Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from August 3, 1996, through August 16, 1996. The

last biweekly notice was published on August 14, 1996 (61 FR 42274).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

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

publication date and page number of this Federal Register notice.

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

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

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

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

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

petitions for leave to intervene is discussed below.

By September 27, 1996, 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

[[Page 44354]]

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

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

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

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

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

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

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

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

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

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

an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

operator should be given Datagram Identification Number N1023 and the

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

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

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

North Carolina

Date of amendment request: July 19, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 3/4.6.2, Containment Spray

System, to extend the surveillance interval for performance of an air

or smoke flow test through containment spray nozzles from once per 5

years to once per 10 years. This change is consistent with the guidance

in NRC Generic Letter 93-05, ``Line Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operations,'' and NUREG-1366, ``Improvements To Technical

Specifications Surveillance Requirements.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed extended testing frequency of containment spray

nozzles will not affect any initiators of any previously evaluated

accidents or change the manner of operation for any system or

component. The containment spray system serves a mitigating function

by removing heat and fission products from a post accident

containment atmosphere. Increasing the surveillance test interval

will not affect the system's ability to provide this function.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

[[Page 44355]]

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

new or different kind of accident from any accident previously

evaluated.

Since the proposed change affects only a surveillance frequency,

it will not involve any physical alterations to plant equipment or

alter the manner in which any safety-related system performs its

function. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed change does not affect any Final Safety Analysis

Report (FSAR) Chapter 15 accident analyses or impact the margin of

safety for the containment spray system as defined in the Bases to

the Technical Specifications. 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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

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 amendment request: June 10, 1996

Description of amendment request: To change the technical

specifications to reflect the transition from General Electric Company

(GE) to Siemens Power Corporation (SPC) as the fuel supplier for the

Quad Cities Nuclear Power Station, Units 1 and 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits will be established consistent with NRC

approved methods to ensure that fuel performance during normal,

transient, and accident conditions is acceptable. The proposed

Technical Specifications amendment reflects previously approved SPC

methodology used to analyze normal operations, including anticipated

operational occurrences (AOOs), and to determine the potential

consequences of accidents.

Licensing Methods and Models

The proposed amendment is to support operation with NRC approved

fuel and licensing methods supplied from Siemens Power Corporation.

In accordance with FSAR Chapter 15, the same accidents and

transients will be analyzed with the new fuel and methods as were

analyzed by GE for GE fuel. The analysis methods and models are NRC

approved. These approved methods and models are used to determine

the fuel thermal limits (e.g., LHGR, APLHGR, MCPR). The SPC core

monitoring code enables the site to monitor keff as well as rod

density to perform the reactivity anomaly surveillance. This is

consistent with GE methodology. The support systems for minimizing

the consequences of transients and accidents are not affected by the

proposed amendment. Therefore, the change in licensing analysis

methods and models does not significantly increase the probability

of an accident or the consequences of an accident previously

identified.

New Fuel Design

The use of ATRIUM 9B fuel at Quad Cities does not involve a

significant increase in the probability or consequences of any

accident previously evaluated in the FSAR. The ATRIUM-9B fuel is

generically approved for use as a reload BWR fuel type (Reference:

ANF-89-014(P)(A) Rev. 1 Supplement 1, General Mechanical Design for

Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting

postulated occurrences and normal operation have been analyzed using

NRC-approved methods for the ATRIUM 9B fuel design to ensure that

safety limits are protected and that acceptable transient and

accident performance is maintained.

The reload fuel has no adverse impact on the performance of in-

core neutron flux instrumentation or CRD response. The ATRIUM-9B

fuel design will not adversely affect performance of neutron

instrumentation nor will it adversely affect the movement of control

blades relative to the GE fuel. The exterior dimensions of the

ATRIUM-9B fuel have been evaluated by ComEd; the SPC fuel provides

adequate clearances relative to the GE10 fuel installed at Quad

Cities. Thus, no increased interactions with the adjacent control

blade and nuclear instrumentation are created. Additionally, given

the above mentioned overall envelope similarities, no problems are

anticipated with other station equipment such as the fuel storage

racks, the new fuel inspection stand and the spent fuel pool fuel

preparation machine. Therefore, the probability of adverse

interactions between the Siemens fuel and components in the core and

fuel handling equipment is not significantly increased.

The ATRIUM 9B design is neutronically compatible with the

existing fuel types and core components in the Quad Cities core. SPC

tests have demonstrated that the ATRIUM-9B fuel design is

hydraulically compatible with the GE9/GE10 fuel. The bundle pressure

drop characteristics of the ATRIUM 9B bundle are similar to those of

the GE9/GE10 fuel design, hence core thermal-hydraulic stability

characteristics are not adversely affected by the ATRIUM 9B design.

Cycle stability calculations are performed by SPC. Therefore, the

probability of thermal hydraulic instability is not significantly

increased.

An evaluation of the Emergency Procedures is being performed to

ensure that the use of the ATRIUM-9B fuel at Quad Cities does not

alter any assumptions previously made in evaluating the radiological

consequences of an accident at Quad Cities Station. Therefore, the

radiological consequences of accidents are not significantly

increased.

Methods approved by the NRC are being used in the evaluation of

fuel performance during normal and abnormal operating conditions.

The ComEd and SPC methods to be used for the cycle specific

transient analyses have been previously NRC approved. The proposed

methodologies are administrative in nature and do not significantly

affect any accident precursors or accident results; as such, the

proposed incorporation of the SPC methodologies for Quad Cities does

not significantly increase the probability or consequences of any

previously evaluated accidents. The description of the fuel is

modified to include the water box design of the NRC approved ATRIUM-

9B fuel. This change is administrative.

Review of the above concludes that the probability of occurrence

and the consequences of an accident previously evaluated in the

safety analysis report have not been significantly increased.

* * * * *

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

from any accident previously evaluated:

Creation of the possibility of a new or different kind of

accident would require the creation of one or more new precursors of

that accident. New accident precursors may be created by

modifications of the plant configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification amendment reflects

previously approved SPC methodology used to analyze normal

operations, including AOOs, and to determine the potential

consequences of accidents. In accordance with FSAR Chapter 15, the

same accidents and transients will be analyzed with the new fuel and

methods as were analyzed by GE for GE fuel. As stated above, the

proposed changes do not permit modes of operation which differ from

those currently permitted; therefore, the possibility of a new or

different kind of accident is not created. Plant support equipment

is not affected by the proposed changes; therefore, no new failure

modes are created.

New Fuel Design

The basic design concept of a 9x9 fuel pin array with an

internal water box has been used in various lead assembly programs

and in reload quantities in Europe since 1986.

[[Page 44356]]

WNP-2 has loaded reload quantities since 1991. Approximately 650

water box assemblies have been irradiated in the United States

through 1995, with a substantially higher number being irradiated

overseas. The NRC has reviewed and approved the ATRIUM-9B fuel

design (Reference: ANF-89-014(P)(A) Rev. 1 Supplement 1, Generic

Mechanical Design for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR

Reload Fuel). The similarities in fuel design and operation between

GE and SPC, and the previous Boiling Water Reactor experience with

both vendors' fuel indicate there would be no new or different types

of accidents for Quad Cities than have been considered for the

existing fuel. Therefore, the use of ATRIUM-9B fuel at Quad Cities

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

accident from any accident previously evaluated.

* * * * *

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

the following reasons:

The existing margin to safety is provided by the existing

acceptance criteria (e.g., 10CFR50.46 limits). The proposed

Technical Specification amendment reflects previously approved SPC

methodology used to demonstrate that the existing acceptance

criteria are satisfied. The revised methodology has been previously

reviewed and approved by the USNRC for application to reload cores

of GE BWRs. References for the Licensing Topical Reports which

document this methodology, and include the Safety Evaluation Reports

prepared by the USNRC, are added to the Reference section of the

Technical Specifications as part of this amendment.

Licensing Methods and Models

The proposed amendment does not involve changes to the existing

operability criteria. NRC approved methods and established limits

(implemented in the COLR) ensure acceptable margin is maintained.

The ComEd and SPC reload methodologies for the ATRIUM-9B reload

design are consistent with the Technical Specification Bases. The

Limiting Conditions for Operation are taken into consideration while

performing the cycle specific and generic reload safety analyses.

NRC approved methods are listed in Section 6 of the Technical

Specifications.

Analyses performed with NRC-approved methodology have

demonstrated that fuel design and licensing criteria will be met

during normal and abnormal operating conditions. The same margins of

safety are utilized by SPC as GE (e.g., limits on peak cladding

temperature, cladding oxidation, plastic strain). Therefore, there

is not a significant reduction in the margin of safety.

New Fuel Design

The exterior dimensions of the ATRIUM-9B fuel assembly result in

equivalent clearances relative to the GE10B. Thus, no increased

interactions with the adjacent control blade and nuclear

instrumentation are created. The change does not adversely impact

equipment important to safety; therefore,the margin of safety is not

significantly reduced.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee

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

Date of amendment request: August 12, 1996

Description of amendment request: The proposed change would

implement the performance-based containment leak rate testing

provisions of Option B to 10 CFR Part 50 Appendix J for the Type A

(containment) testing program.

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 following analysis is presented, pursuant to 10 CFR 50.91,

to demonstrate that the proposed change will not create a

Significant Hazard Consideration.

1. The proposed change will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Containment leak rate testing is not an initiator of any

accident; the proposed change does not affect reactor operations or

accident analysis, and has no significant radiological consequences.

Therefore, this proposed change will not involve an increase in the

probability or consequences of any previously-evaluated accident.

2. The proposed change will not create the possibility of any

new accident not previously evaluated.

The proposed change does not affect normal plant operations or

configuration, or change any design basis. The proposed changes will

not affect the response of [the] containment during a design basis

accident.

3. There is no significant reduction in a margin of safety.

The proposed changes are based on NRC-accepted provisions, and

maintain necessary levels of reliability of containment integrity.

The performance-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity; this supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible.

Based on the above, no significant hazards consideration is

created by the proposed change.

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

thisreview, 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: May 31, 1996

Description of amendment request: The proposed amendment revises

the surveillance test interval for the reactor protection system

reactor trip breakers, reactor trip modules, and electronic trip relays

from 1 month to 6 months. In addition to requesting a change to the

Arkansas Nuclear One, Unit 1 Technical Specifications, the request also

proposes the same changes to NUREG-1430, Standard Technical

Specifications - Babcock and Wilcox Plants.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The accident mitigation features of the plant are not affected

by the proposed test interval extension. The results of the B&W

Owners Group Topical Report BAW-10167, Supplement 3, ``Justification

for increasing The Reactor Trip System On-Line Test Intervals,''

show that the test interval extension of the reactor protection

system trip devices is not a significant contributor to trip system

unavailability or the risk of core damage.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2. Does Not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The reactor trip device surveillance test interval is not, in

and of itself, considered to be an accident initiator. Failure of a

trip device to function is an analyzed condition and does not

constitute a new or different kind of accident.

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

or different kind of accident from any previously evaluated.

Criterion 3. Does Not Involve a Significant Reduction in the

Margin of Safety.

The results of the B&W Owners Group Topical Report BAW-10167,

Supplement 3, ``Justification for Increasing The Reactor Trip

[[Page 44357]]

System On-Line Test Intervals,'' show that the test interval

extension of the reactor protection system trip devices is not a

significant contributor to trip system unavailability or the risk of

core damage. In addition, the uncertainty analysis contained in BAW-

10167 confirms the robustness of the results by demonstrating that

even with an order of magnitude change in the failure data, the

incremental increase due to an increased test interval is

insignificant. Entergy Operations has reviewed BAW-10167 and found

it applicable to ANO-1.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: William D. Beckner

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

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

Date of amendment request: May 9, 1996

Description of amendment request: The proposed amendment changes

the name of Arkansas Power and Light Company (AP&L) to Entergy

Arkansas, Inc. in both the Operating License and the Technical

Specifications. AP&L is licensed to own and possess Arkansas Nuclear

One (ANO). The company licensed to operate ANO, Entergy Operations,

Inc. is unaffected by this change.

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 Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

The proposed change documents changing the legal name of the

company. The proposed change will not affect any other obligations.

The company will continue to own all of the same assets, will

continue to serve the same customers, and will continue to honor all

existing obligations and commitments. Therefore, this change does

not involve a significant increase in the probability or

consequences of any accident previously evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

The administrative changes in the operating license requirements

do not involve any change in the design of the plant. Therefore,

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

kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The proposed change is administrative in nature and does not

reduce the margin of safety imposed by any current requirements.

Therefore, this change does not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: William D. Beckner

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

Turkey Point Plant Units 3 and 4, Dade County, Florida

Dates of amendment request: July 17, 1996

Description of amendment request: The licensee proposed to change

the Turkey Point Units 3 and 4 Technical Specifications (TS) to

implement 10 CFR 50, Appendix J, Option B, for containment leakage

testing. Changes include relocating the details for containment testing

to the ``containment leakage rate testing program'' and adding the

requirements of the containment leakage rate testing program to TS

6.8.4, which describes facility programs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

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

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:

a) These proposed changes are all consistent with NRC

requirements and guidance for implementation of 10 CFR 50, Appendix

J, Option B.

b) Based on industry and NRC evaluations performed in support of

developing Option B, these changes potentially result in a minor

increase in the consequences of an accident previously evaluated due

to the expanded testing intervals. However, the proposed changes do

not result in an increase in the core damage frequency since the

containment system is used for mitigation purposes only.

c) These changes are expected to result in increased attention

to components with poor leakage test history as part of the

performance-based nature of Option B, such that the marginally

increased consequences from the expanded testing intervals may be

further reduced or negated.

Therefore, these changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the implementation of a performance-based program for containment

leakage rate testing, since the proposed changes do not involve the

addition or modification of equipment, nor do they alter the design

or operation of affected plant systems, structures, or components.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are basically unchanged by the

proposed amendments due to the following reasons:

a) The acceptance criteria for total integrated containment

leakage of 1.0 La is consistent with the current technical

specifications and is within the design basis accident assumptions,

and therefore does not reduce the margin of safety.

b) The increase in intervals between leak-test surveillances

will not significantly reduce the margin of safety as shown by

findings in NUREG 1493, ``Performance-Based Containment Leak-Test

Program'', which was based on implementation of the performance-

based testing of Option B.

Therefore these changes do not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

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

NRC Project Director: Frederick J. Hebdon

[[Page 44358]]

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of amendment request: May 21, 1996

Description of amendment request: The proposed change to the

condensate storage tank (CST) level indication would ensure that the

water level is sufficient to provide 50,000 gallons of water for core

spray makeup to the reactor pressure vessel.

Technical Specification (TS) Surveillance Requirement (SR)

3.5.2.2.b for ECCS - Shutdown states: ``Condensate storage tank (CST)

water level is [greater than or equal to] 12 feet.'' The corresponding

Bases state: ''... the CST contains [greater than or equal to] 150,000

gallons of water, equivalent to 12 feet, ensures that the CS System can

supply at least 50,000 gallons of makeup water to the RPV.''

Subsequent licensee analyses confirmed that Plant Hatch Units 1 and

2 CST configurations are different; that is, for both CSTs, a water

level of 12 feet is not equivalent to the required capacity of 150,000

gallons of water. Based on these calculations, the correct level for

the Unit 1 CST is 13 feet, and the correct level for the Unit 2 CST is

15 feet.

The proposed change would revise Unit 1 and Unit 2 SR 3.5.2.2.b to

require a CST water level of greater than or equal to 13 feet and

greater than or equal to 15 feet, respectively, to ensure at least

50,000 gallons of water are available for core spray (CS) makeup to the

reactor pressure vessel (RPV).

The associated Bases for each unit will be revised accordingly.

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, because this administrative change to the CST

water level does not alter the operation of any plant system or

component. The proposed change does not involve a physical

modification to any structure, system, or component. The minimum CST

water level for each unit is being increased to account for the

height of the CS suction standpipe within each CST and the

differences in the Unit 1 and

Unit 2 CST diameters (gallons/ft of water) as follows:

a. Unit 1 - The proposed minimum water level is calculated as:

CS suction standpipe height of 9 ft + (50,000 gallons divided by

12,704 gallons/ft) = 12.93 ft or 13 ft.

b. Unit 2 - The proposed minimum water level is calculated as:

CS suction standpipe height of 10 ft + (50,000 gallons divided by

11,343 gallons/ft) = 14.4 ft or 15 ft.

The revised minimum levels ensure at least 50,000 gallons of

water are provided above the top of the standpipe in each unit's CST

and are available for CS makeup to the RPV, as stated in the

applicable Bases. The TS Limiting Conditions for Operation (LCO)

remain unaffected by the proposed change.

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

new or different kind of accident from any accident previously

evaluated. Revising Surveillance Requirement acceptance criteria

does not result in any physical modification to the plant or

operation of any existing equipment.

3. The proposed TS change does not involve a significant

reduction in a margin of safety, since this administrative change

only ensures the existing TS Bases are satisfied by increasing the

minimum CST water level requirement to ensure at least 50,000

gallons of water are available for CS injection to the RPV. The

proposed change does not involve a physical modification to any

structure, system or component, and does not modify the operation of

any existing equipment.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

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

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

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

Matagorda County, Texas

Date of amendment request: July 8, 1996

Description of amendment request: The proposed amendment would

clarify that the component cooling water system surge tank level

instrumentation can be demonstrated operable by performing a channel

calibration test during any plant mode of operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to Technical Specification Surveillance

Requirement 4.7.3.b.3 will not effect any accident initiators or

precursors and will not alter the design assumptions for the systems

or components used to mitigate the consequences of an accident.

Calibration is performed on level instrumentation of Component

Cooling Water System trains that are out of service for scheduled

maintenance. Isolation redundancy is provided by instrumentation

associated with the trains that are in service during the

calibration. Since the surveillance will continue to be performed at

the specified interval, this proposed change will not increase the

probability of occurrence of an accident previously evaluated. The

surveillance does not differ from those previously performed;

therefore, there is no impact on the 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.

Clarifying the surveillance interval for surge tank level

instrumentation does not involve installation or operation of new or

different kinds of equipment. There is no change in the procedures

as described in the Technical Specifications. The change only

clarifies the interval at which the subject calibration will be

performed. 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 specified surveillance will remain as stated in the

Technical Specifications. Consequently, there is no reduction in the

effectiveness of the surveillance in ensuring equipment operability.

Calibration is performed on level instrumentation of Component

Cooling Water System trains that are out of service for scheduled

maintenance. Isolation redundancy is provided by instrumentation

associated with the trains that are in service during the

calibration. Consequently, clarifying the interval at which the

calibration is performed will have no significant impact on 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, TX 77488

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

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

NRC Project Director: William D. Beckner

[[Page 44359]]

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

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

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

Matagorda County, Texas

Date of amendment request: August 8, 1996

Description of amendment request: The proposed amendment would

allow the transition from Mode 4 to Mode 3 with the turbine-driven

auxiliary feedwater pump inoperable and allow a 72-hour period after

the entry into Mode 3 to complete all necessary operability testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will allow entry into Mode 3 with an

inoperable Turbine Driven Auxiliary Feedwater pump. Since the

operability test on the Turbine Driven Auxiliary Feedwater pump can

only be performed once steam pressure is greater than or equal to

1000 psig, this change will allow the plant to reach the Mode where

steam pressure greater than or equal to 1000 psig is available to

perform the operability testing on the Turbine Driven Auxiliary

Feedwater pump. The allowance of 72 hours to complete the

surveillance testing will make the surveillance requirements

consistent with the allowed outage time already established in the

Action Statements. The proposed change does not affect the

probability of an accident. The Turbine Driven Auxiliary Feedwater

pump is not assumed to be an initiator of any analyzed event. The

consequences of an accident previously evaluated remain unchanged by

allowing the pump to be inoperable until suitable conditions exist

to perform the operability testing. The operability testing will

continue to demonstrate that the Turbine Driven Auxiliary Feedwater

pump will perform as required prior to entry into Mode 2. This

change will not alter assumptions relative to the mitigation of an

accident or transient event. Therefore, this change will 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 previously evaluated.

This change will not physically alter the plant (no new or

different type of equipment will be installed). The changes in

methods governing normal plant operation are consistent with current

safety analysis assumptions. The proposed change will allow entry

into Mode 3 with the Turbine Driven Auxiliary Feedwater pump

inoperable in order to perform the pump Operability Test on the

turbine driven AFW [Auxiliary Feedwater] pump once steam pressure is

greater than or equal to 1000 psig. Therefore, this 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 proposed change will allow entry into Mode 3 with the

Turbine Driven AFW pump inoperable in order to perform the pump

Operability Test on the turbine driven AFW pump once steam pressure

is greater than or equal to 1000 psig. This will allow time for the

plant to obtain suitable test conditions with steam pressure greater

than or equal to 1000 psig. The margin of safety is not affected by

this change. The operability testing will continue to maintain

assurance that the AFW Pumps will perform as required prior to entry

into Mode 2. The safety analysis assumptions will still be

maintained, thus, no question of safety exists. Therefore, this

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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: William D. Beckner

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

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

Date of amendment request: June 4, 1996

Description of amendment request: The proposed amendment would

modify the Seabrook Station, Unit No. 1 Technical Specifications to

implement Option B to 10 CFR Part 50, Appendix J by referring to

Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test

Program. The following Technical Specifications would be affected by

the proposed amendment:

1. Definitions: Definition 1.7, Containment Integrity (Item d.)

would be revised to reflect that leakage rates would be in accordance

with the Containment Leakage Rate Testing Program.

2. Limiting Conditions for Operation and Surveillance Requirements:

a. Containment Integrity: Surveillance Requirement 4.6.1.1.c would

be deleted because the specific guidance would be contained in the

Containment Leakage Rate Testing Program.

b. Containment Leakage: Limiting Condition for Operation 3.6.1.2.a

through 3.6.1.2.c and Surveillance Requirements 4.6.1.2.a through

4.6.1.2.h would be revised to replace specific guidance with a

reference to the Containment Leakage Rate Testing Program.

c. Containment Leakage: The Action for Limiting Condition for

Operation 3.6.1.2 would be revised to include the equivalent Action as

required for Limiting Condition for Operation 3.6.1.1 when the overall

integrated containment leak rate exceeds 1.0 La.

d. Containment Air Locks: Limiting Conditions for Operation

3.6.1.3.a and 3.6.1.3.b would be deleted and Surveillance Requirements

4.6.1.3.a and 4.6.1.3.b would be revised to replace specific guidance

with a reference to the Containment Leakage Rate Testing Program. The

footnote addressing the exemption to Appendix J regarding testing the

air locks prior to establishing containment integrity would be

maintained in the Containment Leakage Rate Testing Program.

e. Containment Vessel Structural Integrity: Surveillance

Requirement 4.6.1.6 would be revised to replace specific guidance with

a reference to the Containment Leakage Rate Testing Program.

f. Containment Ventilation System: Limiting Condition for Operation

3.6.1.7, Action b. would be revised to replace specific guidance with a

reference to the Containment Leakage Rate Testing Program. Surveillance

Requirement 4.6.1.7.1 would be revised to replace specific guidance

with a reference to the Containment Leakage Rate Testing Program.

g. Containment Enclosure Building: Limiting Condition for Operation

3.6.5.3 and Surveillance Requirement 4.6.5.3 would be revised to

include a reference to the requirements in the Containment Leakage Rate

Testing Program.

3. Bases: Sections 3/4.6.1.2, Containment Leakage; 3/4.6.1.7,

Containment Ventilation System; and 3/4.6.5.3, Containment Enclosure

Building Structural Integrity, would be revised to reflect the above

changes including a reference to the Containment Leakage Rate Testing

Program. In addition, a statement would be added to Section 3/4.6.1.2

to clarify the operability of containment regarding allowable leakage

rates.

4. Administrative Controls: Section 6.15 would be added to

establish a Containment Leakage Rate Testing Program, as specified in

Regulatory Guide 1.163, dated September 1995.

Basis for proposed no significant hazards consideration

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

[[Page 44360]]

licensee has provided its analysis of the issue of no significant

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

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

review is presented below.

A. The changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated (10

CFR 50.92(c)(1)) because the proposed changes merely revise the

testing criteria for containment penetrations. The revised criteria

will be based on the guidance in Regulatory Guide 1.163,

``Performance-Based Containment Leak-Test Program.''

This guidance allows for the use of relaxed testing frequencies

for containment penetrations that have performed satisfactorily on a

historical basis.

To support consideration of Option B to Appendix J, the NRC

staff reviewed the potential impact of performance-based testing

frequencies for containment penetrations. The NRC staff review is

documented in NUREG-1493 ``Performance-Based Containment Leak-Test

Program.'' One of the staff's conclusions was that reducing the

frequency of Type A tests (Integrated Leak Rate Tests) from three

per 10 years to one per 10 years leads to a marginal increase in

risk. For Type B and C testing (Local Leak Rate Tests), the change

in testing frequency will not have significant impact since, under

existing requirements, leakage contributes less than 0.1 percent of

the overall accident risk. The use of a performance-based testing

program will continue to provide assurance that the accident

analysis assumptions remain bounding.

B. The changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated

(10 CFR 50.92(c)(2)) because they do not affect the manner by which

the facility is operated or involve changes to structures, systems,

or components that affect the operational characteristics of the

facility. The changes merely revise the testing criteria for the

containment penetrations, and establish a Containment Leakage Rate

Testing Program to ensure that the performance history of each

penetration is satisfactory prior to changing any test frequency.

Since there is no change to the facility or the way in which the

facility is operated, there is no possibility of creating a new or

different kind of accident than previously analyzed.

C. The changes do not involve a significant reduction in a

margin of safety (10 CFR 50.92(c)(3)). During the development of 10

CFR Part 50, Appendix J, Option B, the NRC staff determined the

reduction in safety associated with the implementation of the

performance-based testing program. The staff concluded that reducing

the frequency of Type A tests (Integrated Leak Rate Tests) from

three per 10 years to one per 10 years would have an imperceptible

impact upon risk. For Type B and C testing (Local Leak Rate Tests),

the change in testing frequency will not have significant impact

since this leakage contributes less than 0.1 percent of the overall

risk based on the existing regulations. The use of Option B will

have minimal impact on the radiological release rates since most

penetration leakage is well below the specified limits. The staff

noted that the accident risk is relatively insensitive to

containment leakage rate because accident risk is dominated by

accident sequences that result in failure of or bypass of the

containment. The use of a performance-based testing program will

continue to provide assurance that the accident analysis assumptions

remain bounding.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270

NRC Project Director: Phillip F. McKee

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 17, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to relocate the operability

requirements for shock suppressors (snubbers) from the TS to the

Updated Safety Analysis Report (USAR) and incorporate snubber

examination and testing requirements into TS 3.3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will relocate operability requirements for

shock suppressors (snubbers) from the Technical Specifications (TS)

to the Updated Safety Analysis Report (USAR) and/or plant

procedures. On July 16, 1993, the NRC issued a Final Policy

Statement on Technical Specification Improvements for Nuclear Power

Reactors. The Final Policy Statement contains four criteria which

can be used to determine which constraints on the design and

operation of nuclear power plants are appropriate for inclusion in

TS. The NRC has incorporated these criteria into 10 CFR 50.36,

``Technical specifications.'' Snubbers do not meet any of the four

criteria for inclusion as a Limiting Condition for Operations within

the TS, and therefore it is proposed that these requirements be

relocated from the TS. The proposed change would not reduce or

revise any of the current requirements for snubber operability, only

relocate the requirements. Any changes to the requirements contained

in the USAR and/or plant procedures can be made without NRC approval

only when the changes meet the criteria of 10 CFR 50.59. Changes to

the snubber operability requirements that do not meet the criteria

of 10 CFR 50.59 must be approved by the NRC by license amendment.

Therefore, the relocation of the requirements on snubber operability

from the TS to the USAR does not increase the probability or

consequences of any accident previously analyzed.

The proposed change also deletes sections of the TS which are

redundant or in conflict with the American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel Code. Snubbers are

required to be examined and tested in accordance with ASME Section

XI by 10 CFR 50.55a. The proposed change will ensure that the TS

implement ASME Section XI examination and testing requirements for

snubbers in accordance with 10 CFR 50.55a. Where differences between

the deleted sections of the TS and ASME Section XI requirements

exist, the Section XI requirements are similar or more conservative

than the TS. For example, although the functional test sample size

differs between the methodologies, both ensure that a very high

percentage of the snubbers in the plant are operable within

acceptance limits. Therefore, the proposed revision does not reduce

the effectiveness of snubber examination and testing.

The proposed change would not reduce the operability

requirements, acceptance criteria, or examination and testing of

snubbers. Therefore, the proposed change 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.

There will be no physical alterations to the plant

configuration, changes to setpoint values, or changes to the

implementation of setpoints or limits as a result of this proposed

change.

The proposed change deletes duplicate or conflicting

requirements between the TS and the ASME Section XI. In these areas,

the proposed deletions would remove the TS requirements and testing

would be conducted in accordance with ASME Section XI as directed by

10 CFR 50.55a. Although the requirements of ASME Section XI differ

from the TS in some cases, the differences do not decrease the

effectiveness of testing and examination as compared to the TS

requirements. Other areas, such as snubber operability requirements

and service life monitoring, which are presently addressed by TS,

but are not covered under ASME Section XI, will be maintained in the

USAR so that these requirements cannot be deleted without NRC

approval.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not reduce the operability,

examination, or testing requirements for snubbers. Snubbers will

still be required to meet the requirements of ASME Section XI and 10

CFR 50.55a except where specific written relief has been granted by

the NRC. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

[[Page 44361]]

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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 20, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to clarify surveillance test

requirements of TS 3.1, Tables 3-1, 3-2, 3-3, 3-3A, and 3-5.

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 changes to the Table of Contents are administrative in

nature to reflect the removal of incore instrumentation

(Specification 2.10.3) from the TS by Amendment 167 and for

consistency. Amendment 169 inadvertently reinserted incore

instrumentation back into the Table of Contents.

The change to Specification 2.1.7(1)b is necessary because the

requirement to test the signal to alarm meter relay located in

Specification 3.1, Table 3-3, Item 6 is being deleted. The test,

which verifies the high and low pressurizer level alarm settings and

the pressurizer heater cutout function is unnecessary. Operating

experience has shown that a shiftly pressurizer level verification

as proposed for Specification 3.1, Table 3-3, Item 6.a is sufficient

to detect any level deviation and verify that operation is within

safety analyses assumptions. The level alarms serve as early warning

devices but do not provide an accident mitigation function.

Replacing the monthly test with a channel check is in accordance

with NUREG-1432, Combustion Engineering (CE), Standard Technical

Specifications (STS), Surveillance Requirement (SR) 3.3.11.1 (post

accident monitoring instrumentation). The monthly channel check

supplements the shiftly level verification.

The Basis of Specification 3.1 is revised to clarify

expectations regarding a channel check of channels that are normally

off scale when the surveillance is required. In this situation, the

channel check only verifies that they are off scale in the same

direction. Off scale low current loop channels are verified to be

reading at the bottom of the range and not failed downscale. These

statements are taken from the Bases of CE STS SR 3.3.4.1 Engineered

Safety Features Actuation System (ESFAS) Instrumentation (Analog).

In addition, the Basis of Specification 3.1 is revised to

clarify that power operated relief valve (PORV) actuation is not

required during the channel functional test of the PORV low

temperature setpoint (Table 3-3, Item 18.a). PORV actuation is not

required because it could depressurize the reactor coolant system.

This clarification is modeled after a similar statement from the

Bases of SR 3.4.12.6 (Low Temperature Overpressure Protection (LTOP)

System) of the CE STS.

Changing Specification 3.1, Tables 3-1, 3-2, 3-3, and 3-3A by

using defined terms to enable the Surveillance Method to match the

Surveillance Function is an administrative change designed to

simplify the tables. Removal of the extraneous text does not alter

the surveillance because the defined terms are equivalent in meaning

to the deleted text.

The reordering of several items in the tables into a Check-Test-

Calibrate sequence adds consistency to the tables. Text revisions in

the Channel Description or Surveillance Function columns of Tables

3-1 and 3-2 add clarity and/or consistency. Footnote No. 1 in Table

3-1 concerning the bistable trip tester was deleted because it is

unnecessary.

The Surveillance Function of Table 3-1, Item 1.c (Power Range

Safety Channels) is being changed to ``Test'' from ``Calibrate and

Test.'' It is not necessary for Item 1.c to require both because

Item 1.b already requires the power range safety channel adjustment

(calibration) to be performed daily. As stated in the Basis of

Specification 3.1, ``The minimum calibration frequencies of once-

per-day for the power range safety channels, ...are considered

adequate.'' To further clarify the issue, the Basis of Specification

3.1 is being revised to note that the daily calibration is a heat

balance adjustment only.

Changing Table 3-1, Item 4 (Thermal Margin/Low Pressure (TM/LP))

to use the defined term CHANNEL CALIBRATION will allow OPPD to relax

the current TM/LP calibration requirements with a negligible impact

on safety. Calibration of the temperature input and pressure input

will still require calibration to known standards (i.e., resistance

and pressure), but will allow the calibrations to be done separately

instead of coincidently. The channel functional test that follows

the channel calibration verifies proper function of the TM/LP

circuitry.

Removing the word ``Instruments'' from the Channel Description

of Table 3-2, Item 14 makes the Channel Description consistent with

the Surveillance Method. Table 3-2, Item 14 is not intended to

verify safety injection tank (SIT) instrumentation operability but

rather that the parameters level and pressure are within limits.

Generic Letter (GL) 93-05, Item 7.4, states that the operability of

SIT instrumentation is not directly related to the capability of a

SIT to perform its safety function. GL 93-05 concludes that the

surveillance should only confirm that the parameters defining SIT

operability are within their specified limits.

Items 22 & 24 are being added to Table 3-2 to clearly state the

requirement for testing manual actuation of the Engineered Safety

Features (ESF) channels for Off-site Power Low Signal (OPLS) and

Auxiliary Feedwater. Although testing manual actuation of these

channels is done via the existing Specifications, the requirement to

do so is not clearly stated. Reordering Table 3-2, Item 23 into a

Check-Test-Calibrate Surveillance Frequency sequence adds clarity

and consistency.

The addition of Footnote No. 7 to Table 3-2 clarifies that the

refueling frequency ESF channel functional test pertains to the

backup channels such as derived circuits and equipment that cannot

be tested when the plant is at power. Operating certain relays

during power operation could cause plant transients or equipment

damage.

The revisions to Table 3-3, Item 6, clarify that pressurizer

level is the parameter to be verified and not the pressurizer level

instruments. The revision to Item 6.a is consistent with CE STS SR

3.4.9.1 (pressurizer water level). Reordering Item 6 into a Check-

Test-Calibrate Surveillance Function sequence makes Item 6

consistent with the ordering of the other items in Table 3-3. The

requirement to test the signal to alarm meter relay currently

located in Specification 3.1, Table 3-3, Item 6.c is unnecessary.

Operating experience has shown that a shiftly pressurizer level

verification as proposed for Specification 3.1, Table 3-3, Item 6.a

is sufficient to detect any level deviation and verify that

operation is within safety analyses assumptions. Thus, the monthly

``Test'' requirement will be replaced with a ``Check'' to supplement

the less formal but more frequent shiftly level verification of Item

6.a.

Table 3-3, Items 21 (PORV Operation & Acoustic Position

Indication Channel) and 23 (Safety Valve Acoustic Position

Indication Channel) should be revised to a channel functional test

from a channel/circuit check. An oscillator and installed impactors

are used to generate noise signals and therefore, this surveillance

is more accurately described as a channel functional test rather

than a channel check.

Table 3-3, Items 21 and 22 (PORV Block Valve Operation &

Position Indication) should have the requirement to verify operation

on the emergency power supply deleted. Permanent Class 1E power

supplies the PORV and PORV Block Valve. Therefore, verification of

PORV or PORV Block Valve operability while powered from the

emergency power supply system provides no additional benefit.

(Operability of the emergency power supply system is tested in

accordance with Specification 3.7.) The proposed revision is in

accordance with the exception for plants with a permanent Class 1E

power supply to these valves as stated in CE STS, SR 3.4.11.4.

Deletion of the requirement of TS 3.2, Table 3-5, Item 15, to

test spent fuel pool surveillance coupons for a change in hardness

corrects an oversight in the Application for Amendment dated

December 7, 1992.

As stated in the Safety Evaluation Report enclosed with

Amendment 155, ``Each

[[Page 44362]]

coupon, upon its removal from the mounting jacket, will be analyzed

according to the following tests:

visual observation and photography

neutron attenuation

dimensional measurements (length, width, and thickness)

weight and specific gravity.''

The tests listed above are sufficient to detect degradation of

the Boral material and do not require that the surveillance coupons

be tested for hardness.

Based on the above discussion, the proposed changes clarify and

standardize existing surveillance requirements, remove redundant

requirements, correct minor oversights from previous amendment

requests or are in accordance with CE STS. Thus, none of the

requested changes 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 revisions will not result in any physical

alterations to the plant configuration, changes to setpoint values,

or changes to the application of setpoints or limits. Therefore, the

proposed changes do 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 proposed changes clarify existing surveillance requirements,

remove redundant requirements, correct minor oversights from

previous amendment requests or are in accordance with CE STS. Thus,

none of the requested changes involves 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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

Pennsylvania Power and Light Company, Docket No. 50-388 Susquehanna

Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: May 20, 1996, as supplemented by letter

dated July 25, 1996

Description of amendment request: This amendment request would

modify the Technical Specifications for the unit by: changing the

Minimum Critical Power Ratio safety limit values, adding a reference to

reflect the use of the ANF-B Critical Power Correlation, and modifying

the associated Bases.

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 change to the ANFB correlation and corresponding MCPR Safety

Limits does not physically change the plant systems, structures, or

components. Thus, the probability of an event evaluated in the SAR

is not increased. The acceptance criterion for the MCPR Safety Limit

(i.e., 99.9% of the fuel rods expected to avoid boiling transition)

is not changed. Only the methodology used to demonstrate compliance

is changed.

Therefore, the consequences of anticipated operational

occurrences (which must show the Safety Limit is not violated) are

not changed. Results of incorporating this change will not

significantly increase the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

As stated above, this methodology change does not impact the

acceptance criteria for the MCPR Safety Limits and does not

physically change the plant systems, structures, or components.

Since no changes to the physical plant are being made, this 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.

A cycle specific MCPR Safety Limit analysis was performed by SPC

[Siemens Power Corporation]. This analysis used NRC approved methods

described in the SPC report: ANF-524(P)(A), Revision 2 and

Supplement 1, Revision 2. The MCPR Safety Limit value is calculated

such that at least 99.9% of the fuel rods are expected to avoid

boiling transition during normal operation or anticipated operation

occurrences. Both the existing analysis using XN-3 and the new

analysis using ANFB utilize NRC approved methods to accomplish this

same objective. Therefore, the change to an ANFB based Safety Limit

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: Osterhout Free Library,

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

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

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

NRC Project Director: John F. Stolz

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

Peak Steam Electric Station (CPSES), Units 1 and 2, Somervell

County, Texas

Date of amendment request: July 31, 1996

Brief description of amendments: Based on analyses of the core

configuration and expected operation for CPSES Unit 1, Cycle 6, the

proposed amendments would revise core safety limit curves and

Overtemperature N-16 reactor trip setpoints. In addition, the TU

Electric Small Break LOCA Topical Report on the Core Operating Limits

Report Technical Specification is incorporated. The topical report

change is applicable to both Units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1.a. Revision to the Unit 1 Core Safety Limits

Analyses of reactor core safety limits are required as part of

reload calculations for each cycle. TU Electric has performed the

analyses of the Unit 1, Cycle 6 core configuration to determine the

reactor core safety limits. The methodologies and safety analysis

values result in new operating curves which, in general, permit

plant operation over a similar range of acceptable conditions. This

change means that if a transient were to occur with the plant

operating at the limits of the new curve, a different temperature

and power level might be attained than if the plant were operating

within the bounds of the old curves. However, since the new curves

were developed using NRC approved methodologies which are wholly

consistent with and do not represent a change in the Technical

Specification BASES for safety limits, all applicable postulated

transients will continue to be properly mitigated. As a result,

there will be no significant increase in the consequences, as

determined by accident analyses, of any accident previously

evaluated.

1.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip

Setpoints, Parameters and Coefficients

As a result of changes discussed, the Overtemperature N-16

reactor trip setpoint has been recalculated. These trip setpoints

help ensure that the core safety limits are maintained and that all

applicable limits of the safety analysis are met.

Based on the calculations performed, the safety analysis value

for Overtemperature N-16 reactor trip setpoint has changed. This

essentially means if a transient were to occur, the actual

temperature and power level achievable prior to initiating a reactor

trip could be slightly higher. However, the analyses performed show

that, using the TU Electric methodologies, all applicable limits of

the safety analysis are met. This setpoint

[[Page 44363]]

provides a trip function which allows the mitigation of postulated

accidents and has no impact on accident initiation. Therefore, the

changes in safety analysis values do not involve an increase in the

probability of an accident and, based on satisfying all applicable

safety analysis limits, there is no significant increase in the

consequences of any accident previously evaluated.

In addition, sufficient operating margin has been maintained in

the overtemperature setpoint such that the risk of turbine runbacks

or reactor trips due to upper plenum flow anomalies or other

operational transients will be minimized, thus reducing potential

challenges to the plant safety systems.

1.c. Incorporation of TU Electric Small Break LOCA Topical

Report, RXE-95-0001-P.

TU Electric has submitted the topical report ``Small Break Loss

of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans

to use the report to support Unit 1 Cycle 6. In order to accomplish

this activity, it is necessary to include the topical report in the

list of NRC-approved methodologies in Technical Specification

6.9.1.6b. Use of this topical report is contingent upon NRC

approval; therefore, inclusion of this report in Section 6 of the

Technical Specifications is administrative in nature and does not

change the probability or consequences of an accident.

2. The proposed changes involve the use of revised safety

analysis values and the calculation of new reactor core safety

limits and reactor trip setpoints. As such, the changes play an

important role in the analysis of postulated accidents but none of

the changes effect plant hardware or the operation of plant systems

in a way that could initiate an accident. Therefore, the proposed

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

accident from any accident previously evaluated.

3. In reviewing and approving the methods used for safety

analyses and calculations, the NRC has approved the safety analysis

limits which establish the margin of safety to be maintained. While

the actual impact on safety is discussed in response to question 1,

the impact on margin of safety is discussed below:

3.a.

Revision to the Unit 1 Reactor Core Safety Limits

The TU Electric reload analysis methods have been used to

determine new reactor core safety limits. All applicable safety

analysis limits have been met. The methods used are wholly

consistent with Technical Specification BASES 2.1 which is the bases

for the safety limits. In particular, the curves assure that for

Unit 1, Cycle 6, the calculated DNBR is no less than the safety

analysis limit and the average enthalpy at the vessel exit is less

than the enthalpy of saturated liquid. The acceptance criteria

remains valid and continues to be satisfied; therefore, no change in

a margin of safety occurs.

3.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip

Setpoints, Parameters and Coefficients

Because the reactor core safety limits for CPSES Unit 1, Cycle 6

are recalculated, the Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor trip setpoint which

protect the reactor core safety limits must also be recalculated.

The Overtemperature N-16 reactor trip setpoint helps prevent the

core and Reactor Coolant System from exceeding their safety limits

during normal operation and design basis anticipated operational

occurrences. The most relevant design basis analysis in Chapter 15

of the CPSES Final Safety Analysis Report (FSAR) which is affected

by the change in the safety analysis value for the CPSES Unit 1

Overtemperature N-16 reactor trip setpoint is the Uncontrolled Rod

Cluster Control Assembly Bank Withdrawal at Power (FSAR Section

15.4.2). This event has been re-analyzed with the revised safety

analysis value for the Overtemperature N-16 reactor trip setpoint to

demonstrate compliance with event specific acceptance criteria.

Because all event acceptance criteria are satisfied, there is no

degradation in a margin of safety.

The nominal Reactor Trip System instrumentation setpoints values

for the Overtemperature N-16 reactor trip setpoint (Technical

Specification Table 2.2-1) are determined based on a statistical

combination of all of the uncertainties in the channels to arrive at

a total uncertainty. The total uncertainty plus additional margin is

applied in a conservative direction to the safety analysis trip

setpoint value to arrive at the nominal and allowable values

presented in Technical Specification Table 2.2-1. Meeting the

requirements of Technical Specification Table 2.2-1 assures that the

Overtemperature N-16 reactor trip setpoint assumed in the safety

analyses remains valid. The CPSES Unit 1, Cycle 6 Overtemperature N-

16 reactor trip setpoint is different from previous cycles which

provides more operational flexibility to withstand mild transients

without initiating automatic protective actions. Although the

setpoint is different, the Reactor Trip System instrumentation

setpoint values for the Overtemperature N-16 reactor trip setpoint

are consistent with the safety analysis assumption which has been

analytically demonstrated to be adequate to meet the applicable

event acceptance criteria. Thus, there is no reduction in a margin

of safety.

3.c. Revise 6.9.1.6b to include Topical Report RXE-95-001-P,

``Small Break Loss of Coolant Accident Methodology''

TU Electric has submitted the topical report ``Small Break Loss

of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans

to use the report to support Unit 1 Cycle 6. In order to accomplish

this activity, it is necessary to include the topical report in the

list of NRC-approved methodologies in Technical Specification

6.9.1.6b. Use of this topical report is contingent upon NRC

approval; therefore, inclusion of this report in Section 6 of the

Technical Specifications is administrative in nature and does not

reduce the margin of safety.

Using the NRC approved TU Electric methods, the reactor core

safety limits are determined such that all applicable limits of the

safety analyses are met. Because the applicable event acceptance

criteria continue to be met, there is no significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Project Director: William D. Beckner

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

Steam Electric Station (CPSES), Units 1 and 2, Somervell County,

Texas

Date of amendment request: July 31, 1996

Brief description of amendments: The proposed amendments would

revise the Technical Specifications by (1) changing the battery charger

ratings; (2) by clarifying the meaning of the term ``associated

inverter''; and by (3) deleting the protection channel and the vital

bus ratings for the instrument busses identified for Mode 1 through 4.

These changes are associated with a plant modification in which the

inverters and battery chargers are being replaced and an installed

spare inverter is being added for each safety train. These changes are

equally applicable to CPSES Units 1 and 2.

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. DO THE PROPOSED CHANGES INVOLVE A SIGNIFICANT INCREASE IN THE

PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED?

CHANGE TO IDENTIFY BATTERY CHARGER RATINGS

The first proposed change replaces the test amperes with the

design value for the replacement battery charger and allows a

voltage range (greater than or equal to 130 volts) instead of a

single value. The intent of the surveillance requirement or the

surveillance frequency is not changed. The replacement inverters and

battery chargers will continue to provide the capacity needed to

perform the required safety functions. The revised surveillance will

continue to assure that the battery chargers are capable of

performing as designed. Therefore this change does not impact the

probability or the consequences of an accident previously evaluated.

CLARIFICATION TO DEFINE ASSOCIATED INVERTER

The second proposed change adds a foot note to clarify the term

``associated inverter'' by describing it as, ''... the dedicated

inverter or installed spare inverter.'' Also the Bases

[[Page 44364]]

for this specification is revised to reflect the basis for this

change. This change allows use of an installed spare inverter (for

each train) having the capability to energize the Instrument Bus for

the protection channel or the vital bus. Procedural controls and

interlocks ensure that the spare is available to feed only one of

the protection channel or vital bus Instrument Bus at a time, in the

event the dedicated inverter is not available. Procedural controls

and interlocks also ensure that the installed spare inverter is fed

from the same power source as that of the dedicated inverter not in

service and whose loads are being fed by the spare inverter. This

proposed design only allows the spare inverter for a safety train to

be manually aligned to replace only one of the four inverters in

that train at a time.

The installation of a spare inverter for each train and the

associated design configuration increases the availability of

energized Instrument Bus for the protection channel and vital bus.

These changes do not involve an increase in the probability or

consequences of an accident previously evaluated.

DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR

INSTRUMENT BUS

The third proposed change deletes specifying of the protection

channel and vital bus KVA ratings for the Instrument Bus. The

ratings of inverter that feeds these instrument buses are being

described in other Licensing Bases Documents or Design Basis

Documents. There is no change proposed to the intent of the action

statements.

This is considered an administrative change and does not impact

the probability or consequences of an accident previously evaluated.

2. DO THE PROPOSED CHANGES CREATE THE POSSIBILITY OF A NEW OR

DIFFERENT KIND OF ACCIDENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED?

CHANGE TO IDENTIFY BATTERY CHARGER RATINGS

Replacing the inverters and battery chargers and changing the

parameters of the battery charger surveillance test to match the

replacement chargers does not alter the functional modes of this

portion of the design and does not result in any new failure modes.

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

accident from any previously evaluated.

CLARIFICATION TO DEFINE ASSOCIATED INVERTER

The second proposed change allows use of an installed spare

inverter for each train to energize the one of the Instrument Bus

for the protection channel and vital bus at a time for the

respective safety train while its dedicated inverter is not

available. The spare inverter is such that it has the capability to

support the maximum load for the protection channel or vital bus.

Manually aligning the installed inverter to replace on[e] of the

dedicated inverters is essentially equivalent to a repair activity

which replaces a faulted inverter with a new inverter. In addition,

procedural controls and interlocks are provided to ensure the proper

alignment of the installed spare when it is used. The proposed

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

from any previously evaluated.

DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR

INSTRUMENT BUS

The third proposed change as discussed earlier does not change

intent of the Technical Specifications action statements. This is an

administrative change which does not introduce new failure modes and

has no new or different accidents from any previously evaluated are

created.

3. DO THE PROPOSED CHANGES INVOLVE A SIGNIFICANT REDUCTION IN

MARGIN OF SAFETY?

The relevant Technical Specification sections proposed for

changes: (1) ensure that the battery charger is capable of charging

the battery by performing the surveillance at 18 month frequency;

(2) establish operability requirements of the Instrument Bus for the

protection channel and vital bus in MODES 1 through 6; and (3)

identify the actions required for not meeting item 2.

These proposed changes do not alter the intent of the above

requirements; however replacement of the currently installed

inverters with inverters which are expected to be more reliable and

available and the addition of a spare inverter per safety train to

energize Instrument Bus for protection channel and vital bus does

increase the reliability of the instrument busses for the train.

Allowing credit for this spare inverter in meeting the operability

requirements of Instrument Bus for the protection channel and vital

bus, minimize potential plant shutdowns due to non-energized

instrument from its dedicated inverter. These changes do not involve

a significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Project Director: William D. Beckner

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 9, 1996

Description of amendment request: The proposed amendment would

revise the Safety Limits for Minimum Critical Power Ratio (MCPR) based

upon a Vermont Yankee plant and cycle specific analysis, performed by

General Electric. The revised MCPR Safety Limits are needed to

accommodate Vermont Yankee's core design for upcoming refueling cycle

number 19. Specifically, the MCPR Safety Limits of 1.07 and 1.08 in the

Vermont Yankee Technical Specifications (TS) section 1.1.A are proposed

to be increased to 1.10 and 1.12 for two loop and single loop

operation, respectively.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The Safety Limit Minimum Critical Power Ratio (MCPR) is

defined to ensure that during normal operation and Anticipated

Operational Transients (AOTs), at least 99.9% of the fuel rods in

the core do not experience transition boiling. Core MCPR operating

limits are developed to ensure these Safety Limits are maintained in

the event of the worst case transient. Since the Safety Limit MCPR

will be maintained at all times, operation under the proposed

changes will ensure at least 99.9% of the fuel rods in the core do

not experience transition boiling and no significant radiological

release will result. Therefore, this Safety Limit MCPR change does

not affect the probability or consequences of a previously evaluated

accident.

(2) The proposed changes do not involve any new modes of

operation or any plant modifications. Establishment and monitoring

of the operating limits will continue as per established procedure.

The proposed changes to these limits do not result in the creation

of any new precursors to an accident. Therefore, the proposed change

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

accident from any previously analyzed.

(3) The Safety Limit MCPR values were evaluated by General

Electric based upon a cycle specific Vermont Yankee analysis, using

NRC approved methods. The resulting limits are more conservative

than the previous generic limits and will continue to assure that at

least 99.9% of the fuel rods in the core do not experience

transition boiling during analyzed transients. This acceptance

criteria ensures the safety design limit of ``no damage to a nuclear

system process barrier shall result from forces associated with

AOTs.'' Therefore, the implementation of the proposed change does

not involve a significant reduction in [a] margin of safety.

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

that, although the proposed change does not involve a plant

modification, the reason for the proposed higher safety limit MCPRs is

the cycle-specific core design and the local power distribution in the

slightly higher enriched fresh GE-9B fuel bundles. This new fuel will

be loaded during the September/October 1996 refueling outage. In

conjunction with the proposed safety limit MCPRs and the core operating

limits determined in accordance with Vermont Yankee TS 6.7.A.4, the new

fuel load will not involve a significant increase in the probability or

consequences of an

[[Page 44365]]

accident previously evaluated nor a significant reduction in a margin

of safety. In addition, the new fuel load does not create the

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

previously evaluated. Based on this review, it appears that the three

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

to determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

Attorney for licensee: R. K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: May 1, 1996

Brief description of amendment: The proposed amendment will modify

the definition of ``Core Alteration,'' and the limiting condition for

operation, Surveillance conditions and Bases section associated with

Technical Specification 3.7.C, ``Secondary Containment.''

Date of issuance: August 12, 1996

Effective date: August 12, 1996

Amendment No.: 166

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28606) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 12, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

Date of application for amendment: July 17, 1995, as supplemented

May 2, 1996, and July 1, 1996.

Brief description of amendment: The change revises technical

specification (TS) section 3.8 to specify that the spent fuel building

refueling filter fan and at least one containment purge fan shall be

shown to operate within plus or minus 10 percent of the design flow.

Date of issuance: August 6, 1996

Effective date: August 6, 1996

Amendment No. 172

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

Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47615). The May 2, and July 1, 1996, letters provided clarifying

information that did not affect the proposed no significant hazards

consideration. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 6, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

Date of application for amendment: June 6, 1996

Brief description of amendment: The amendment revises technical

specifications (TS) Section 4.2.3 to allow the licensee to defer the

ultrasonic inspection of the reactor coolant pump flywheel for one

operating cycle.

Date of issuance: August 9, 1996

Effective date: August 9, 1996

Amendment No. 173

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

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34888) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 9, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

North Carolina

Date of application for amendment: May 31, 1996

Brief description of amendment: The amendment revises Technical

Specifications (TS) Table 3.3-7, Seismic Monitoring Instrumentation,

and TS Table 4.3-4, Seismic Monitoring Instrumentation Surveillance

Requirements, to correct the location described for one of the three

Triaxial Peak Accelerograph recorders.

Date of issuance: August 7, 1996

Effective date: August 7, 1996

Amendment No. 66

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34888) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 7, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Cameron Village Regional

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

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

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

Date of application for amendments: April 16, 1996

Brief description of amendments: The amendments revise the

Technical

[[Page 44366]]

Specifications (TSs) to eliminate selected response time testing

requirements based on analyses performed by the Boiling Water Reactor

Owners' Group as documented in NEDO-32291. The affected TS sections are

3/4.3.1, ``Reactor Protection System Instrumentation;'' 3/4.3.2,

``Isolation Actuation Instrumentation;'' and 3/4.3.3, ``Emergency Core

Cooling System Actuation Instrumentation.''

Date of issuance: August 14, 1996

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

Amendment Nos.: 114 and 99

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

revised the Technical Specifications.

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

25702) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 14, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment: December 21, 1995

Brief description of amendment: The amendment revises the Technical

Specifications (TS) to implement 10 CFR Part 50, Appendix J - Option B,

by referring to Regulatory Guide 1.163, ``Performance-Based Containment

Leak-Test Program.'' Specifically, changes have been made to TS Section

3/4.6.1.2, ``Primary Containment Leakage,'' TS 3/4.6.1.3, ``Primary

Containment Air Locks,'' TS 3/4.6.1.5, ``Primary Containment Structural

Integrity,'' TS 6.0, ``Administrative Controls,'' and their associated

Bases.

Date of issuance: August 8, 1996

Effective date: August 8, 1996, with full implementation within 45

days.

Amendment No.: 108

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7551) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 8, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: April 19, 1996, and supplements

dated May 10 and May 28, 1996.

Brief description of amendments: The amendment changes the

Technical Specifications to address frequency extension on a periodic

basis, deletes separate notification requirements for an inoperable

startup transformer, and allows the operating residual heat removal

loop to be removed from operation, under certain conditions, during

refueling.

Date of Issuance: August 6, 1996

Effective Date: August 6, 1996

Amendment Nos.: 189 and 183Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34892) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 6, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: July 26, 1995, and supplemented

March 13, May 3, and May 9, 1996.

Brief description of amendments: Change TS 6.9.1.7, Core Operating

Limits Report, resulting from a reanalysis of the small break loss-of-

coolant accident for the Turkey Point Units using the NOTRUMP code

including the COSI safety injection (SI) condensation model.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment Nos. 190 and 184Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47618). The supplements dated March 13, May 3, and May 9, 1996

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 August 13, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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

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

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

Matagorda County, Texas

Date of amendment request: May 1, 1996

Brief description of amendments: The amendments changed the

technical specifications to implement 10 CFR Part 50, Appendix J,

Option B, by referring to Regulatory Guide 1.163, ``Performance-Based

Containment Leak-Test Program.'' Part of the requested change, that

regarding the frequency of leakage rate testing the normal containment

purge valves and the supplementary containment purge valves, was

denied.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment Nos.: 84 and 71

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

revised the Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28616) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 13, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Wharton County Junior College,

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

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

Linn County, Iowa

Date of application for amendment: November 30, 1995

Brief description of amendment: The amendment implements the Option

I-D long-term stability solution and removes the existing SIL-380 Rev.

1-based specifications. In addition, the amendment requires a plant

scram be initiated should the plant enter natural circulation

conditions and prohibits restarting a recirculation pump while in

natural circulation. Finally, this amendment deletes Technical

Specification (TS) actions and surveillance requirements related to

core plate differential pressure noise while in single recirculation

pump operation (SLO).

Date of issuance: August 7, 1996

Effective date: August 7, 1996

Amendment No.: 215

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications.

[[Page 44367]]

Date of initial notice in Federal Register: March 13, 1996 (61 FR

10394) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 7, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401

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

Linn County, Iowa

Date of application for amendment: November 15, 1995, as

supplemented April 9, 1996

Brief description of amendment: The amendment revises the

requirements for the End of Cycle Recirculation Pump Trip logic to

match more closely the assumptions applicable to the turbine trip

events for which it was installed. The surveillance requirements are

also revised, based on those same assumptions.

Date of issuance: August 8, 1996

Effective date: August 8, 1996

Amendment No.: 216

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications.

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

1629) The April 9, 1996, submittal was clarifying in nature and did not

affect the no significant hazards determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated August 8, 1996. No significant hazards consideration comments

received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401

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

Linn County, Iowa

Date of application for amendment: January 18, 1996

Brief description of amendment: The amendment revises the setpoint

at which the Reactor Water Cleanup (RWCU) system isolates, based on

reactor vessel water level. In particular, the amendment changes the

Group 5 isolation from isolating on ``reactor water level low'' to

``reactor water level low-low.''

Date of issuance: August 8, 1996

Effective date: August 8, 1996, and shall be implemented prior to

startup from RFO 14.

Amendment No.: 217

Facility Operating License No. DPR-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 14, 1996 (61

FR 5814) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 8, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S. E., Cedar Rapids, Iowa 52401

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of application for amendments: January 12, 1996 (AEP:NRC:1233)

Brief description of amendments: The amendments modify the

Technical Specifications to delete the surveillance requirement

demonstrating operability of the emergency power supply for the

pressurizer power operated relief valves and block valves.

Date of issuance: August 15, 1996

Effective date: August 15, 1996, with full implementation within 45

days

Amendment Nos.: 211 and 196

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7554) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 15, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske

Memorial Library, 500 Market Street, St. Joseph, Michigan 49085

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

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

Date of application for amendment: February 7, 1996, as

supplemented July 26, 1996.

Brief description of amendment: The amendment revises the operating

license, TSs and associated Bases to implement Option B ``Performance-

Based Requirements'' of Appendix J to 10 CFR Part 50 for Type A, B, and

C leakage rate testing.

Date of issuance: August 13, 1996

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

30 days.

Amendment No.: 74

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

Technical Specifications and operating license.

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

20849) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 13, 1996. 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.

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 3, 1996

Brief description of amendment: The amendment removes, on a one-

time basis during the cycle 13 mid-cycle offload/reload activities, the

Technical Specification (TS) requirement that the boron concentration

in all filled portions of the reactor coolant system be ``uniform.''

The requested change also adds a footnote indicating that it is

acceptable for the boron concentration of the water volumes in the

steam generators and the connecting piping to be as low as 1300 parts

per million. The TS Bases are also updated to reflect the one-time TS

change.

Date of issuance: August 12, 1996

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

within 30 days.

Amendment No.: 201

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

Technical Specifications.

Date of initial notice in Federal Register: July 11, 1996 (61 FR

36583) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 12, 1996. 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, CT 06360, and Waterford Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385

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

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

Pennsylvania

Date of application for amendments: February 29, 1996

Brief description of amendments: These amendments relocate

Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam

Electric Station Technical Requirements Manual, a document which is

controlled under the requirements of 10 CFR 50.59.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment Nos.: 159 and 130

[[Page 44368]]

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15992) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated August 13, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,

Alabama

Date of application for amendments: May 20, 1996 (TS 373)

Brief description of amendment: The amendments incorpore the

guidance of Generic Letter 87-09 in the technical specifications,

allowing a 24-hour delay in implementing action requirements due to a

missed surveillance requirement.

Date of issuance: August 5, 1996

Effective Date: August 5, 1996

Amendment Nos.: 230, 245 and 205

Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:

Amendments revised the Technical Specifications.

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

31185) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 5, 1996. No significant

hazards consideration comments received: None

Local Public Document Room location: Athens Public library, South

Street, Athens, Alabama 35611

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: May 29, 1996

Brief description of amendment: The amendment authorizes revision

of the Final Safety Analysis Report (FSAR) to incorporate a

modification to the facility that will reduce the single failure trip

potential for the main feedwater and bypass valves.

Date of issuance: August 13, 1996

Effective date: August 13, 1996

Amendment No.: 115

Facility Operating License No. NPF-30: The amendment revised the

Final Safety Analysis Report.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34900) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 13, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: June 4, 1996

Brief description of amendment: The amendment revises the Technical

Specifications by reducing the surveillance test frequencies for the

radiation monitoring system (Table TS 4.1-1) and the control rods

(Table TS 4.1-3) in accordance with the guidance of Generic Letter 93-

05, ``Line-Item Technical Specifications Improvements to Reduce

Surveillance Requirements for Testing During Power Operation,'' dated

September 27, 1993.

Date of issuance: August 7, 1996

Effective date: August 7, 1996

Amendment No.: 125

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34901) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated August 7, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 29, 1994, as superseded by letter

dated September 15, 1995, and subsequently supplemented by letters

dated March 8, 1996, April 18, 1996, June 14, 1996, and July 12, 1996.

Brief description of amendment: The amendment revises TS 3/4.8.1,

``Electric Power Systems - A.C. Sources,'' and its associated Bases to

achieve an overall improvement in emergency diesel generator

reliability and availability.

Date of issuance: August 9, 1996

Effective date: August 9, 1996, to be implemented within 90 days of

the date of issuance.

Amendment No.: 101

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

Technical Specifications.

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

25716) The June 14, 1996, and July 12, 1996, supplemental letters

provided Bases page changes and did not change the initial no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated August 9, 1996. 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 21st day of August 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II Office of Nuclear Reactor

Regulation

[Doc. 96-21813 Filed 8-27-96; 8:45 am]

BILLING CODE 7590-01-F