies.

Date of issuance: March 24, 1999.

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

within 60 days.

Amendment Nos.: 219 and 201.

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

revised the Technical Specifications.

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

56257).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 24, 1999.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

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

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

Date of application for amendments: April 6, 1995, as supplemented

on August 21, 1995. (TS 95-19).

Brief description of amendments: The amendments change the licenses

for Sequoyah Nuclear Plant, Units 1 and 2 by removing the license

conditions that reference the post-accident sampling system (PASS). The

PASS information has been placed in the Sequoyah Final Safety Analysis

Report (FSAR). This Change is consistent with NUREG-1431, ``Standard

Technical Specifications--Westinghouse Plants.''

Date of issuance: March 16, 1999.

Effective date: March 16, 1999.

Amendment Nos.: 243 and 233.

Facility Operating License Nos. DPR-77 and DPR-79: The amendments

revise the licenses.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20527). The August 21, 1995, letter provided clarifying information

that 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 16, 1999.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: April 23, 1998, as supplemented

on January 25, 1999.

Brief description of amendment: The amendment changes the existing

requirements for the Residual Heat Removal Service Water (RHRSW),

Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS)

as identified in Technical Specifications (TSs) 4.5.C and 3/4.5.D.

Specifically, the changes are as follows:

(1) Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised

to delete the existing allowance for 7 days of operation after both SSW

subsystems are made or found to be inoperable.

(2) Specification 4.5.C.1 and Specification 4.5.D.1: These

requirements have been revised to relocate testing information related

to pump flow and pressure testing characteristics for the RHRSW and SSW

Systems, respectively, to the Technical Requirements Manual.

(3) Specifications 3.5.D.1, 3.5.D.2, 3.5.D.3, 4.5.D.2, 4.5.D.3, and

associated Bases: All references to SSW ``subsystem'' have been

replaced by ``essential equipment cooling loop'' to more accurately

reflect the Vermont Yankee design and operation. In addition, certain

operability clarifications have been made to the Bases relative to

affected Specifications.

(4) Bases for Specification 3.5.D: The Bases have been revised to

omit statements that imply that the ACS could provide adequate heat

removal following a postulated accident. Other Bases additions have

been made that include certain operability clarifications relative to

affected Specifications.

Date of Issuance: March 11, 1999.

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

within 30 days.

Amendment No.: 169.

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

Technical Specifications.

[[Page 17040]]

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6713).

The January 25, 1999, supplement did not affect the original

proposed no significant hazards consideration.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated March 11, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301.

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

Generating Station, Coffey County, Kansas

Date of amendment request: March 20, 1998, as supplemented by

letters dated May 28, June 30, August 28, September 4, November 20, and

December 8, 1998.

Brief description of amendment: The amendment revised the technical

specifications (TS) to support a modification to the plant to increase

the storage capacity of the spent fuel pool and increase the nominal

fuel enrichment to 5% weight percent of U-235. The amendment also

revised the TS to allow the storage of an additional 279 assemblies in

the cask loading pit.

Date of issuance: March 22, 1999.

Effective date: March 22, 1999, to be fully implemented no later

than December 31, 1999, except that the racks in the cask loading pit

may be installed at a future time after the completion of the next

refueling outage.

Amendment No.: 120.

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 13, 1998 (63 FR

37601). The June 30, August 28, September 4, November 20, and December

8, 1998, supplemental letters provided additional clarifying

information, did not expand the scope of the application as originally

noticed, and did not change the staff's proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 22, 1999.

No significant hazards consideration comments received: No.

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.

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

Generating Station, Coffey County, Kansas

Date of amendment request: February 4, 1998, as supplemented by

letter dated October 20, 1998.

Brief description of amendment: The amendment revises the

requirements in Technical Specification Tables 3.3-3, 3.3-4 and 4.3-2

regarding the engineered safety features actuation system (ESFAS)

Functional Unit 6.f, and adds a note to Table 4.3-2 to clarify the

verification of time delays associated with ESFAS Functional Units 8.a

and 8.b.

Date of issuance: March 23, 1999.

Effective date: March 23, 1999, to be implemented within 30 days

from the date of issuance.

Amendment No.: 121.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 25, 1998 (63 FR

14491). The October 20, 1998, supplemental letter provided additional

clarifying information, did not expand the scope of the application as

originally noticed and did not change the staff's original proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated March 23, 1999.

No significant hazards consideration comments received: No.

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.

Dated at Rockville, Maryland, this 31st day of March 1999.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Acting Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-8503 Filed 4-6-99; 8:45 am]

BILLING CODE 7590-01-P