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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

publish notice of any 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, September 6, 2002, through September 19,

2002. The last biweekly notice was published on September 17, 2992 (67

FR 58635).

[[Page 61675]]

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administrative 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 Commission's Public

Document Room (PDR), located at One White Flint North, 11555 Rockville

Pike (first floor), Rockville, Maryland. The filing of requests for a

hearing and petitions for leave to intervene is discussed below.

By October 31, 2002, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

persons should consult a current copy of 10 CFR 2.714,\1\ which is

available at the Commission's PDR, located at One White Flint North,

11555 Rockville Pike (first floor), Rockville, Maryland. Publicly

available records will be accessible from the Agencywide Documents

Access and Management System's (ADAMS) Public Electronic Reading Room

on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-

collections/cfr/. 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.

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\1\ 1. The most recent version of Title 10 of the Code of

Federal Regulations, published January 1, 2002, inadvertently

omitted the last sentence of 10 CFR 2.714(d) and subparagraphs

(d)(1) and (2), regarding petitions to intervene and contentions.

Those provisions are extant and still applicable to petitions to

intervene. Those provisions are as follows: ``In all other

circumstances, such ruling body or officer shall, in ruling on--

(1) A petition for leave to intervene or a request for hearing,

consider the following factors, among other things:

(i) The nature of the petitioner's right under the Act to be

made a party to the proceeding.

(ii) The nature and extent of the petitioner's property,

financial, or other interest in the proceeding.

(iii) The possible effect of any order that may be entered in

the proceeding on the petitioner's interest .

(2) The admissibility of a contention, refuse to admit a

contention if:

(i) The contention and supporting material fail to satisfy the

requirements of paragraph (b)(2) of this section; or

(ii) The contention, if proven, would be of no consequence in

the proceeding because it would not entitle petitioner to relief.''

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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

[[Page 61676]]

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: Rulemaking and

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

located at One White Flint North, 11555 Rockville Pike (first floor),

Rockville, Maryland, by the above date. Because of continuing

disruptions in delivery of mail to United States Government offices, it

is requested that petitions for leave to intervene and requests for

hearing be transmitted to the Secretary of the Commission either by

means of facsimile transmission to 301-415-1101 or by e-mail to

[email protected]. A copy of the request for hearing and petition

for leave to intervene should also be sent to the Office of the General

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

and because of continuing disruptions in delivery of mail to United

States Government offices, it is requested that copies be transmitted

either by means of facsimile transmission to 301-415-3725 or by e-mail

to [email protected]. A copy of the request for hearing and

petition for leave to intervene should also be sent 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 PDR, located at One White Flint North, 11555 Rockville

Pike (first floor), Rockville, Maryland. Publicly available records

will be accessible from the Agencywide Documents Access and Management

System's (ADAMS) Public Electronic Reading Room on the Internet at the

NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not

have access to ADAMS or if there are problems in accessing the

documents located in ADAMS, contact the NRC PDR Reference staff at 1-

800-397-4209, 304-415-4737 or by e-mail to [email protected].

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: August 28, 2002.

Description of amendment request: The amendment would revise

Technical Specification (TS) 3/4.9.9, ``Containment Ventilation

Isolation System'' and associated Bases to allow the use of

administrative controls on open containment penetrations during core

alterations.

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 changes modify TS requirements similar to that

previously reviewed and approved by the NRC in Harris Nuclear Plant

(HNP) License Amendment 104. The administrative controls proposed by

this change are currently being used for the same applicable

penetrations as part of TS 3.9.4. This change would permit opening

up the applicable penetrations under administrative controls if the

containment ventilation isolation system were inoperable. HNP has

demonstrated (in License Amendment 104) that the radiological

consequences were acceptable for a fuel handling accident occurring

simultaneously with an open penetration. For the purpose of the

applicable analysis, no credit was given for isolating the

penetration and dose consequences remained below applicable

regulatory limits. The proposed change does not modify the design or

operation of equipment used to move spent fuel or to perform core

alterations.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Containment penetrations are designed to form part of the

containment pressure boundary. The proposed change provides for

administrative controls and operating restrictions for containment

penetrations consistent with guidance approved by the NRC staff.

Containment penetrations are not an accident initiating system as

described in the Final Safety Analysis Report [FSAR]. The proposed

change does not affect other Structures, Systems, or Components. The

operation and design of containment penetrations in operational

modes 1-4 will not be affected by this proposed change.

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 changes modify similar required Actions previously

reviewed and approved by the NRC in HNP License Amendment 104. The

proposed change to containment penetrations does not significantly

affect any of the parameters that relate to the margin of safety as

described in the Bases of the TS or the FSAR. Accordingly, NRC

Acceptance Limits are not significantly affected by this change.

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.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Kahtan N. Jabbour, Acting.

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

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

Carolina

Date of amendment request: August 30, 2002.

Description of amendment request: The amendment would revise

Technical Specifications Definitions 1.13, Engineered Safety Features

(ESF) Response Time and 1.29, Reactor Trip System (RTS) Response Time.

Also proposed in this change request are revisions to Surveillance

Requirements 4.3.1.2 and 4.3.2.2 and Bases Sections B 3/4.3.1 and B 3/

4.3.2. These changes will revise the definition and surveillance

requirements for response

[[Page 61677]]

time testing of the Engineered Safety Feature Actuation System (ESFAS)

and the RTS. These changes are in conformance with changes approved in

WCAP-13632-P-A, Revision 2, and WCAP-14036-P-A, Revision 1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The change to the Harris Nuclear Plant (HNP) Technical

Specification (TS) does not result in a condition where the design,

material, and construction standards that were applicable prior to

the change are altered. The same RTS and ESFAS instrumentation is

being used; the time response allocations/modeling assumptions in

the Final Safety Analysis Report (FSAR) Chapter 15 analyses are

still the same; only the method of verifying the time response is

changed. The proposed change will not modify any system interface

and could not increase the likelihood of an accident since these

events are independent of this change. The proposed change will not

change, degrade or prevent actions or alter any assumptions

previously made in evaluating the radiological consequences of an

accident described in the FSAR.

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

new or different kind of accident from any accident previously

evaluated.

This change does not alter the performance of process protection

racks, Nuclear Instrumentation, and logic systems used in the plant

protection systems. Replacement transmitters will still have

response time verified by testing before being placed in operational

service. Changing the method of periodically testing these systems

(assuring equipment operability) from response time testing to

calibration and channel checks will not create any new accident

initiators or scenarios. Periodic surveillance of these systems will

continue and may be used to detect degradation that could cause the

response time to exceed the total allowance. The total time response

allowance for each function bounds all degradation that cannot be

detected by periodic surveillance. Implementation of the proposed

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

accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

This change does not affect the total system response time

assumed in the safety analysis. The periodic system response time

verification method for the process protection racks, Nuclear

Instrumentation, and logic systems is modified to allow the use of

actual test data or engineering data. The method of verification

still provides assurance that the total system response is within

that defined in the safety analysis, since calibration tests will

continue to be performed and may be used to detect any degradation

which might cause the system response time to exceed the total

allowance. The total response time allowance for each function

bounds all degradation that cannot be detected by periodic

surveillance. Based on the above, it is concluded that the proposed

change does not result in a significant reduction in margin with

respect to plant safety.

Pursuant to 10 CFR 50.91, the preceding analysis provides a

determination that the proposed Technical Specifications change

poses no significant hazard as delineated by 10 CFR 50.92.

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.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Kahtan N. Jabbour, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power

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

Date of amendment request: August 12, 2002.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.8.2.3, ``Electrical Power

Systems, D.C. Distribution--Operating,'' TS 3.8.2.4, ``Electrical Power

Systems, D.C. Distribution--Shutdown,'' and TS 3.8.2.5, ``Electrical

Power Systems, D.C. Distribution Systems (Turbine Battery)--Operating''

to use standard technical specification terminology in order to provide

enhanced readability and usability. The proposed amendment would also

provide additional criteria for determining battery operability upon

restoration from a recharge or equalizing charge.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed Technical Specifications changes for relocation of

information which defines the operability of the D.C. electrical

power subsystems will not create any new failure modes, will not

cause an accident to occur, and will not result in any change in the

operation of accident mitigation equipment. Relocation of this

information will not have an adverse impact on any accident

initiators. Proper operation of the D.C. electrical power subsystems

will still be verified. As a result, the design basis accidents will

remain the same postulated events described in the Millstone Unit

No. 2 Final Safety Analysis Report, and the consequences of the

design basis accidents will remain the same. Therefore, the proposed

changes will not increase the probability or consequences of an

accident previously evaluated.

The proposed changes for deletion of redundant actions

requirements and reformatting of surveillance requirements

associated with the D.C. electrical power subsystems will not cause

an accident to occur and will not result in any change in the

operation of associated accident mitigation equipment. The proposed

changes will not have an adverse impact on any accident initiators.

Proper operation of the D.C. electrical power subsystems will still

be verified. As a result, the design basis accidents will remain the

same postulated events described in the Millstone Unit No. 2 Final

Safety Analysis Report, and the consequences of the design basis

accidents will remain the same. Therefore, the proposed changes will

not increase the probability or consequences of an accident

previously evaluated.

The proposed changes to the surveillance requirements for the

D.C. electrical power subsystems to add additional criteria relating

to physical damage or deterioration and its impact on battery

performance do not affect any existing accident initiators or

precursors. The proposed changes will not create any adverse

interactions with other systems that could result in initiation of a

design basis accident. Proper operation of the D.C. electrical power

subsystems batteries will still be verified. As a result, the design

basis accidents will remain the same postulated events described in

the Millstone Unit No. 2 Final Safety Analysis Report, and the

consequences of the design basis accidents will remain the same.

Therefore, the proposed changes will not increase the probability or

consequences of an accident previously evaluated.

The proposed changes to the surveillance requirements for the

D.C. electrical power subsystems to add additional criteria relating

to demonstrating battery operability following a recharge or

equalizing charge will not have an adverse affect on battery

operability. The proposed changes will not create any adverse

interactions with other systems that could result in initiation of a

design basis accident. Proper operation of the D.C. electrical power

subsystems batteries will still be verified. As a result, the design

basis accidents will remain the same postulated events described in

the Millstone Unit No. 2 Final Safety Analysis Report, and the

consequences of the design basis accidents will remain the same.

Therefore, the proposed changes will not increase the probability or

consequences of an accident previously evaluated.

[[Page 61678]]

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

from any accident previously evaluated.

The proposed changes do not create any new or different accident

initiators or precursors. The proposed changes do not create any new

failure modes for the components of the D.C. electrical power

subsystems and do not affect the interaction between the D.C.

electrical power subsystems and any other system. The proposed

changes do not alter the plant configuration (no new or different

type of equipment will be installed) or require any new or unusual

operator actions. The proposed changes do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The components of the D.C.

electrical power subsystems will continue to function as before, and

will continue to be declared inoperable if their ability to perform

a safety function is impaired. Therefore, the proposed changes do

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

from any accident previously evaluated.

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

The proposed changes will not reduce the margin of safety since

they have no impact on any accident analysis assumption. The

proposed changes do not decrease the scope of equipment currently

required to be operable or subject to surveillance testing, nor do

the proposed changes affect any instrument setpoints or equipment

safety functions. The Technical Specifications will continue to

require that a battery be declared inoperable if physical damage or

abnormal deterioration of the cells, cell plates, or racks that

would degrade battery performance is observed. The proposed changes

do not alter the requirements of the Technical Specification with

respect to the capacity of any battery. The effectiveness of

Technical Specifications will be maintained since the changes will

not alter the operation of any component or system, nor will the

proposed changes affect any safety limits or safety system settings

which are credited in a facility accident analysis. Therefore, there

is no 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.

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

Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT

06385.

NRC Section Chief: James W. Andersen, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power

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

Date of amendment request: August 14, 2002.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) related to Containment

Systems. Specifically, the proposed changes would: (1) Add

clarification to TS 1.7 ``Definitions--Containment Integrity'' (2) add

clarifying information as well revise a portion of Surveillance

Requirement (SR) 4.6.1.1 associated with the affected section of TS

3.6.1.1 ``Containment Integrity;'' (3) revise TS 3.6.3, ``Containment

Isolation Valves,'' to make editorial changes, to add clarifying

information and to add an Action item that would increase the allowed

outage time (AOT) from 4 hours to 72 hours for Containment Isolation

Valves (CIVs) in closed systems, and (4) other changes that are

clarifying and/or administrative in nature. In addition, the TS Bases

would be revised to address the proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed Technical Specification changes associated with

both containment integrity and CIVs that will remove ambiguity,

improve usability, and increase AOT for CIVs in closed systems, will

not cause an accident to occur. Operability requirements for

containment integrity and CIVs will remain the same. The ability of

the equipment associated with the proposed changes to mitigate the

design basis accidents will not be affected. The proposed Technical

Specification requirements are sufficient to ensure the required

accident mitigation equipment will be available and function

properly for design basis accident mitigation. The proposed allowed

outage time is reasonable and consistent with standard industry

guidelines to ensure the accident mitigation equipment will be

restored in a timely manner. In addition, the design basis accidents

will remain the same postulated events described in the Millstone

Unit No. 3 Final Safety Analysis Report, and the consequences of

those events will not be affected. Therefore, the proposed changes

will not increase the probability or consequences of an accident

previously evaluated.

The additional proposed changes to the Technical Specifications

(e.g., relocating information to the Bases, renumbering of

footnotes, renumbering a requirement) will not result in any

technical changes to the current requirements. Therefore, these

additional changes will not increase the probability or consequences

of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes to the Technical Specifications do not

impact any system or component that could cause an accident. The

proposed changes will not alter the plant configuration (no new or

different type of equipment will be installed) or require any

unusual operator actions. The proposed changes will not alter the

way any structure, system, or component functions, and will not

alter the manner in which the plant is operated. The response of the

plant and the operators following an accident will not be different.

In addition, the proposed changes do not introduce any new failure

modes. Therefore, the proposed changes will not create the

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

previously evaluated.

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

The proposed Technical Specification changes associated with

both containment integrity and CIVs that will remove ambiguity,

improve usability, and increase AOT for CIVs in closed systems, will

not cause an accident to occur. Operablity requirements for

containment integrity and CIVs will remain the same. The equipment

associated with the proposed Technical Specification changes will

continue to be able to mitigate the design basis accidents as

assumed in the safety analysis. The proposed allowed outage time is

reasonable and consistent with standard industry guidelines to

ensure the accident mitigation equipment will be restored in a

timely manner. In addition, the proposed changes will not affect

equipment design or operation, and there are no changes being made

to the Technical Specification required safety limits or safety

system settings. The proposed Technical Specification changes will

provide adequate control measures to ensure the accident mitigation

functions are maintained. Therefore, the proposed changes will not

result in a reduction in a margin of safety.

The additional proposed changes to the Technical Specifications

(e.g., relocating information to the Bases, renumbering of

footnotes, renumbering a requirement) will not result in any

technical changes to the current requirements. Therefore, these

additional changes will not result in a reduction in a margin of

safety.

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

this review, it appears that the 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.

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

Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT

06385.

NRC Section Chief: James W. Andersen, Acting.

[[Page 61679]]

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 15, 2002.

Description of amendment request: The proposed amendment would

revise the River Bend Station (River Bend or RBS) reactor vessel

surveillance program required by Title 10 of the Code of Federal

Regulations (10 CFR) part 50, appendix H, section IIIB.3. The change

will incorporate the Boiling Water Reactor Vessel & Internals Project

Integrated Surveillance Program into the RBS licensing basis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Response: No.

Pressure-temperature (P/T) limits (RBS Technical Specifications

Figure 3.4.11-1) are imposed on the reactor coolant system to ensure

that adequate safety margins against nonductile or rapidly

propagating failure exist during normal operation, anticipated

operational occurrences, and system hydrostatic tests. The P/T

limits are related to the nil-ductility reference temperature,

RTNDT, as described in ASME [American Society of

Mechanical Engineers Boiler and Pressure Vessel Code (Code)] Section

III, Appendix G. Changes in the fracture toughness properties of RPV

[reactor pressure vessel] beltline materials, resulting from the

neutron irradiation and the thermal environment, are monitored by a

surveillance program in compliance with the requirements of 10CFR50,

Appendix H. The effect of neutron fluence on the shift in the nil-

ductility reference temperature of pressure vessel steel is

predicted by methods given in RG [Regulatory Guide] 1.99, Rev[ision]

2.

River Bend's current P/T and Power Uprate limits were

established based on adjusted reference temperatures developed in

accordance with the procedures prescribed in RG 1.99, Rev 2,

Regulatory Position 1. Calculation of adjusted reference temperature

by these procedures includes a margin term to ensure conservative,

upper-bound values are used for the calculation of the P/T limits.

When permitted (two or more credible surveillance data sets

available), Regulatory Position 2 (or other NRC [U.S. Nuclear

Regulatory Commission]-approved) methods for determining adjusted

reference temperature will be followed.

This change is not related to any accidents previously

evaluated. This change will not affect P/T limits as given in RBS

Technical Specifications Figure 3.4.11-1 or USAR [Updated Safety

Analysis Report] Figures 5.3-4a and 5.3-4b. This change will not

affect any plant safety limits or limiting conditions of operation.

The proposed change will not affect reactor pressure vessel

performance as no physical changes are involved and RBS vessel P/T

limits will remain conservative in accordance with Reg[ulatory]

Guide 1.99, Rev 2 requirements. The proposed change will not cause

the reactor pressure vessel or interfacing systems to be operated

outside of their design or testing limits. Also, the proposed change

will not alter any assumptions previously made in evaluating the

radiological consequences of accidents. Therefore, the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

Response: No.

The proposed change revises the RBS license basis to reflect

participation in the ISP [Integrated Surveillance Program]. This

proposed change does not involve a modification of the design of

plant structures, systems, or components. The proposed change will

not impact the manner in which the plant is operated as plant

operating and testing procedures will not be affected by the change.

The proposed change will not degrade the reliability of structures,

systems, or components important to safety as equipment protection

features will not be deleted or modified, equipment redundancy or

independence will not be reduced, supporting system performance will

not be downgraded, the frequency of operation of equipment will not

be increased, and increased or more severe testing of equipment will

not be imposed. No new accident types or failure modes will be

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

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

kind of accident from that previously evaluated.

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

margin of safety?

Response: No.

As stated in the River Bend SER [Safety Evaluation Report],

``Appendices G and H of 10CFR50 describe the conditions that require

pressure-temperature limits and provide the general bases for these

limits. These appendices specifically require that pressure-

temperature limits must provide safety margins at least as great as

those recommended in the ASME Code, Section III, Appendix G. * * *

Until the results from the reactor vessel surveillance program

become available, the staff will use Regulatory Guide (RG) 1.99,

Revision 1 [now Revision 2], to predict the amount of neutron

irradiation damage. * * * The use of operating limits based on these

criteria--as defined by applicable regulations, codes, and

standards--will provide reasonable assurance that nonductile or

rapidly propagating failure will not occur, and will constitute an

acceptable basis for satisfying the applicable requirements of

General Design Criteria (GDC) 31.''

Bases for RBS Technical Specification 3.4.11 states: ``The P/T

limits are not derived from Design Basis Accident (DBA) analyses.

They are prescribed during normal operation to avoid encountering

pressure, temperature, and temperature rate of change conditions

that might cause undetected flaws to propagate and cause nonductile

failure of the RCPB [Reactor Coolant Pressure Boundary], a condition

that is unanalyzed. * * * Since the P/T limits are not derived from

any DBA, there are no acceptance limits related to the P/T limits.

Rather, the P/T limits are acceptance limits themselves since they

preclude operation in an unanalyzed condition.''

The proposed change will not affect any safety limits, limiting

safety system settings, or limiting conditions of operation. The

proposed change does not represent a change in initial conditions,

or in a system response time, or in any other parameter affecting

the course of an accident analysis supporting the Bases of any

Technical Specification. The proposed change does not involve

revision of the P/T limits but rather a revision to the surveillance

capsule withdrawal schedule. The current P/T limits were established

based on adjusted reference temperatures for vessel beltline

materials calculated in accordance with Regulatory Position 1 of RG

1.99, Rev 2. P/T limits will continue to be revised as necessary for

changes in adjusted reference temperature due to changes in fluence

according to Regulatory Position 1 until two or more credible

surveillance data sets become available. When two or more credible

surveillance data sets become available, P/T limits will be revised

as prescribed by Regulatory Position 2 of RG 1.99, Rev 2, or other

NRC-approved guidance. Therefore, the proposed change does not

involve a significant reduction in any margins 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.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 21, 2002.

Description of amendment request: The proposed amendment would

revise Surveillance Requirement (SR) 3.0.3 to extend the delay period,

before entering a Limiting Condition for Operation, following a missed

surveillance. The delay period would be extended from the current limit

of ``* * * up to 24 hours or up to the limit of the specified

Frequency, whichever is less'' to ``* * * up to 24 hours or up to the

limit of the

[[Page 61680]]

specified Frequency, whichever is greater.'' In addition, the following

requirement would be added to SR 3.0.3: ``A risk evaluation shall be

performed for any Surveillance delayed greater than 24 hours and the

risk impact shall be managed.''

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice

of opportunity for comment in the Federal Register on June 14, 2001 (66

FR 32400), on possible amendments concerning missed surveillances,

including a model safety evaluation and model no significant hazards

consideration (NSHC) determination, using the consolidated line item

improvement process. The NRC staff subsequently issued a notice of

availability of the models for referencing in license amendment

applications in the Federal Register on September 28, 2001 (66 FR

49714). The licensee affirmed the applicability of the following NSHC

determination in its application dated August 21, 2002.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), an analysis of the issue

of NSHC is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant

Increase in the Probability or Consequences of an Accident Previously

Evaluated

The proposed change relaxes the time allowed to perform a missed

surveillance. The time between surveillances is not an initiator of

any accident previously evaluated. Consequently, the probability of

an accident previously evaluated is not significantly increased. The

equipment being tested is still required to be operable and capable

of performing the accident mitigation functions assumed in the

accident analysis. As a result, the consequences of any accident

previously evaluated are not significantly affected. Any reduction

in confidence that a standby system might fail to perform its safety

function due to a missed surveillance is small and would not, in the

absence of other unrelated failures, lead to an increase in

consequences beyond those estimated by existing analyses. The

addition of a requirement to assess and manage the risk introduced

by the missed surveillance will further minimize possible concerns.

Therefore, this change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a

New or Different Kind of Accident From Any Previously Evaluated

The proposed change does not involve a physical alteration of

the plant (no new or different type of equipment will be installed)

or a change in the methods governing normal plant operation. A

missed surveillance will not, in and of itself, introduce new

failure modes or effects and any increased chance that a standby

system might fail to perform its safety function due to a missed

surveillance would not, in the absence of other unrelated failures,

lead to an accident beyond those previously evaluated. The addition

of a requirement to assess and manage the risk introduced by the

missed surveillance will further minimize possible concerns. Thus,

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

kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant

Reduction in the Margin of Safety

The extended time allowed to perform a missed surveillance does

not result in a significant reduction in the margin of safety. As

supported by the historical data, the likely outcome of any

surveillance is verification that the LCO [Limiting Condition for

Operation] is met. Failure to perform a surveillance within the

prescribed frequency does not cause equipment to become inoperable.

The only effect of the additional time allowed to perform a missed

surveillance on the margin of safety is the extension of the time

until inoperable equipment is discovered to be inoperable by the

missed surveillance. However, given the rare occurrence of

inoperable equipment, and the rare occurrence of a missed

surveillance, a missed surveillance on inoperable equipment would be

very unlikely. This must be balanced against the real risk of

manipulating the plant equipment or condition to perform the missed

surveillance. In addition, parallel trains and alternate equipment

are typically available to perform the safety function of the

equipment not tested. Thus, there is confidence that the equipment

can perform its assumed safety function.

Therefore, this change does not involve a significant reduction

in a margin of safety.

Based upon the reasoning presented above and the previous

discussion of the amendment request, the requested change does not

involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005.

NRC Section Chief: Robert A. Gramm.

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

No. 2, Pope County, Arkansas

Date of amendment request: May 14, 2002, as supplemented by letter

dated September 9, 2002. The May 14, 2002, application was originally

noticed in the Federal Register on July 23, 2002 (67 FR 48216).

Description of amendment request: The proposed amendment would

revise Surveillance Requirement (SR) 4.0.3 to extend the delay period,

before entering a Limiting Condition for Operation, following a missed

surveillance. The delay period would be extended from the current limit

of ``* * * up to 24 hours to permit the completion of the surveillance

when the allowable outage time limits of the ACTION requirements are

less than 24 hours'' to ``* * *up to 24 hours or up to the limit of the

specified interval, whichever is greater.'' In addition, the following

requirement would be added to SR 4.0.3: ``A risk evaluation shall be

performed for any Surveillance delayed greater than 24 hours and the

risk impact shall be managed.'' Also, the addition of a Bases Control

Program is proposed as Technical Specification (TS) 6.5.14,

clarifications are proposed for SR 4.0.1, and other minor changes are

proposed for SR 4.0.3, consistent with NUREG-1432, Revision 2,

``Standard Technical Specifications, Combustion Engineering Plants.''

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice

of opportunity for comment in the Federal Register on June 14, 2001 (66

FR 32400), on possible amendments concerning missed surveillances,

including a model safety evaluation and model no significant hazards

consideration (NSHC) determination, using the consolidated line item

improvement process. The NRC staff subsequently issued a notice of

availability of the models for referencing in license amendment

applications in the Federal Register on September 28, 2001 (66 FR

49714). The licensee affirmed the applicability of the model NSHC

determination in its application dated May 14, 2002, as supplemented by

letter dated September 9, 2002. The NRC staff has augmented the model

NSHC to address the ANO-2 plant-specific items regarding the addition

of a Bases Control Program, clarifications for SR 4.0.1, and other

minor changes for SR 4.0.3 (because the model NSHC assumes a plant's

TSs already have these improvements), as presented below.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), an analysis of the issue

of no significant hazards consideration is presented below:

[[Page 61681]]

Criterion 1--The Proposed Change Does Not Involve a Significant

Increase in the Probability or Consequences of an Accident Previously

Evaluated

The proposed change relaxes the time allowed to perform a missed

surveillance. The time between surveillances is not an initiator of

any accident previously evaluated. Consequently, the probability of

an accident previously evaluated is not significantly increased. The

equipment being tested is still required to be operable and capable

of performing the accident mitigation functions assumed in the

accident analysis. As a result, the consequences of any accident

previously evaluated are not significantly affected. Any reduction

in confidence that a standby system might fail to perform its safety

function due to a missed surveillance is small and would not, in the

absence of other unrelated failures, lead to an increase in

consequences beyond those estimated by existing analyses. The

addition of a requirement to assess and manage the risk introduced

by the missed surveillance will further minimize possible concerns.

The addition of a Bases Control Program formalizes a means for

processing changes to the Bases of the TSs and does not change the

meaning of any TS. The clarifications proposed for SR 4.0.1

regarding surveillances that are not met, do not change the current

intent or practice of the TSs. The other minor changes to SR 4.0.3

regarding the discovery of surveillances that were not performed,

address the delay time period and make other editorial changes that

do not change the current intent or practice of the TSs. As such,

none of these changes affects the initiator of any accident

previously evaluated nor the ability of safety systems to mitigate

any accident previously evaluated.

Therefore, the changes discussed above do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a

New or Different Kind of Accident From Any Previously Evaluated

The proposed change does not involve a physical alteration of

the plant (no new or different type of equipment will be installed)

or a change in the methods governing normal plant operation. A

missed surveillance will not, in and of itself, introduce new

failure modes or effects and any increased chance that a standby

system might fail to perform its safety function due to a missed

surveillance would not, in the absence of other unrelated failures,

lead to an accident beyond those previously evaluated. The addition

of a requirement to assess and manage the risk introduced by the

missed surveillance will further minimize possible concerns.

Likewise, formalizing a program to control changes to the Bases,

clarifying SR 4.0.1, and the other minor changes to SR 4.0.3, do not

change the meaning of any TS and thus do not involve a physical

alteration of the plant or change the methods governing normal plant

operation.

Therefore, the changes discussed above do not create the

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

previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant

Reduction in the Margin of Safety

The extended time allowed to perform a missed surveillance does

not result in a significant reduction in the margin of safety. As

supported by the historical data, the likely outcome of any

surveillance is verification that the LCO [Limiting Condition for

Operation] is met. Failure to perform a surveillance within the

prescribed frequency does not cause equipment to become inoperable.

The only effect of the additional time allowed to perform a missed

surveillance on the margin of safety is the extension of the time

until inoperable equipment is discovered to be inoperable by the

missed surveillance. However, given the rare occurrence of

inoperable equipment, and the rare occurrence of a missed

surveillance, a missed surveillance on inoperable equipment would be

very unlikely. This must be balanced against the real risk of

manipulating the plant equipment or condition to perform the missed

surveillance. In addition, parallel trains and alternate equipment

are typically available to perform the safety function of the

equipment not tested. Thus, there is confidence that the equipment

can perform its assumed safety function.

Likewise, formalizing a program to control changes to the Bases,

clarifying SR 4.0.1, and the other minor changes to SR 4.0.3, do not

change the meaning of any TS and thus will not cause equipment that

is relied upon to perform a safety function, to become inoperable.

Therefore, the changes discussed above do not involve a

significant reduction in a margin of safety.

Based upon the reasoning presented above and the previous

discussion of the amendment request, the requested change does not

involve a significant hazards consideration.

The NRC staff has reviewed the above 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.

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

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

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of amendment request: August 7, 2002.

Description of amendment request: The proposed amendment would

revise the Limiting Condition for Operation (LCO), the associated

Conditions and Required Actions of TS 3.7.1, and the values in Table

3.7.1-1. The proposed changes would revise the LCO by requiring five

MSSVs per steam generator to be operable consistent with the accident

analyses assumptions. The proposed change would modify the associated

Required Actions of TS 3.7.1 by adding a requirement to reduce the

Power Range Neutron Flux--High reactor trip setpoint when one or more

steam generators with one or more MSSVs are inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change adds a requirement to appropriately reduce

the Power Range Neutron Flux--High reactor trip setpoint when one or

more steam generators with one or more MSSVs are inoperable. The

proposed TS change does not affect the design of the MSSV or

increase the likelihood of MSSV failures. Reducing the Power Range

Neutron Flux--High reactor trip setpoint does not affect initiators

of any accident sequence analyzed in the Byron/Braidwood Stations'

Updated Final Safety Analysis Report (UFSAR). Therefore, the

probability of occurrence of a previously evaluated accident is not

increased.

The design basis for the MSSVs is to limit the secondary system

pressure to 1, the ISP will continue to adequately

monitor the RPV materials; therefore, the possibility of an RPV

embrittlement event is not created.

3. Does the proposed change involve a significant decrease in

the margin of safety.

[[Page 61685]]

The ISP will use materials that adequately represent a

particular RPV, including Plant Hatch. A test matrix, as provided in

BWRVIP-86: [``]BWR Vessel and Internals Project, BWR Integrated

Surveillance Program Implementation Plan,'' includes representative

materials from other plants to be used for the Hatch Units. A

representative material is a plate or weld that is selected from

among all the existing plant surveillance programs to represent the

corresponding limiting plate or weld material in a plant. The choice

of material considers chemistry, heat number, fabricator and the

welding process. These are factors that determine the best

representative material. As a result, the Hatch RPV will be

adequately monitored for radiation embrittlement and 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

amendment request involves no significant hazards consideration.

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

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

20037.

NRC Section Chief: John A. Nakoski.

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

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

Date of amendment request: August 19, 2002.

Description of amendment request: The proposed amendment revises

Technical Specification (TS) Section 3/4.3.2, ``Engineered Safety

Features Actuation System Instrumentation.''

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:

Pursuant to 10 CFR 50.91, this analysis provides a determination

that the proposed change to the Technical Specifications described

previously, does not involve any significant hazards consideration

as defined in 10 CFR 50.92, as described below:

[(1)] Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Response: No.

This change to the Technical Specifications will not result in a

condition where the design, material, and construction standards

that were applicable prior to the change are altered. The same ESFAS

[engineered safety features actuation system] instrumentation will

be used and the same ESFAS system reliability is expected. The

proposed change will not modify any system interface or function and

could not increase the likelihood of an accident because these

events are independent of this change. The proposed activity will

not change, degrade, or alter any assumptions previously made in

evaluating the radiological consequences of an accident described in

the safety analysis report.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

[(2)] Does the proposed change create the possibility of a new

or different kind of accident from any accident previously

evaluated?

Response: No.

The proposed change will not alter the performance of the ESFAS

mitigation systems assumed in the plant safety analysis. Changing

the interval for periodically verifying ESFAS slave relays (assuring

equipment operability) will not create any new accident initiators

or scenarios. Only the testing frequency is changed. No physical

changes will be made to the Solid State Protection System or the ESF

Actuation System as a result of this change.

Therefore, the proposed change does not create the possibility

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

evaluated.

[(3)] Does the proposed change involve a significant reduction

in a margin of safety?

Response: No.

The proposed change will not affect the total ESFAS response

assumed in the safety analysis because the reliability of the slave

relays will not be significantly affected by the increased

surveillance interval. The relays have demonstrated a high

reliability and insensitivity to short term wear and aging effects.

The overall reliability, redundancy, and diversity assumed available

for the protection and mitigation of accident and transient

conditions is unaffected by this proposed Technical Specification

change.

Therefore, the proposed change does not involve a reduction in a

margin of safety.

Based on the above safety evaluation, the South Texas Project

concludes that the change proposed by this License Amendment Request

satisfies the no significant hazards consideration standards of 10

CFR 50.92(c) and, accordingly, a finding of no significant hazards

is justified.

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.

Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, &

Bockius, 1111 Pennsylvania NW., Washington, DC 20004.

NRC Section Chief: Robert A. Gramm.

STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,

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

Date of amendment request: August 20, 2002.

Description of amendment request: The proposed amendment would

delete the Appendix C of the Operating License, regarding antitrust

conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

STPNOC has determined whether a significant hazards

consideration is involved with the proposed amendment by focusing on

the three criteria set forth in 10 CFR 50.92 as discussed below:

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Response: No.

This request involves an administrative change only. The

Operating Licenses are being changed to remove unnecessary and

outdated antitrust conditions. No actual plant equipment or accident

analyses will be affected by the proposed changes. Therefore, this

request will have no impact on the probability or consequences of

any type of accident: new, different, or previously evaluated.

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

different kind of accident from any accident previously evaluated?

Response: No.

This request involves an administrative change only. The

Operating Licenses are being changed to remove unnecessary and

outdated antitrust conditions. No actual plant equipment or accident

analyses will be affected by the proposed change and no failure

modes not bounded by previously evaluated accidents will be created.

Therefore, this request will have no impact on the possibility of

any type of accident: new, different, or previously evaluated.

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

margin of safety?

Response: No.

Margin of safety is associated with confidence in the ability of

the fission product barriers (i.e., fuel and fuel cladding, Reactor

Coolant System pressure boundary, and containment structure) to

limit the level of radiation dose to the public. This request

involves an administrative change only. The Operating Licenses are

being changed to remove unnecessary and outdated antitrust

conditions.

No actual plant equipment or accident analyses will be affected

by the proposed change. Additionally, the proposed change will not

relax any criteria used to establish safety limits, safety systems

settings, or any limiting conditions of operations. Therefore, this

request will not impact [a] margin of safety.

Based on the above, STPNOC concludes that the proposed amendment

involves no significant hazards consideration under the criteria set

forth in 10 CFR 50.92 and, accordingly, a finding of ``no

significant hazards consideration'' is justified.

[[Page 61686]]

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.

Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis, &

Bockius, 1111 Pennsylvania NW., Washington, DC 20004.

NRC Section Chief: Robert A. Gramm.

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

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

Date of amendment request: August 21, 2002.

Description of amendment request: The proposed amendment revises

Technical Specifications (TS) 3/4.4.1.4.2 and 3/4.9.1.3 to delete the

specific reference to the valves required to be secured to isolate

uncontrolled boron dilution flow paths in MODE 5 with the loops not

filled and in MODE 6.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

STPNOC has evaluated whether or not a significant hazards

consideration is involved with the proposed amendment by focusing on

the three standards set forth in 10 CFR 50.92, ``Issuance of

amendment,'' as discussed below.

[(1)] Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

Response: No.

There is no technical change in the requirements imposed by the

Technical Specifications. The proposed changes to replace the TS

reference to the specific valves to be used to isolate boron

dilution flow paths with new Technical Specification requirements to

assure the flow paths are secured provides the same level of

assurance that the boron dilution event will be precluded.

[(2)] Does the proposed change create the possibility of a new

or different kind of accident from any accident previously

evaluated?

Response: No.

The proposed change allows alternate, equally effective,

locations where the potential boron dilution flow paths can be

isolated to preclude an uncontrolled boron dilution event in MODE 5

with the loops not filled and in MODE 6. Consequently, the

possibility of the dilution event is unchanged. The proposed change

does not otherwise alter how the plant is operated or change its

design basis so that the possibility of a new accident is not

created.

[(3)] Does the proposed change involve a significant reduction

in a margin of safety?

Response: No.

The proposed changes to replace the TS reference to the specific

valves to be used to isolate boron dilution flow paths with new

Technical Specification requirements to assure the flow paths are

secured provides the same level of assurance that the boron dilution

event will be precluded.

Based upon the analysis provided herein, the proposed amendments

do not involve a significant hazards consideration.

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.

Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, &

Bockius, 1111 Pennsylvania NW., Washington, DC 20004.

NRC Section Chief: Robert A. Gramm.

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

Generating Station, Coffey County, Kansas

Date of amendment request: August 16, 2002.

Description of amendment request: The amendment would revise

Technical Specification 3.6.3, ``Containment Isolation Valves,'' by (1)

deleting the Note and adding the acronym ``(CIV)'' for containment

isolation valve in Condition A of the Actions for the Limiting

Condition for Operation, (2) revising the Completion Time for Required

Condition A.1 from 4 hours to as much as 7 days depending on the

category of the CIVs, (3) deleting Condition C, and (4) renumbering the

later Conditions D and E. The proposed amendment is based on Topical

Report WCAP-15791-P, ``Risk-Informed Evaluation of Extensions to

Containment Isolation Valve Completion Times.''

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 changes to the Completion Times do not change the

response of the plant to any accidents and have an insignificant

impact on the reliability of the containment isolation valves. The

containment isolation valves will remain highly reliable and the

proposed changes will not result in a significant increase in the

risk of plant operation. This is demonstrated by showing that the

impact on plant safety as measured by the large early release

frequency (LERF) and incremental conditional large early release

probabilities (ICLERP) is acceptable. These changes are consistent

with the acceptance criteria in [the risk-informed] Regulatory

Guides 1.174 and 1.177. Therefore, since the containment isolation

valves will continue to perform their [safety] functions with high

reliability as originally assumed and the increase in risk as

measured by LERF and ICLERP is acceptable, there will not be a

significant increase in the consequences of any accidents.

The proposed changes do not adversely affect accident initiators

or precursors nor alter the design assumptions, conditions, or

configuration of the facility or the manner in which the plant is

operated and maintained. The proposed changes do not alter or

prevent the ability of structures, systems, and components (SSCs)

from performing their intended [safety] function to mitigate the

consequences of an initiating event within the assumed acceptance

limits. The proposed changes do not affect the source term,

containment isolation, or radiological release assumptions used in

evaluating the radiological consequences of an accident previously

evaluated. Further, the proposed changes do not increase the types

or amounts of radioactive effluent that may be released offsite, nor

significantly increase individual or cumulative occupational/public

radiation exposures. The proposed changes are consistent with the

safety analysis assumptions and resultant consequences [in Chapter

15, ``Accident Analysis,'' of the Updated Final Safety Analysis

Report (USAR) for the plant].

Therefore, it is concluded that this change does not increase

the probability of occurrence of a malfunction of equipment

important to safety.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not result in a change in the manner in

which the containment isolation valves provide plant protection.

There are no design changes associated with the proposed changes.

The changes to Completion Times do not change any existing accident

scenarios, nor create any new or different accident scenarios.

The changes do not involve a physical alteration of the plant

(i.e., no new or different type of equipment will be installed) or a

change in the methods governing normal plant operation. In addition,

the changes do not impose any new or different requirements or

eliminate any existing requirements. The changes do not alter

assumptions made in the safety analysis. The proposed changes are

consistent with the safety analysis assumptions and current plant

operating practice.

Therefore, the possibility of a new or different malfunction of

safety related equipment is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not alter the manner in which safety

limits, limiting safety system settings or limiting conditions for

[[Page 61687]]

operation are determined. The safety analysis acceptance criteria

are not impacted by these changes. The proposed changes will not

result in plant operation in a configuration outside the design

basis. The calculated impact on risk is insignificant and is

consistent with the acceptance criteria contained in Regulatory

Guides 1.174 and 1.177.

Therefore, it is concluded that 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.

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

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

NRC Section Chief: Stephen Dembek.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power

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

Date of amendment request: July 19, 2002.

Brief description of amendment request: The proposed amendment

would revise Technical Specification Surveillance Requirement (SR)

4.0.3 to extend the delay period, before entering a Limiting Condition

for Operation, following a missed surveillance. The delay period would

be extended from the current limit of ``* * * up to 24 hours'' to ``* *

* up to 24 hours or up to the limit of the specified surveillance

interval, whichever is greater.'' In addition, the following

requirement would be added to SR 4.0.3: ``A risk evaluation shall be

performed for any surveillance delayed greater than 24 hours and the

risk impact shall be managed.'' The proposed amendment would also make

administrative changes to SRs 4.01 and 4.03 to be consistent with

NUREG-1431, Revision 2.

Date of publication of individual notice in Federal Register:

September 4, 2002 (67 FR 56604).

Expiration date of individual notice: October 4, 2002.

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, located at One White Flint

North, 11555 Rockville Pike (first floor), Rockville, Maryland.

Publicly available records will be accessible from the Agencywide

Documents Access and Management Systems (ADAMS) Public Electronic

Reading Room on the internet at the NRC Web site, http://www.nrc.gov/

reading-rm/adams.html. If you do not have access to ADAMS or if there

are problems in accessing the documents located in ADAMS, contact the

NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-

415-4737 or by e-mail to [email protected].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: September 19, 2001, as

supplemented on January 17 and July 1, 2002.

Brief description of amendment: The amendment revises Technical

Specifications Subsections 3.5.A.5.b and c, concerning operability of

suppression chamber-to-drywell vacuum breakers.

Date of Issuance: September 11, 2002.

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

within 30 days of issuance.

Amendment No.: 230.

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

Technical Specifications.

Date of initial notice in Federal Register: December 20, 2001 (66

FR 65749). The January 17 and July 1, 2002, letters provided clarifying

information within the scope of the original application and did not

change the staff's initial proposed no significant hazards

consideration determination. The Commission's related evaluation of

this amendment is contained in a Safety Evaluation dated September 11,

2002.

No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: September 10, 2001.

Brief description of amendment: The amendment revised the

requirements in Technical Specifications, Sections 3.4.A.7.c and

3.4.A.8.c, changing confirmation of operability of core spray pumps and

system components from testing to verification.

Date of Issuance: September 10, 2002.

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

within 30 days of issuance.

Amendment No.: 231.

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

Technical Specifications.

Date of initial notice in Federal Register: March 5, 2002 (67 FR

10008). The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated September 10, 2002.

No significant hazards consideration comments received: No.

[[Page 61688]]

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: August 1, 2001, as supplemented

on June 19 and September 9, 2002.

Brief description of amendment: The amendment revised Technical

Specifications Section 6.3, ``Facility Staff Qualifications,'' deletes

Section 6.4, ``Training,'' and revises the Table of Contents to reflect

deletion of Section 6.4. These changes reflect updating of requirements

that had been outdated based on licensed operator training programs

being accredited by the Institute of Nuclear Power Operations, and

promulgation of applicable regulations.

Date of Issuance: September 18, 2002.

Effective date: September 18, 2002, and shall be implemented within

30 days of issuance.

Amendment No.: 232.

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

Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR

55009). The June 19 and September 9, 2002, letters provided clarifying

information within the scope of the original application and did not

change the staff?s initial proposed no significant hazards

consideration determination. The Commission's related evaluation of

this amendment is contained in a Safety Evaluation dated September 18,

2002.

No significant hazards consideration comments received: No.

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

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

1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: May 15, 2002, as supplemented

by letter dated August 29, 2002.

Brief description of amendments: The amendments revise Limiting

Condition for Operation (LCO) 3.9.3, ``Containment Penetrations.'' The

amendments would (1) modify the requirement in LCO 3.9.3.b that one

door in each air lock is closed by adding the words ``capable of

being'' before the word ``closed'' and (2) add a note to LCO 3.9.3

stating that containment penetration flow paths providing direct access

from the containment to the outside atmosphere may be unisolated under

administrative controls. The amendments would allow the containment air

lock and other penetrations that provide direct access to the outside

atmosphere to be open during core alterations or movement of irradiated

fuel assemblies within containment.

Date of issuance: September 11, 2002.

Effective date: September 11, 2002, and shall be implemented within

60 days of the date of issuance, including completing the changes to

the Technical Specification Bases, as described in the licensee's

letters of May 15 and August 29, 2002.

Amendment Nos.: Unit 1--144, Unit 2--144, Unit 3--144.

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

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 25, 2002 (67 FR

42816). The Commission's related evaluation of the amendments are

contained in a Safety Evaluation dated September 11, 2002.

No significant hazards consideration comments received: No.

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: March 26, 2002, as supplemented

June 19 and August 8, 2002.

Brief description of amendment: This amendment extends the 10-year

performance-based Type A test interval on a one-time basis to require

the performance of a Type A test within 12.1 years from the last test,

which was performed on April 9, 1992.

Date of issuance: September 16, 2002.

Effective date: September 16, 2002.

Amendment No.: 193.

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

Technical Specifications.

Date of initial notice in Federal Register: May 28, 2002 (67 FR

36928). The June 19, and August 8, 2002, supplements contained

clarifying information only, and did not change the initial proposed no

significant hazards consideration determination or expand the scope of

the initial application. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 16, 2002.

No significant hazards consideration comments received: No.

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: February 21, 2002, as

supplemented May 14 and August 2, 2002.

Brief description of amendment: The amendment modifies the

containment vessel spray nozzle testing frequency from testing every

``10 years'' to testing ``following activities which could result in

nozzle blockage.''

Date of issuance: September 19, 2002.

Effective date: September 19, 2002.

Amendment No.: 194.

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

Technical Specifications.

Date of initial notice in Federal Register: April 30, 2002 (67 FR

21285). The May 14 and August 2, 2002, supplements contained clarifying

information only and did not change the initial proposed no significant

hazards consideration determination or expand the scope of the initial

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

contained in a Safety Evaluation dated September 19, 2002.

No significant hazards consideration comments received: No.

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

Nuclear Plant, Unit 1, Wake and Chatham Counties, North Carolina

Date of application for amendment: July 8, 2002.

Brief Description of amendment: The amendment deleted the level

value in Technical Specification (TS) 3/4.8.1.1, ``Electrical Power

Systems--A.C. Sources--Operating'' and TS 3/4.8.1.2, ``Electrical Power

Systems--A.C. Sources--Shutdown.''

Date of issuance: September 12, 2002.

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

within 60 days from date of issuance.

Amendment No.: 111.

Facility Operating License No. NPF-63: Amendment changes the

Technical Specifications.

Date of initial notice in Federal Register: August 6, 2002 (67 FR

50950). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 12, 2002.

No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear

Plant, Charlevoix County, Michigan

Date of amendment request: June 11, 2002, as supplemented by letter

dated July 3, 2002.

Brief description of amendment: The amendment revises Defueled

Technical Specification (DTS) Section 5.2, ``Storage and Inspection of

Spent Fuel,'' and DTS Section 6.6.2.9, ``Spent Fuel Pool Water

Chemistry Program,'' by adding applicability statements that specify

that these specifications apply

[[Page 61689]]

only when irradiated fuel is stored in the spent fuel pool.

Date of issuance: September 11, 2002.

Effective date: The license amendment is effective as of the date

of issuance and shall be implemented within 45 days from the date of

issuance.

Amendment No.: 124.

Facility Operating License No. DPR-6: The amendment revised the

Defueled Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR

45562). The July 3, 2002, supplemental letter provided clarifying

information that did not change the scope of the original Federal

Register notice or the original no significant hazards consideration

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

contained in a Safety Evaluation dated September 11, 2002.

No significant hazards considerations comments received: No.

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

Michigan

Date of application for amendment: May 23, 2002.

Brief description of amendment: The amendment deletes Technical

Specification 5.5.3, ``Post Accident Sampling System (PASS),'' and

thereby eliminates the requirements to have and maintain the PASS at

Fermi 2.

Date of issuance: September 5, 2002.

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

within 90 days.

Amendment No.: 150.

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

Technical Specifications.

Date of initial notice in Federal Register: June 25, 2002 (67 FR

42816). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,

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

Date of application for amendment: June 6, 1998; April 5, 1999;

April 7, April 19, July 31, and September 28, 2000; March 19, June 11,

September 21, and December 20, 2001.

Brief description of amendment: The amendment revises the Millstone

Power Station, Unit No. 3 licensing basis related to operation of the

supplementary leak collection and release system after a postulated

accident. Specifically, the proposed revision to the Final Safety

Analysis Report (FSAR) would address: (1) The manual actions required

to trip the non-safety grade fans and the time requirements for control

room ventilation realignment, and (2) the input assumptions and results

of the loss-of-coolant accident/control rod ejection accident analyses.

Date of issuance: September 16, 2002.

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

within 60 days from the date of issuance.

Amendment No.: 211.

Facility Operating License No. NPF-49: Amendment revised the FSAR.

Date of initial notice in Federal Register: July 1, 1998 (63 FR

35992). The April 5, 1999; April 7, April 19, July 31, and September

28, 2001; March 19, June 11, September 21, and December 20, 2001,

letters provide clarifying information that was within the scope of the

original application and did not change the staff's proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated September 16, 2002.

No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: December 7, 2001, as

supplemented by letter dated July 22, 2002.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) to permit implementation of containment

local leakage rate testing addressed by 10 CFR Part 50, Appendix J,

Option B, and to reference Regulatory Guide 1.163, ``Performance-Based

Containment Leak Test Program,'' dated September 1995. In addition, the

TS are revised regarding soap bubble testing and leak testing of

containment purge valves with resilient seals for upper and lower

compartments and instrument rooms.

Date of issuance: September 4, 2002.

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

within 30 days from the date of issuance.

Amendment Nos.: 207 & 188.

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 26, 2001 (67

FR 66464). The supplement dated July 22, 2002, provided clarifying

information that did not change the scope of the December 7, 2001,

application nor the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No.

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

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

Date of application of amendments: July 11, 2002.

Brief description of amendments: The amendments revised the

Technical Specifications to incorporate several administrative changes.

Date of Issuance: September 5, 2002.

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

within 30 days from the date of issuance.

Amendment Nos.: 328, 328 & 329.

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

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 6, 2002 (67 FR

50951). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear

Generating Unit No. 2, Westchester County, New York

Date of application for amendment: January 8, 2002, as supplemented

on August 22, 2002.

Brief description of amendment: The amendment revised Technical

Specifications Section 3.7.C, ``Gas Turbine Generators,'' and Section

4.6, ``Emergency Power System Periodic Tests,'' to change the minimum

amount of fuel oil required to be stored from 54,200 gallons to 94,870

gallons. The amendment also revised the minimum electrical output of

the gas turbine generator that is required to be tested monthly to 2000

kilowatts from the previous value of 750 kilowatts.

Date of issuance: September 18, 2002.

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

60 days.

Amendment No.: 233.

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

Technical Specifications.

Date of initial notice in Federal Register: March 5, 2002 (67 FR

10012). The August 22, 2002, letter provided clarifying information

that did not

[[Page 61690]]

change the initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated September 18, 2002.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: June 7, 2002, supplemented July

17, 2002.

Brief description of amendment: The amendment changes the Technical

Specifications to allow relaxation of secondary containment operability

requirements while handling irradiated fuel in the secondary

containment. The amendment replaces the current accident source term

use in selected design basis radiological analyses with an alternative

source term pursuant to 10 CFR 50.67, ``Accident Source Term.''

Date of issuance: September 12, 2002.

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

30 days.

Amendment No.: 276.

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

Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR

45568). The July 17, 2002, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 12, 2002.

No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,

Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,

Vermont

Date of application for amendment: March 19, 2002, as supplemented

on June 4, July 16 and 24, August 22 and September 4, 2002.

Brief description of amendment: The amendment revises the technical

specifications to reflect the removal of the automatic reactor scram

and main steam isolation valve closure functions of the main steam line

radiation monitors (MSLRM). An explicit requirement for periodic

functional test and calibration of the MSLRM is added to maintain

operability of the mechanical vacuum pump trip function.

Date of Issuance: September 18, 2002.

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

implemented within 60 days.

Amendment No.: 212.

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

Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR

45573). The July 16 and 24, August 22, and September 4, 2002,

supplements were within the scope of the original application and did

not change the staff's proposed no significant hazards consideration

determination. The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated September 18, 2002.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station (GGNS), Unit 1, Claiborne

County, Mississippi

Date of application for amendment: November 15, 2001, as

supplemented by letters dated March 1 and June 19, 2002.

Brief description of amendment: This amendment revises the GGNS

Unit 1 Technical Specification Surveillance Requirements (SRs)

pertaining to testing of the standby emergency diesel generators (DGs)

to allow DG testing during reactor operation. The change removes the

restriction associated with these SRs that prohibits conducting the

required testing of the DGs during reactor operating Modes 1, 2, or 3.

Date of issuance: September 5, 2002.

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

within 60 days from the date of issuance.

Amendment No: 153.

Facility Operating License No. NPF-29: The amendment revises the

Technical Specifications and Surveillance Requirements.

Date of initial notice in Federal Register: December 26, 2001 (66

FR 66464). The supplemental letters dated March 1 and June 19, 2002,

provided clarifying information that did not change the scope of

original Federal Register notice or the original no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: None.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: February 25, 2002, as

supplemented by letters dated August 16 and 22, 2002.

Brief description of amendment: This amendment adds a new Technical

Specification 3.10.9, ``Suppression Pool Makeup-MODE 3,'' to allow

installation of reactor cavity gate 2 in the Upper Containment Pool

(UCP) and draining the reactor cavity pool portion of the UCP while

still in MODE 3, with the reactor pressure less than 230 pounds per

square inch gauge (psig). It also modifies the applicability of the UCP

gates surveillance requirement (TS Section 3.6.2.4, ``Suppression Pool

Makeup (SPMU) System,'') to allow installation of UCP gates in MODES 1,

2, and 3.

Date of issuance: September 6, 2002.

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

within 60 days of issuance.

Amendment No: 154.

Facility Operating License No. NPF-29: The amendment revises the

Technical Specifications and Surveillance Requirements.

Date of initial notice in Federal Register: April 30, 2002 (67 FR

21289). The August 16 and 22, 2002, supplemental letters provided

clarifying information that did not change the scope of the original

Federal Register notice or the original no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: February 19, 2002, as

supplemented by letter dated July 17, 2002.

Brief description of amendment: This amendment revises Technical

Specification 3.8.1, ``AC Sources--Operating,'' to remove all current

Mode restrictions associated with testing the High Pressure Core Spray

Diesel Generator 13 during normal operation. The proposed changes

remove the restriction associated with Surveillance Requirements (SRs)

that prohibit performing the required testing in

[[Page 61691]]

Modes 1, 2, or 3. The specific SRs addressed in this amendment are: SR

3.8.1.11, 3.8.1.12, 3.8.1.16, and 3.8.1.19.

Date of issuance: September 10, 2002.

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

within 60 days from the date of issuance.

Amendment No: 155.

Facility Operating License No. NPF-29: The amendment revises the

Technical Specifications and Surveillance Requirements.

Date of initial notice in Federal Register: April 30, 2002 (67 FR

21288). The supplemental letter dated July 17, 2002, provided

clarifying information that did not change the scope of original

Federal Register notice or the original no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 10, 2002.

No significant hazards consideration comments received: None.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of application for amendments: March 8, 2002.

Brief description of amendments: The amendments revise TS 3.8.4,

``DC Sources-Operating,'' 3.8.5, ``DC Sources-Shutdown,'' 3.8.6,

``Battery Cell Parameters,'' and 3.8.8, ``Inverter-Shutdown.'' The

changes also include the relocation of the following TS items to a

licensee-controlled program: (1) A number of Surveillance Requirements

(SRs) that require the performance of preventive maintenance, and (2)

TS Table 3.8.6-1, ``Battery Cell Parameter Requirements.'' The

amendments also add new actions and their associated completion times

to TS 3.8.6 for out-of-limits conditions for battery cell voltage,

electrolyte level, and electrolyte temperature. In addition, SRs are

added for verification of these parameters.

Date of issuance: September 19, 2002.

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

within 60 days.

Amendment Nos.: 129, 129, 124 & 124.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: May 14, 2002 (67 FR

34485). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 19, 2002.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden

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

Date of application for amendments: August 1, 2001, as supplemented

June 19 and September 9, 2002.

Brief description of amendments: The amendments revise Technical

Specification 5.3, ``Unit Staff Qualifications,'' concerning approval

of the education and experience eligibility requirements for operator

license applicants.

Date of issuance: September 17, 2002.

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

within 60 days.

Amendment Nos.: 194 & 187.

Facility Operating License Nos. DPR-19 and DPR-25: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR

55018). The supplements dated June 19 and September 9, 2002, provided

additional information that clarified the application, did not expand

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

the staff's original proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle

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

Date of application for amendments: August 01, 2001, as

supplemented June 19 and September 09, 2002.

Brief description of amendments: The amendments revise Technical

Specifications requirements regarding Facility Staff Qualifications for

licensed operator and non-licensed personnel training programs. The

changes revise requirements that have been superseded based on licensed

operator training programs being accredited by the Institute of Nuclear

Power Operations, promulgation of the revised 10 CFR part 55,

``Operators' Licenses,'' which became effective on May 26, 1987, and

adoption of a systems approach to training as required by 10 CFR

50.120, ``Training and qualification of nuclear power plant

personnel.''

Date of issuance: September 17, 2002.

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

within 60 days.

Amendment Nos.: 154 & 140.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR

55018). The supplements dated June 19 and September 09, 2002, provided

additional information that clarified the application, did not expand

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

the staff's original proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-

277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York

County, Pennsylvania

Date of application for amendments: August 1, 2001, as supplemented

June 19 and September 9, 2002.

Brief description of amendments: The amendments revised Technical

Specification 5.3.1 to state that the licensed operators shall comply

with the qualification requirements in 10 CFR part 55, rather than the

American National Standards Institute's (ANSI) standard ANSI N18.1-

1971.

Date of issuance: September 17, 2002.

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

implemented within 30 days.

Amendments Nos.: 245, 249.

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR

55018). The June 19 and September 9, 2002, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination or expand the application beyond

the scope of the original Federal Register notice. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois

Date of application for amendments: August 1, 2001, as supplemented

June 19 and September 9, 2002.

[[Page 61692]]

Brief description of amendments: The amendments revise Technical

Specification requirements that have been superceded based on the

licensed operator training program being accredited by the Institute of

Nuclear Power Operations, promulgation of the revised 10 CFR part 55,

and adoption of a systems approach to training as required by 10 CFR

50.120.

Date of issuance: September 18, 2002.

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

within 60 days.

Amendment Nos.: 208 & 203.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 31, 2001 (66 FR

55018). The supplements dated June 19 and September 9, 2002, provided

additional information that clarified the application, did not expand

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

the staff's original proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 18, 2002.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: November 9, 2000.

Brief description of amendment: This amendment revises the allowed

outage time from 72 hours to 7 days for one low pressure injection

train, and one containment spray system train. The supporting analysis

for the request is based on the Babcock & Wilcox Owners Group (B&WOG)

Topical Report BAW-2295A, Revision 1 & 2, ``Justification for the

Extension of Allowed Outage Time for Low pressure Injection and Reactor

Building Spray Systems,'' and its review by the staff documented in a

Safety Evaluation Report. The Davis-Besse Nuclear Power Station is the

lead B&WOG plant requesting these changes to be made to the Technical

Specifications.

Date of issuance: September 17, 2002.

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

within 120 days.

Amendment No.: 253.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 27, 2000 (65

FR 81919). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 17, 2002.

No significant hazards consideration comments received: No.

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

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

Date of application for amendments: January 18, 2002.

Brief description of amendments: These amendments revised Technical

Specifications to relocate specific working hour limits and controls to

administrative procedures.

Date of issuance: September 10, 2002.

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

within 60 days of issuance.

Amendment Nos.: 185 and 128.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67

FR 7418). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 10, 2002.

No significant hazards consideration comments received: No.

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

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: July 26, 2002, as supplemented

August 23, 2002

Brief description of amendments: The amendments will add a license

condition to the Operating Licenses for both units, allowing a one-time

140-hour allowed outage time for the essential service water (ESW)

system, to allow ESW pump replacement during plant operation.

Date of issuance: September 9, 2002.

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

within 20 days.

Amendment Nos.: 270 and 251.

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

revised the Facility Operating License.

Date of initial notice in Federal Register: August 8, 2002 (67 FR

51603). The August 23, 2002, letter provided clarifying information

within the scope of the original application and 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 September 9, 2002.

No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile

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

Date of application for amendment: October 19, 2001, as

supplemented June 17, 2002.

Brief description of amendment: The amendment revised the Technical

Specifications to implement programmatic controls for radiological

effluent technical specifications in the Administrative Controls

section, to relocate certain procedural details to licensee-controlled

documents, and to add new programs to accommodate existing NRC

requirements and guidance.

Date of issuance: September 11, 2002.

Effective date: September 11, 2002.

Amendment No.: 176.

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

Technical Specifications.

Date of initial notice in Federal Register: January 8, 2002 (67 FR

928). The June 17, 2002, supplemental letter did not expand the scope

of the application as originally noticed and did not change the

proposed no significant hazards consideration determination. The

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

Safety Evaluation dated September 11, 2002.

No significant hazards consideration comments received: No.

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

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

Date of amendment request: August 9, 2001, as supplemented

September 17, 2001, and June 24, 2002.

Description of amendment request: The amendment combines Technical

Specifications (TSs) 3/4.9.9, ``Containment Purge and Exhaust Isolation

System,'' and 3/4.9.4, ``Containment Building Penetrations.'' By

combining these two TSs, the amendment updates the Seabrook TSs related

to refueling operations by adopting portions of NUREG-1431, ``Standard

Technical Specifications, Westinghouse Plants,'' Revision 2. The

amendment also changes the TS index pages and the associated TS Bases.

By letter dated June 24, 2002, the licensee withdrew that part of the

application associated with relocation of TS 3/4.9.4, ``Decay Time,''

to the Seabrook Station Technical Requirements Manual.

Date of issuance: September 5, 2002.

Effective date: As of its date of issuance, and shall be

implemented within 90 days.

[[Page 61693]]

Amendment No.: 85.

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

Technical Specifications.

Date of initial notice in Federal Register: September 19, 2001 (66

FR 48290). The supplements dated September 17, 2001, and June 24, 2002,

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination. The

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

Safety Evaluation dated September 5, 2002.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,

Van Buren County, Michigan

Date of application for amendment: January 28, 2002.

Brief description of amendment: The amendment revises the Core

Operating Limits Report analytical methods referenced in Technical

Specification (TS) 5.6.5.b. Specifically, the amendment adds references

to two NRC-approved Framatome ANP, Inc., reports: (1) EMF-2310(P)(A),

Revision 0, ``SRP [Standard Review Plan] Chapter 15 Non-LOCA [loss-of-

coolant accident] Methodology for Pressurized Water Reactors [PWRs],''

dated May 2001, and (2) EMF-2328(P)(A), Revision 0, ``PWR Small Break

LOCA Evaluation Model, S-RELAP5 Based,'' dated March 2001. The

amendment also deletes previous references in TS 5.6.5.b describing

Exxon Nuclear Company's large-break LOCA evaluation model.

Date of issuance: September 13, 2002.

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

within 60 days.

Amendment No.: 209.

Facility Operating License No. DPR-20: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 19, 2002 (67

FR 7420). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 13, 2002.

No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,

Salem County, New Jersey

Date of application for amendment: April 3, 2002.

Brief description of amendment: This amendment consists of changes

to the Technical Specifications (TSs) which allow the relocation of TS

3/4.4.4, ``Reactor Coolant System--Chemistry,'' and the associated

bases from the TSs to the Hope Creek Updated Final Safety Analysis

Report (UFSAR).

Date of issuance: September 18, 2002.

Effective date: September 18, 2002, and shall be implemented within

60 days.

Amendment No.: 140.

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications and the UFSAR.

Date of initial notice in Federal Register: May 14, 2002 (67 FR

34492). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 18, 2002.

No significant hazards consideration comments received: No.

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

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

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

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

Date of application for amendments: May 24, 2002.

Brief description of amendments: The amendments revised the

Technical Specifications to allow Mode 2 (startup) operation with two

out of four, rather than three out of four, required intermediate range

monitor channels per trip system.

Date of issuance: September 12, 2002.

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

within 30 days from the date of issuance.

Amendment Nos.: 233/175.

Renewed Facility Operating License Nos. DPR-57 and NPF-5:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: July 9, 2002 (67 FR

45572). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 12, 2002.

No significant hazards consideration comments received: No.

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

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

Date of amendment request: May 31, 2001, as supplemented by letters

dated June 14, August 13, October 16, November 7, 2001, August 14,

2002, and September 4, 2002.

Brief description of amendments: The amendment grants conforming

amendments to the operating licenses to reflect the direct transfer of

Reliant Energy Incorporated's ownership interest to Texas Genco, LP.

Date of issuance: September 4, 2002.

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

within 30 days from the date of issuance.

Amendment Nos.: Unit 1-142; Unit 2-130.

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

revised the facility operating licenses.

Date of initial notice in Federal Register: September 28, 2001 (66

FR 49711). The supplemental information did not expand the scope of the

application as originally noticed in the Federal Register. The

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

Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: April 1, 2002, as supplemented by letter

dated June 6, 2002.

Brief description of amendments: The amendments include addition of

topical report ERX-2001-005, ``ZIRLO\TM\ Cladding and Boron Coating

Models for TXU Electric's Loss of Coolant Accident Analysis

Methodologies,'' to the list of approved methodologies for use in

generating the Core Operating Limits Report in Technical Specification

(TS) 5.6.5, ``Core Operating Limits Report (COLR).'' In addition, the

proposed changes include ZIRLO\TM\ clad in the description of the fuel

assemblies in TS 4.2.1, ``Fuel Assemblies.''

Date of issuance: September 4, 2002.

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

within 60 days from the date of issuance.

Amendment Nos.: 99 and 99.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 14, 2002 (67 FR

34493). The June 6, 2002, supplemental letter provided clarifying

information that did not change the scope of the original Federal

Register notice or the original no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No.

[[Page 61694]]

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: March 27, 2002.

Brief description of amendments: The amendments revise Technical

Specification (TS) 5.3.1 to require that each member of the unit staff,

with the exception of licensed Reactor Operators (ROs) and licensed

Senior Reactor Operators (SROs), shall meet or exceed the minimum

qualifications of Regulatory Guide (RG) 1.8, ``Qualification and

Training of Personnel for Nuclear Power Plants,'' Revision 2, 1987.

Also, a new TS 5.3.2 is added to require that the ROs and SROs shall

meet or exceed the minimum qualifications of RG 1.8, Revision 3, May

2000, and the current TS 5.3.2 is renumbered to TS 5.3.3.

Date of issuance: September 4, 2002.

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

within 60 days from the date of issuance.

Amendment Nos.: 100 and 1000.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 14, 2002 (67 FR

34493). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated September 4, 2002.

No significant hazards consideration comments received: No.

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

Callaway County, Missouri

Date of application for amendment: June 17, 2002 (ULNRC-04684).

Brief description of amendment: The amendment revised Technical

Specification 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' by

adding Surveillance Requirement (SR) 3.3.1.16 to Function 3 of TS Table

3.3.1-1. SR 3.3.1.16 verifies that the reactor trip system response

times are within limits every 18 months on a staggered test basis.

Date of issuance: September 3, 2002.

Effective date: September 3, 2002, and shall be implemented within

60 days from the date of issuance.

Amendment No.: 151.

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

Technical Specifications.

Date of initial notice in Federal Register: July 23, 2002 (67 FR

48222). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated September 3, 2002.

No significant hazards consideration comments received: No.

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

Callaway County, Missouri

Date of application for amendment: February 15, 2001, as

supplemented by letters dated April 20 and November 7, 2001, and March

1 and August 5, 2002.

Brief description of amendment: The amendment revises paragraph

d.1.j) 2) of Technical Specification (TS) 5.5.9, ``Steam Generator (SG)

Tube Surveillance Program,'' to (1) delete the requirement that all SG

tubes containing an Electrosleeve TM, a Framatome

proprietary process, be removed from service within two operating

cycles following installation of the first ElectrosleeveTM;

(2) add the requirement that ElectrosleevesTM will not be

installed in the outermost periphery tubes of the SG bundles where

potentially locked tubes would cause high axial loads; (3) revise the

references describing electrosleeving; and (4) add the requirement that

all sleeves with detected inside diameter flaw indications will be

removed from service upon detection. In addition, if an

ElectrosleeveTM tube pull is performed by the licensee, the

licensee has agreed to provide the results of the tube examination to

the NRC staff within 60 days of when the final results of the

examination are made available to the licensee.

Date of issuance: September 13, 2002.

Effective date: September 13, 2002, and shall be implemented within

60 days of the date of issuance.

Amendment No.: 153.

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

Technical Specifications.

Date of initial notice in Federal Register: May 14, 2002 (67 FR

34494). The supplemental letter of August 5, 2002, provided additional

information that clarified the application, did not expand the scope of

the application as originally noticed, and did not change the NRC

staff's original proposed no significant hazards consideration

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

contained in a Safety Evaluation dated September 13, 2002.

No significant hazards consideration comments received: No.

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

Generating Station, Coffey County, Kansas

Date of amendment request: June 27, 2000, and its supplements dated

January 31, 2001, May 2, 2001, October 30, 2001, and May 10, 2002.

Brief description of amendment: The amendment revised the antitrust

conditions for Kansas Gas and Electric Company (KGE) in Appendix C to

the operating license. The revisions (1) add a statement that the

antitrust conditions do not restrict the rights of Kansas Electric

Power Cooperative, Inc. (KEPCo) or the duties of KGE, that may exist

beyond, and are not inconsistent with, the antitrust conditions, (2)

define ``KGE members in licensee's service area'' in the appendix to

include all KEPCo members with facilities in Western Resources' and

KGE's combined service area, (3) delete license conditions restricting

KEPCo's use of the power from WCGS, (4) remove out-of-date conditions,

and (5) update conditions to be consistent with the terms and

conditions of Western Resources' Federal Energy Regulatory Commission

open access transmission tariff. Western Resources is the parent

company of KGE.

Date of issuance: September 6, 2002.

Effective date: September 6, 2002, and shall be implemented within

90 days from the date of issuance.

Amendment No.: 147.

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

Appendix C, ``Antitrust Conditions for Kansas Gas and Electric

Company,'' to the operating license.

Date of initial notice in Federal Register: July 26, 2000 (65 FR

46010). The supplemental letters dated January 31, 2001, May 2, 2001,

October 30, 2001, and May 10, 2002, provided additional clarifying

information that did not expand the application beyond the scope of the

initial notice or change the staff's proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated September 6, 2002.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 20th day of September, 2002.

For the Nuclear Regulatory Commission.

Stuart A. Richards,

Acting Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 02-24616 Filed 9-30-02; 8:45 am]

BILLING CODE 7590-01-P