[Federal Register Volume 64, Number 202 (Wednesday, October 20, 1999)]
[Notices]
[Pages 56526-56543]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-27210]


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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

[[Page 56527]]

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 25, 1999, through October 7, 1999. 
The last biweekly notice was published on October 6, 1999 (64 FR 
54370).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By November 19, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
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

[[Page 56528]]

and make it immediately effective, notwithstanding the request for a 
hearing. Any hearing held would take place after issuance of the 
amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: September 14, 1999
    Description of amendments request: Request No. 1: The proposed 
administrative change to Technical Specification (TS) 5.5.2, Primary 
Coolant Sources Outside Containment, would delete the references to the 
post-accident sampling return piping of the radioactive waste gas 
system and the post-accident sampling return piping of the liquid 
radwaste system because the Palo Verde post-accident sampling system 
does not have return lines to the radioactive waste gas or liquid 
radwaste systems.
    Request No. 2: This proposed TS amendment would also delete the 
administrative requirement in TS 5.6.2, Annual Radiological 
Environmental Operating Report, that states: ``[t]he report shall 
identify the TLD [thermoluminescence dosimeter] results that represent 
collocated dosimeters in relation to the NRC TLD program and the 
exposure period associated with each result.''
    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:

Request No. 1

Standard 1--Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

    No--This proposed administrative change to Technical 
Specification (TS) 5.5.2 to delete references to the radioactive 
waste gas system and liquid radwaste system in the context of the 
post accident sampling system (PASS) does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Leak testing requirements of the PASS return 
piping are included in the TS 5.5.2 requirements that are not being 
changed. The appropriate PASS piping, including return piping, is 
leak tested per the prescribed requirements in TS 5.5.2. This 
administrative change would simply clarify TS 5.5.2, since the PASS 
return piping is not part of the waste gas or liquid radwaste 
systems. There is no physical connection between the PASS piping and 
the radioactive waste gas or liquid radwaste systems. The 
radioactive waste gas system and the liquid radwaste system are not 
part of PASS and would not contain highly radioactive fluids during 
a serious transient or accident to be subject to TS 5.5.2. This 
administrative change would involve no change to the design or 
maintenance of the plant and no changes in the functional 
requirements of any system.

Standard 2--Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    No--This proposed administrative change to delete references to 
the radioactive waste gas system and liquid radwaste system in the 
context of PASS does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Leak testing requirements of the PASS return piping are implicitly 
included in the TS 5.5.2 requirements that are not being changed. 
The appropriate PASS piping, including return piping, is leak tested 
per the prescribed requirements in TS 5.5.2. There is no physical 
connection between the PASS piping and the radioactive waste gas or 
liquid radwaste systems. The radioactive waste gas system and the 
liquid radwaste system are not part of PASS and would not contain 
highly radioactive fluids during a serious transient or accident to 
be subject to TS 5.5.2. This administrative change would involve no 
change to the design or maintenance of the plant and no changes in 
the functional requirements of any system. This administrative 
change would simply clarify TS 5.5.2, since the PASS return piping 
is not part of the waste gas or liquid radwaste systems.

Standard 3--Does the proposed change involve a significant reduction in 
a margin of safety?

    No--This proposed administrative change does not involve a 
significant reduction in a margin of safety. There is no margin of 
safety associated with this proposed administrative change to 
Technical Specification 5.5.2. Leak testing requirements of the PASS 
return piping are implicitly included in the TS 5.5.2 requirements 
that are not being changed. The appropriate PASS piping, including 
return piping, is leak tested per the prescribed requirements in TS 
5.5.2. This administrative change would involve no change to the 
design or maintenance of the plant and no changes in the functional 
requirements of any system. This administrative change would simply 
clarify TS 5.5.2, since the PASS return piping is not part of the 
waste gas or liquid radwaste systems.

Request No. 2

Standard 1--Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

    No--This proposed administrative change to Technical 
Specification (TS) 5.6.2 does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
This proposed TS amendment would delete the administrative 
requirement in TS 5.6.2, Annual Radiological Environmental Operating 
Report, that states: ``[t]he report shall identify the TLD results 
that represent collocated dosimeters in relation to the NRC TLD 
program and the exposure period associated with each result.'' The 
NRC ended their TLD program at the end of 1997. The requirements of 
TS 5.6.2 and the changes being made with this request are purely 
administrative reporting requirements that have no effect on the 
design, operation, or maintenance of the plant. Since there is no 
effect on the design, operation, or maintenance of the plant, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Standard 2--Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    No--This proposed administrative change to TS 5.6.2 does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. This change only affects 
administrative reporting requirement and has no effect on the 
design, operation, or maintenance of the plant. Since this proposed 
change is purely administrative and would have no effect on the 
design, operation, or maintenance of the plant, this change will not 
create possibility of a new or different type of accident than any 
previously evaluated.

[[Page 56529]]

Standard 3--Does the proposed change involve a significant reduction in 
a margin of safety?

    No--This proposed administrative change to TS 5.6.2 does not 
involve a significant reduction in a margin of safety. This TS 
establishes requirements for reporting radiological monitoring 
information to the NRC. Since TS 5.6.2 contains an administrative 
reporting requirement, and this proposed change would simply delete 
an administrative requirement associated with a discontinued NRC 
monitoring program, there is no margin of safety associated [with] 
this TS or with the proposed changes to the requirements of TS 
5.6.2. Also, since this involves only administrative reporting, this 
change has no [e]ffect on any other margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Section Chief: Stephen Dembek

CBS Corporation (Licensee), Westinghouse Test Reactor, Waltz Mill Site, 
Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-2

    Date of amendment request: September 7, 1999, as supplemented on 
October 1, 1999
    Description of amendment request: CBS Corporation is the licensee 
for the Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania. 
The licensee is authorized to only possess the reactor and a 
decommissioning plan has been approved. The licensee is planning to 
revise the decommissioning plan by reassigning the responsibilities of 
the Site Manager, who works for the Westinghouse Electric Company (a 
contractor to CBS) to the TR-2 Decommissioning Project Director who 
works for CBS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendment to a license of a facility involves no 
significant hazards consideration if 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 the margin of safety.
    The staff agrees with the licensee's no significant hazards 
consideration determination submitted on September 7, 1999, for the 
following reason:
    In order to complete the decommissioning of the WTR facility as 
described in the Decommissioning Plan, CBS has established contractual 
agreements with the Westinghouse Electric Company to supply continued 
site support and services to the Westinghouse Test Reactor Facility. 
CBS has also entered into contracts with other third party 
organizations as described in the Decommissioning Plan. These contracts 
will remain in place between CBS and each respective third party so 
that there will be no effective change in the personnel associated with 
the on-going decommissioning project under the TR-2 License. CBS 
continues to retain full responsibility for the project.
    The only change being made is that the responsibilities of the 
Westinghouse Electric Company Site Manager, as it pertains to the WTR 
and the TR-2 License, has been assigned to the TR-2 Decommissioning 
Project Director, who works for CBS. The Westinghouse Electric Company 
personnel who reported to the Site Manager will now report directly to 
CBS through the contract.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment does not modify the WTR facility 
configuration or licensed activities. Thus no new accident initiators 
are introduced. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated, and 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.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: September 16, 1999.
    Description of amendment request: The amendments would revise 
Surveillance Requirements (SRs) 3.8.4.8 and 3.8.4.9 of the Technical 
Specifications and Bases SR 3.8.4.8 to allow testing of the direct 
current (DC) channel batteries with the units on line. The proposed 
change to SR 3.8.4.8 would also prohibit the diesel generator (DG) 
batteries from being service tested while the units are on line.
    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:

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Approval of this amendment will have no 
significant effect on accident probabilities or consequences. The 
125 Volt DC Vital Instrumentation and Control Power System is not an 
accident initiating system; therefore, there will be no impact on 
any accident probabilities by the approval of this amendment. The 
design of the system is not being modified by this proposed 
amendment. It has been shown that the required battery testing can 
be performed safely with the unit on line well within the allowed 
outage time for an inoperable DC channel. Both safety trains would 
continue to be capable of performing their required design functions 
in the event of an accident. Therefore, there will be no impact on 
any accident consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant which will introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed

[[Page 56530]]

amendment. It has already been shown that both safety trains of the 
125 Volt DC Vital Instrumentation and Control Power System will 
continue to be able to perform their accident mitigation functions 
should they be required. In addition, the probabilistic risk 
analysis conducted for this proposed amendment demonstrated that 
there is no appreciable increase in overall plant risk incurred by 
its implementation. No safety margins will be impacted.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: July 29, 1999, as supplemented by letter 
dated August 30, 1999.
    Description of amendment request: The proposed amendment would 
delete a license condition that required installation of a neutron flux 
monitoring system, in the form of excore wide range monitors (WRM), in 
conformance with Regulatory Guide 1.97, ``Instrumentation for Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
Conditions During and Following an Accident.'' WNP-2 installed the WRM 
system in the spring of 1989. Removal of the license condition would 
allow WNP-2 to deactivate the WRM system. 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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. As stated in the NRC safety evaluation approving NEDO-
31558-A (Reference 2) [in licensee's August 30,1999 letter], 
Category 1 neutron flux monitoring instrumentation is not needed for 
existing BWRs to cope with Loss-of-Coolant Accident (LOCA), 
Anticipated Transient Without SCRAM (ATWS), or other accidents that 
do not result in severe core damage conditions. Instrumentation to 
monitor the progression of core melt accidents would best be 
addressed by the current severe accident management program. Also, 
WRM is not included in the WNP-2 IPE/PSA models and WRM is not 
relied upon for operator actions in the Emergency Operating 
Procedures (EOPs) or actions accounted for in Severe Accident 
Management. Therefore, no individual precursors of an accident are 
affected and the elimination of the WRM does not impact or change 
the probabilities of accidents previously evaluated. In addition, 
since the operability of plant systems designed to mitigate accident 
consequence has not changed, the consequences of an accident 
previously evaluated are not expected to increase.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
procedures that may create the potential for new or different 
personnel errors. The elimination of the WRM system does not create 
the possibility of a new or different kind of accident because plant 
crews are trained to use the Neutron Monitoring System (NMS) in 
normal evolutions and under emergency conditions according to EOP 
guidance. In addition, NEDO-31558-A concludes that the failure of 
all neutron flux monitoring instrumentation does not prevent the 
operator from determining the shutdown condition of the reactor. 
Sufficient information is available on which to base operational 
decisions and to conclude that reactivity control has been 
accomplished. For example, Rod Position Information System (RPIS) is 
powered from an uninterruptible source and remains available even 
during Station Blackout (SBO) conditions to provide full core 
control rod position information as a backup reactor power indicator 
based on calculations of rod worth and shutdown margin. The proposed 
change does not introduce any new modes of operation or alter system 
setpoints which could create a new or different kind of accident. 
Therefore, no new precursors of an accident and no new or different 
kinds of accidents are created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The elimination of the WRM system does not result in a reduction 
of the margin of safety. The neutron power indications necessary for 
operator response to ATWS are provided by the NMS not WRM. Based on 
a WNP-2 specific evaluation against the alternate criteria specified 
in NEDO-31558-A, there is sufficient confidence that the 
instrumentation would still be available to confirm that the reactor 
is shutdown. In addition, failure of the existing neutron flux 
monitoring instrumentation does not prevent plant operators from 
determining the shutdown condition of the reactor. Sufficient 
information is available to the operator to make operational 
decisions and to conclude that reactivity control has been 
accomplished. The proposed changes will not impact the basis for any 
Technical Specification related to the establishment or maintenance 
of nuclear safety margins. Therefore, operation of the facility in 
accordance with the proposed amendment does not involve a reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: February 19, 1999.
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 Improved Technical Specifications 
Sections 5.6.2.7, 5.6.2.8, and 5.7.2.b, related to the Containment 
Tendon Surveillance Program. The proposed changes are a result of 
revisions to 10 CFR 50.55a which are required to be fully implemented 
by September 9, 2001. These revised requirements affect the 
surveillance methods for the containment tendons and the conduct of 
containment visual inspections, and the methods of reporting the 
results of the required inspections to the NRC.
    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?
    No. The proposed change to the Crystal River Unit 3 (CR-3) 
Improved Technical Specifications (ITS) replaces the previous 
programmatic commitment to implement a Containment Tendon 
Surveillance Program based on Regulatory Guide 1.35, Revision 3,

[[Page 56531]]

with a Containment Inspection Program that complies with the current 
requirements of 10 CFR 50.55a. Effective September 9, 1996, 10 CFR 
50.55a requires licensees to implement a Containment Inspection 
Program in compliance with the 1992 Edition with the 1992 Addenda of 
Subsection IWE, ``Requirements for Class MC and Metallic Liners of 
Class CC Components of Light-Water Cooled Power Plants,'' and with 
Subsection IWL, ``Requirements for Class CC Concrete Components of 
Light-Water Cooled Power Plants,'' of Section XI, Division 1, of the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code) with additional modifications and limitations as 
stated in 10 CFR 50.55a(b)(2)(ix). Florida Power Corporation (FPC) 
is implementing a Containment Inspection Program to comply with 
these new regulatory requirements. The final rule specifies 
requirements to assure that the critical areas of the containment 
structure are routinely inspected to detect and take corrective 
action for defects that could compromise structural integrity. This 
proposed ITS change is requested to update the ITS to these latest 
10 CFR 50.55a regulatory requirements.
    By complying with the regulatory requirements described in 10 
CFR 50.55a, the probability of a loss of containment structural 
integrity is maintained as low as reasonably achievable. Maintaining 
containment structural integrity is independent of the operation of 
the reactor coolant system (RCS), and independent of the reactor 
protection system (RPS) and emergency core cooling system (ECCS). 
The Containment Inspection Program ensures that the containment will 
function as designed to provide an acceptable barrier to release of 
radioactive materials to the environment. By assuring the 
effectiveness of this barrier through appropriate inspection, and by 
implementing corrective actions for any degradation discovered 
during these inspections that might lead to containment structural 
failures, the probability or consequences of accidents will not be 
greater than that previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    No. Maintaining containment structural integrity is independent 
of the operation of the RCS, and independent of the RPS and ECCS. By 
implementing corrective actions for any degradation discovered 
during the required inspections of the containment, the possibility 
of a new or different kind of accident will not be created.
    3. Involve a significant reduction in a margin of safety?
    No. The margin of safety as defined by the CR-3 ITS has not been 
reduced. By complying with the regulatory requirements described in 
10 CFR 50.55a, the probability of a loss of containment structural 
integrity is maintained as low as reasonably achievable. The 
Containment Inspection Program ensures that the containment will 
function as designed to provide an acceptable barrier to release of 
radioactive materials to the environment. By implementing the 
Containment Inspection Program, the existing margin of safety is 
preserved.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel (MAC-
BT15A), Florida Power Corporation, P. O. Box 14042, St. Petersburg, 
Florida 33733-4042.
    NRC Section Chief: Sheri R. Peterson.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: July 7, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to change the component 
surveillance frequencies for the following TSs to indicate a frequency 
of once per 3 months: Core Spray System TS 4.4.A.1 and 4.4.A.2, 
Containment Cooling System TS 4.4.C.1, Emergency Service Water System 
TS 4.4.D.1, Fire Protection System TS 4.4.F (isolation valves only), 
and Pressure Suppression Chamber--Drywell Vacuum Breakers TS 4.5.F.5.a. 
The TSs currently stipulate a component surveillance frequency of once 
per month. Also, the amendment would revise TS pages 4.4-1 and 4.4-2 to 
incorporate editorial format changes and TS page 4.4-3 to accommodate 
the expanded text.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed surveillance interval change does not alter the 
actual surveillance requirements, nor does it alter the limits and 
restrictions on plant operations. The reliability of systems and 
components relied upon to prevent or mitigate the consequences of 
accidents previously evaluated is not degraded by the proposed 
change to the surveillance interval. Assurance of system and 
equipment availability is maintained. The proposed change does not 
alter any system or equipment configuration.
    Based on the above, the proposed change does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed surveillance interval change does not alter the 
actual surveillance requirements, nor does it alter the limits and 
restrictions on plant operations. Assurance of system and equipment 
availability is maintained. The proposed change does not alter any 
system or equipment configuration nor does it introduce any new 
mechanisms which could contribute to the creation of a new or 
different kind of accident than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed change extends the surveillance interval for 
verifying the operability of the specified pumps and valves from 
once per month to once per three months. The proposed change does 
not alter the actual surveillance requirements, the limits and 
restriction on plant operations nor the design, function or manner 
of operation of any structures, systems or components. System 
availability and reliability are maintained. Accordingly, the 
proposed TS change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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 amendment requests: September 17, 1999.
    Description of amendment requests: The proposed amendments would 
allow credit in the applicable subcriticality analysis for the negative 
reactivity provided by insertion of the rod cluster control assemblies 
(RCCAs) during realignment from a cold leg recirculation to a hot leg 
recirculation configuration. This realignment, which is referred to as 
hot leg switchover, is performed following a loss-of-coolant accident. 
This methodology change, when evaluated in accordance with 10 CFR 
59.59, resulted in an unreviewed safety question that will require 
prior approval by the NRC staff in accordance with the provisions of 10 
CFR 50.90

[[Page 56532]]

prior to implementation. The proposed change would also affect the 
Bases for Technical Specification (T/S) 3/4.5.5, ``Refueling Water 
Storage Tank,'' and several sections of the Updated Final Safety 
Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    No. I&M [Indiana Michigan Power Company] proposes to credit RCCA 
insertion of negative reactivity for criticality control during the 
core cooling flow path realignment from cold leg recirculation to 
hot leg recirculation following the postulated cold leg LBLOCA 
[large-break loss-of-coolant accident]. No physical modifications 
will be made to plant systems, structures, or components.
    Credit for RCCAs is only being applied to demonstrate core 
subcriticality upon hot leg switchover (HLSO) following a cold leg 
LBLOCA. The performance criteria codified in 10 CFR 50.46 continue 
to be met. The ability of the RCCAs to insert under LOCA and seismic 
conditions was a function important to safety as part of the 
original CNP [Cook Nuclear Plant] design basis. This is supported by 
the conclusion presented in NRC (at the time, the Atomic Energy 
Commission) Safety Evaluation Report (SER), Section 3.3, 
``Mechanical Design of Reactor Internals,'' dated January 14, 1969. 
The SER includes the statements that, ``[t]he control rod guide 
tubes are designed so that each finger of each control rod assembly 
is always partially inserted in the guide tube. Deflection limits on 
the guide tubes have been chosen so that deflections caused by blow-
down forces during a loss-of-coolant accident will not prevent 
control rod insertion,'' and that the ``* * * mechanical design of 
internals, fuel assemblies, and control elements is acceptable.'' 
However, the licensing basis safety analyses for the LBLOCA scenario 
have conservatively not taken credit for insertion of the RCCAs.
    No physical modifications will be made to plant systems, 
structures, or components in order to implement the proposed 
methodology change. The safety functions of the safety related 
systems and components, which are related to accident mitigation, 
have not been altered. Therefore, the reliability of RCCA insertion 
is not affected. As such, taking credit for RCCA insertion does not 
alter the probability of an LBLOCA (the design basis accident at 
issue). The Westinghouse analyses provided as Attachments 6 and 7 
[to the licensee's application] demonstrate that RCCA insertion will 
occur, with substantial margin, following a design basis cold leg 
LBLOCA combined with a seismic event. Crediting RCCA insertion does 
not affect mechanisms for a malfunction that could impact the HLSO 
subcriticality analysis, or mechanisms that could initiate a LOCA. 
Taking credit for the negative reactivity available from insertion 
of the RCCAs, which is currently assumed for various accident 
analyses within the CNP licensing basis (e.g., small break LOCA, 
main steamline break, feedline break, steam generator tube rupture), 
does not affect equipment malfunction probability directly or 
indirectly. Therefore, crediting the RCCAs as a source of negative 
reactivity for post-LOCA criticality control at the time of HLSO 
does not significantly increase the probability of an accident 
previously evaluated.
    Furthermore, the traditional conservative assumption that the 
most reactive RCCA is stuck fully out of the core is being 
maintained. A malfunction that results in one RCCA to fail to insert 
is a credible scenario, and is being considered for the post-LOCA 
subcriticality analysis following a cold leg LBLOCA. There will be 
sufficient negative reactivity, even with the most reactive RCCA 
stuck fully out of the core, to assure core subcriticality post-
LOCA, as supported by the subcriticality analysis that is confirmed 
each and every fuel cycle as part of the reload documentation (i.e., 
the Reload Safety Evaluations). The core is shown to remain 
subcritical during the post-LOCA long-term cooling period, 
specifically while HLSO is performed. Thus, no additional 
radiological source terms are generated, and the consequences of an 
accident previously evaluated in the UFSAR will not be significantly 
increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed change involves crediting the negative 
reactivity that is available from the RCCAs for an analysis 
applicable several hours after the initiation of a cold leg LBLOCA. 
As such, this change involves post-LOCA recovery actions several 
hours after the break has occurred and does not involve accident 
initiation. As discussed above, the original design requirements for 
the CNP reactor internals, core fuel assemblies, and RCCAs were 
based upon assuring the ability of the RCCAs to insert following a 
double-ended rupture LOCA with seismic loadings. Thus, the safety 
functions of safety related systems and components have not been 
altered by this change. Crediting the negative reactivity that is 
available from the RCCAs for the post-LOCA subcriticality analysis 
upon HLSO does not cause the initiation of any accident, nor does 
the proposed activity create any new credible limiting single 
failure. Crediting the insertion of RCCAs does not result in any 
event previously deemed incredible being made credible nor is there 
any introduction of any new failure mechanisms that are not 
currently considered in the design basis LOCA. There are no changes 
introduced by this amendment concerning how safety related equipment 
is designed to operate under normal or design basis accident 
conditions since the calculations supporting RCCA insertion 
following a cold leg LBLOCA have assumed design basis break sizes in 
conjunction with seismic loadings. Therefore, the possibility of an 
accident of a different type than already evaluated in the UFSAR is 
not created.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. Presently, no credit is taken for RCCA insertion in the 
analysis to demonstrate post-cold leg LOCA subcriticality at the 
time of HLSO. The current subcriticality analysis for this scenario 
relies only on the boron provided by the RWST [refueling water 
storage tank] and the accumulators. Thus, RCCA insertion provides 
another source of negative reactivity (margin of safety). Revising 
the post-cold leg LBLOCA HLSO subcriticality analysis to credit the 
negative reactivity associated with the RCCAs is a means to offset 
the sump dilution associated with the effects of the inactive 
regions of the CNP containment sump. The incorporation of this 
``defense-in-depth'' source of negative reactivity in the HLSO 
subcriticality analysis has been conservatively determined to cause 
a reduction in the margin of safety. 10 CFR 50, Appendix K, I.A.2., 
states, in part, that ``[r]od trip and insertion may be assumed if 
they are calculated to occur,'' and provides for crediting RCCA 
insertion as an acceptable feature of emergency core cooling system 
(ECCS) evaluation models. The proposed change is based upon an 
analysis for CNP that demonstrates that the control rods will indeed 
insert and the resulting negative reactivity can be credited for 
post-LOCA criticality control.
    The proposed change would ensure that post-LOCA subcriticality 
is maintained during HLSO. Subsequently, there would not be a 
challenge to long-term core cooling due to a return to a critical 
condition. This being the case, the requirements of 10 CFR 
50.46(b)(5) that, ``* * * the calculated core temperature shall be 
maintained at an acceptably low value and decay heat shall be 
removed for the extended period of time* * *'' continues to be 
satisfied and the margin of safety in the CNP licensing basis is 
preserved. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: September 29, 1999.
    Description of amendment request: The proposed amendment requests a 
Technical Specification change that

[[Page 56533]]

would extend the allowed out-of-service time for the residual heat 
removal service water system (RHRSW) from 7 days to 11 days on a one-
time basis while modifications are made on the RHRSW ``A'' strainer.
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:
    Involve an increase in the probability or consequences of an 
accident previously evaluated.
    The Conditional Core Damage Probability due to this proposed 
change is calculated to be 6.4 E-8. This value falls below the 
threshold probability of 1 E-6 for risk significance of temporary 
changes to the plant configuration in the EPRI PSA [Electric Power 
Research Institute Probability Assessment] Applications Guide 
(Reference 3) [see application dated September 29, 1999].
    This proposed change does not increase the consequences of an 
accident previously evaluated because all relevant accidents (LOCA) 
[loss-of-coolant accident] would result in the transfer of decay 
heat to the suppression pool. For this scenario, the same complement 
of equipment will be available to achieve and maintain cold shutdown 
as is required by the current Technical Specification LCO [limiting 
condition for operation].
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not physically alter the plant. As 
such, no new or different types of equipment will be installed. The 
new design for the RHRSW strainer packing gland will be evaluated 
under a separate 10 CFR 50.59 evaluation and is considered to be 
functionally equivalent for the purposes of this one-time-only 
proposed Technical Specification change.
    The implementation and use of the contingency plan for achieving 
limited containment heat removal in the event the B division of 
RHRSW is rendered inoperable will be evaluated under the Authority's 
10 CFR 50.59 program.
    Involve a significant reduction in a margin of safety.
    The Conditional Core Damage Probability due to this proposed 
change is calculated to be 6.4 E-8. This value falls below the 
threshold probability of 1 E-6 for risk significance of temporary 
changes to the plant configuration in the EPRI PSA Applications 
Guide (Reference 3).
    The consequences of a postulated accident occurring during the 
extended allowable out-service time are bounded by existing analyses 
therefore there is no significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: December 1, 1998, as supplemented by 
letters of April 21, 1999, and July 19, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to reflect replacing the current 
Model 51 steam generators with Westinghouse Model 54F steam generators. 
The replacement program includes re-analyzing and evaluating loss-of-
coolant-accident (LOCA) and non-LOCA mass and energy releases, 
containment and sub-compartment pressure and temperature responses, 
dose analyses, and the effects on nuclear steam supply and balance of 
plant systems.
    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 changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the [Final Safety Analysis Report] FSAR. The comprehensive 
engineering effort performed to support [steam generator] SG 
replacement has included evaluations or re-analysis of all accident 
analyses including all dose related events. All dose consequences 
have been analyzed or evaluated with respect to these proposed 
changes, and all acceptance criteria continue to be met. Therefore, 
these changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident than any accident already evaluated in 
the FSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed changes. 
The proposed technical specification changes have no adverse effects 
on any safety-related system and do not challenge the performance or 
integrity of any safety-related system. Therefore, these changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed technical specification changes do not involve a 
significant reduction in a margin of safety. All applicable analyses 
supporting the [steam generator] SG replacement reflect these 
proposed values. All acceptance criteria (including LOCA peak clad 
temperature, [departure from nucleate boiling] DNB, containment 
temperature and pressure, and dose limits) continue to be met. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed Southern Nuclear Company's analysis, and 
based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.
    NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 30, 1999 (TS 98-10).
    Description of amendment requests: The proposed amendments would 
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-
79(Unit 2) by updating the current Technical Specification requirements 
for reactor coolant system leakage detection and operational leakage 
specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revisions enhance the Technical specification (TS) 
requirements to provide greater consistency with the standard TS in 
NUREG-1431. This revision proposes changes to the requirements for 
reactor coolant system (RCS) leak detection and RCS operational 
leakage in Specifications 3.4.6.1 and 3.4.6.2, respectively. New 
Specifications

[[Page 56534]]

3.4.6.3 and 3.5.6 for RCS pressure isolation valves and emergency 
core cooling system (ECCS) seal injection flow have been added to 
improve consistency with NUREG-1431. The proposed revisions are not 
the result of changes to plant equipment, system design, testing 
methods, or operating practices. The modified requirements will 
allow some relaxation of current operability criteria, action 
requirements, and surveillance requirements (SRs). These changes 
provide more appropriate requirements in consideration of the safety 
significance and the design capabilities of the plant as determined 
by the improved standard TS industry effort. These specifications 
serve to primarily provide identification and control of the RCS 
fission product barrier leakage and ECCS degradation and are not 
considered to be a contributor to the generation of postulated 
accidents. Since these proposed revisions will continue to support 
the required safety functions, without modification of the plant 
features, the probability of an accident is not increased.
    The proposed changes will allow relaxation of action times for 
inoperable leak detection features and the components that can be 
inoperable. The required actions to ensure acceptable pressure 
isolation valve capability with an inoperable valve have been 
revised to allow isolation by a single valve for a limited period of 
time. These revisions will allow unit operation for a longer period 
of time with reduced system redundancy. However, the redundancy 
reduction and action time increases are not significant and will 
continue to provide an acceptable level of safety considering the 
significance of RCS leakage, other design features or compensatory 
actions that provide equivalent functions, and the unlikely chance 
of an event that would require functions for leakage identification 
during the proposed time interval. These considerations are 
consistent with the basis developed by the industry and NRC for 
NUREG-1431. Surveillances have been removed from the RCS operational 
leakage specification as a result of relocated requirements, 
duplication of other SRs, and testing requirements that do not 
provide a significant benefit in the identification of RCS leakage. 
The SRs that have been retained or relocated to other TS 
specifications will provide acceptable verifications for the timely 
identification of conditions that indicate an unacceptable amount of 
RCS leakage or potential ECCS degradation resulting from excessive 
seal injection flow.
    The limiting condition for operation associated with the seal 
injection flow requirements has been revised to utilize a modified 
operability criteria. The proposed change will provide a range of 
differential pressures and the corresponding seal flows that would 
be representative of the existing single point flow limit. This 
change does not alter the intent of the operability requirements, 
but does allow the flexibility to use equivalent values that provide 
the same level of assurance for ECCS operability. The proposed 
operability condition for seal injection flow enhances the current 
requirement by establishing additional test parameters that will 
ensure that the amount of seal injection flow does not degrade the 
ECCS functions.
    The proposed changes to the SQN TS provide flexibility without 
modifying the functions of required safety systems. In many 
instances the proposed changes ensure that plant conditions for 
surveillance testing are more appropriate for testing purposes and 
the verification of system operability.
    These changes are consistent with the intent of NUREG-1431 and 
result in the enhancement of the SQN TSs based on the latest 
industry and NRC positions. The provisions proposed in this change 
request will continue to maintain an acceptable level of protection 
for the health and safety of the public and will not significantly 
impact the potential for the offsite release of radioactive 
products. The overall effect of the proposed change will result in 
specifications that have equivalent or improved requirements 
compared to existing specifications for RCS leakage and ECCS 
operability and will not significantly increase the consequences of 
an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revisions are not the result of changes to plant 
equipment, system design, testing methods, or operating practices. 
The modified requirements will allow some relaxation of current 
operability criteria, action requirements, and SRs consistent with 
NUREG-1431. These changes provide more appropriate requirements in 
consideration of the safety significance and the design capabilities 
of the plant as determined by the improved standard TS industry 
effort. These specifications serve to primarily provide 
identification and control of the RCS fission product barrier 
leakage and ECCS degradation and are not considered to be a 
contributor to the generation of postulated accidents. Since the 
functions of the associated systems will continue to perform without 
change and were not previously considered to contribute to accident 
generation, the proposed changes will not create the possibility of 
a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes, associated with RCS leakage and ECCS 
functions, will not result in changes to system design or setpoints 
that are intended to ensure timely identification of plant 
conditions that could be precursors to accidents or potential 
degradation of accident mitigation systems. These systems will 
continue to operate without change and only the associated actions 
or testing activities have been altered. Revisions to the actions 
and surveillances provide some relaxation and flexibility such that 
longer intervals are allowed for inoperable components and testing 
requirements are revised to provide conditions that provide more 
accurate results. The increased action times are acceptable 
considering the available redundant features, the compensatory 
measures provided by the actions, and the allowed time intervals 
that have been developed by the industry and NRC and recommended in 
NUREG-1431. The SR changes actually provide test condition 
requirements that enhance the accuracy of the activity even though 
they may allow a delay in the performance of the test. These 
surveillance changes are also in accordance with NUREG-1431 
recommendations.
    These revisions will continue to provide the necessary actions 
to minimize the impact of inoperable equipment to an acceptable 
level and will provide testing activities that will ensure system 
operability. Since the setpoints and design features that support 
the margin of safety are unchanged and actions for inoperable 
systems continue to provide appropriate time limits and compensatory 
measures, the proposed changes will not significantly reduce the 
margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: September 28, 1999 (TS 99-007).
    Description of amendment request: The proposed amendment on 
Response Time Test (RTT) elimination would revise the Watts Bar Nuclear 
Plant Unit 1 Technical Specifications (TS) definitions for ``Engineered 
Safety Feature (ESF) Response Time'' and ``Reactor Trip System (RTS) 
Response Time'' to provide for verification of response time for 
selected components provided that the components and the methodology 
for verification have been previously reviewed and approved by the NRC. 
In addition, associated changes to the Bases for Surveillance 
Requirements would also be made.
    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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the TS does not result in a condition where the 
design, material, and

[[Page 56535]]

construction standards that were applicable prior to the change are 
altered. The same RTS and ESF instrumentation is being used, the 
time response allocations/modeling assumptions in the Chapter 15 
analyses are unchanged; only the method of verifying 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 activity will 
not change, degrade or prevent actions, or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the UFSAR [Updated Final Safety Analysis 
Report]. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. 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 pressure and 
differential pressure transmitters, process protection racks (Eagle 
21), nuclear instrumentation (NIS), and logic system (SSPS) used in 
the plant protection systems. These components/systems will still 
have response time verified by test prior to placing the equipment 
in operational service and after any maintenance that could affect 
the response time of that equipment. Changing the method of 
periodically verifying instrument response time for applicable 
instrumentation from RTT to calibration and channel checks or 
functional test will not create any new accident initiators or 
scenarios. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a 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 selected pressure and pressure differential 
sensors, Eagle 21, NIS, and SSPS is modified to allow use of actual 
test data or engineering data. The method of verification still 
provides assurance that the total system response time is within 
that assumed in the safety analysis, since calibration checks and 
functional tests will detect any degradation which might 
significantly affect equipment response time. Therefore, the 
proposed license amendment request does not result in a significant 
reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri Peterson.

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

    Date of amendment request: June 15, 1999.
    Description of amendment request: The licensee proposed revisions 
to Technical Specifications (TSs) Sections 3.1/4.1 Reactor Protection 
System and 3.2/4.2 Protective Instrument Systems instrumentation, 
tables, and the associated bases to increase the surveillance test 
intervals (STIs), add allowable out-of-service times (AOTs), replace 
generic ECCS actions for inoperable instrument channels with function-
specific actions, and relocate selected trip functions from the TSs to 
a Vermont Yankee (VY) controlled document. In addition, revision to TS 
Section 3.1/4.1 Reactor Protection System and the associated bases is 
proposed to remove the RUN Mode APRM Downscale/IRM High Flux/
Inoperative Scram Trip Function (APRM Downscale RUN Mode SCRAM). The 
submittal also proposes to implement editorial corrections and 
administrative changes that do not alter the meaning or intent of the 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    VY has determined that the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The generic analysis contained in 
Licensing Topical Report NEDC-30851P-A assessed the impact of 
changing SCRAM (RPS) surveillance test intervals for Logic and 
Functional tests (STIs) and adding allowable out-of-service times 
(AOTs) on the SCRAM (RPS) failure frequency, the scram frequency and 
equipment cycling. Specifically, Section 5.7.4, ``Significant 
Hazards Assessment,'' of NEDC-30851P-A states that:
    ``Fewer challenges to the safeguards system, due to less 
frequent testing of the RPS, conservatively results in a decrease of 
approximately one percent in core damage frequency. This decrease is 
based upon the following:
    Based on the plant-specific experience presented in Appendix J, 
the estimated reduction in scram frequency (0.3 scrams/ yr.) 
represents a 1 to 2 percent decrease in core damage frequency based 
on the BWR plant-specific Probabilistic Risk Assessments (PRAS) 
listed in Table 5-8.
    The increase in core damage frequency due to less frequent 
testing is less than one percent. This increase is even lower (less 
than 0.01 percent) when the changes resulting from the 
implementation of the Anticipated Transients Without Scram (ATWS) 
rule are considered. Therefore, this increase is more than offset by 
the decrease in CDF due to fewer scrams.
    The effect of reducing unnecessary cycles on RPS equipment, 
although not easily quantifiable, also results in a decrease in core 
damage frequency.
    The overall impact on core damage frequency of the changes in 
allowable out-of-service times is negligible.''
    From this generic analysis, the BWR Owners' Group concluded that 
the proposed changes do not significantly increase the probability 
or consequences of an accident previously evaluated, namely the 
increase in probability of a scram failure due to SCRAM (RPS) 
unavailability is insignificant, and the overall probability of an 
accident is actually decreased as the time the SCRAM (RPS) 
Instrumentation logic operates as designed is increased resulting in 
less inadvertent scrams during testing and repair. Furthermore, the 
plant specific reports demonstrate[ ] that although VY differs from 
the generic model analyzed in License Topical Report NEDC-30851P-A, 
the net effect of the plant-specific differences do not alter the 
generic conclusions.
    The generic analysis contained in Licensing Topical Reports 
NEDC-30851P-A Suppl 2/NEDC-31677P-A assessed the impact of changing 
STIs and AOTs for BWR Isolation Instrumentation common/not common to 
SCRAM (RPS) and ECCS instrumentation. Specifically, Section 4.0, 
``Summary of Results,'' of NEDC-30851P-A Suppl 2 states that:
    ``The results indicate that the effects on probability of 
failure to initiate isolation are very small and the effects on 
probability or frequency of failure to isolate are negligible in 
nearly every case. In addition, the results indicate that increasing 
the AOT to 24 hours for tests and repairs has a negligible effect on 
the probability of failure of the isolation function. These combined 
with changes to the testing intervals and allowed out-of-service 
times for RPS and ECCS instrumentation provide a net improvement to 
plant safety and operations.''
and Section 5.6, ``Assessment of Net Effect of Changes,'' of NEDC-
31677P-A states that:
    ``A reduction in core damage frequency (CDF) of at least as much 
as estimated in the ECCS instrumentation analysis can be expected 
when the isolation actuation instrumentation STIs are changed from 
one month to three months. The chief contributor to this reduction 
is the channel functional tests for the MSIVs. Inadvertent closure 
of the MSIVs will cause an unnecessary plant scram. This reduction 
in CDF more than compensates for any small incremental

[[Page 56536]]

increase (10% or 1OE-07/year) in calculated isolation function 
failure frequency when the STI is extended to three months.''
    From this generic analysis, the BWR Owners' Group concluded that 
the proposed changes do not significantly increase the consequences 
of an accident previously evaluated, namely the increase in 
probability of an isolation failure due to isolation instrumentation 
unavailability is insignificant, and the overall probability of an 
accident is actually decreased as the time the SCRAM (RPS) 
Instrumentation logic operates as designed is increased resulting in 
less inadvertent scrams during testing and repair.
    The generic analysis contained in Licensing Topical Report NEDC-
30936P-A (Parts 1 and 2) assessed the impact of changing STIs and 
AOTs for all BWR ECCS Actuation Instrumentation. Specifically, 
Section 4.0, ``Technical Assessment of Changes,'' of NEDC-30936P-A 
(Part 2) states that:
    ``The results indicate an insignificant (less than 5E-7 per 
year) increase in water injection function failure frequency when 
STIs are increased from 31 days to 92 days, AOTs for repair of the 
ECCS actuation instrumentation are increased from one hour to 24 
hours, and AOTs for surveillance testing are increased from two to 
six hours. For all four BWR models the increase represents less than 
4% increase in failure frequency. However, when other factors which 
influence the overall plant safety are considered, the net result is 
judged to be an improvement in plant safety.''
    From this generic analysis, the BWR Owners' Group concluded that 
the proposed changes do not significantly increase the probability 
or consequences of an accident previously evaluated, namely the 
increase in probability of a water injection failure due to ECCS 
instrumentation unavailability is insignificant and the net result 
is judged to be an improvement in plant safety. Furthermore, the 
plant specific report demonstrates that although VY differs from the 
generic model analyzed in Licensing Topical Report NEDC30936P-A, the 
net affect of the plant-specific differences do not alter the 
generic conclusions.
    The generic analysis contained in Licensing Topical Report NEDC-
30851 P-A Supp 1, assessed the impact of changing Rod Block STIs on 
Rod Block failure frequency. Specifically, Section 5 (BNL's Tech. 
Eval. Report--Attach. 2 to the NRC SER) of NEDC-30851 P-A Suppl 1 
states that:
    ``The BWR Owners'' Group proposed changes to the Technical 
Specifications concerning the test requirements for BWR control rod 
block instrumentation. The changes consist of increasing the 
surveillance test intervals from one to three months. These test 
interval extensions are consistent with the already approved changes 
to STIs for the reactor protection system. The technical analysis 
reviewed and verified as documented herein indicates that there will 
be no significant changes in the availability of the control rod 
block function if these changes are implemented. In addition, there 
will be a negligible impact on the plant core melt frequency due to 
the decreased testing.''
    From this generic analysis, the BWR Owners' Group concluded that 
the proposed changes do not significantly increase the probability 
of an accident previously evaluated or consequences of an accident 
previously evaluated.
    Bases contained in GE Topical Report GENE-770-06-1 assessed the 
impact of changing STIs and AOTs on selected systems failure 
frequency. Specifically, Section 2.0, ``Summary,'' of GENE 770-06-1 
states that:
    ``Technical bases are provided for selected proposed changes to 
the instrumentation STIs and AOTs that were identified in the BWROG 
Improved BWR Technical Specification activity. These STI and AOT 
changes are consistent with approved changes to the RPS, ECCS, and 
isolation actuation instrumentation. These proposed changes do not 
result in a degradation to overall plant safety.''
    From these Bases, the BWR Owners' Group concluded that the 
proposed changes do not significantly increase the probability of an 
accident previously evaluated or consequences of an accident 
previously evaluated.
    Bases contained in GE Topical Report GENE-770-06-2 assessed the 
impact of changing STIs and AOTs on selected systems (RCIC 
Actuation) failure frequency. Specifically, Section 2.0, 
``Summary,'' of GENE 770-06-2 states that:
    ``The STI and AOT changes to the RCIC actuation instrumentation 
are justified based on their small effect on the water injection 
function unavailability and consistency with comparable changes to 
the actuation instrumentation for the other ECCS subsystems''. These 
STI and AOT changes are consistent with approved changes to the RPS, 
ECCS, and isolation actuation instrumentation. These proposed 
changes do not result in a degradation to overall plant safety.''
    From these Bases, the BWR Owners' Group concluded that the 
proposed changes do not significantly increase the probability of an 
accident previously evaluated or consequences of an accident 
previously evaluated.
    The proposed change will not alter the physical characteristics 
of any plant systems or components and all safety-related systems 
and components remain within their applicable design limits. Thus, 
system and component performance is not adversely affected by this 
change, thereby assuring that the design capabilities of those 
systems and components are not challenged in a manner not previously 
assessed so as to create the possibility of a new or different kind 
of accident.
    The addition of allowable out-of-service times (AOTs) and the 
increase in surveillance test intervals (STIS) does not alter the 
function of the SCRAM (RPS), ECCS, Isolation, Rod Block, and 
Selected Instrument Systems nor involve any type of plant 
modification and no new modes of plant operation are involved with 
these changes.
    No physical change is being made to any systems or components 
that are credited in the safety analysis, therefore there is no 
change in the probability or consequences of any accident analyzed 
in the UFSAR.
    The design basis accident applicable to the startup power region 
is the Control Rod Drop Accident (CRDA). The UFSAR does not credit 
the RUN Mode IRM High Flux/Inoperative with the associated APRM 
downscale scram Trip Function (APRM downscale RUN Mode SCRAM) in the 
termination of this accident, Accident mitigation is provided by the 
APRM 120% power scram. Therefore, elimination of the APRM downscale 
RUN Mode SCRAM function has no adverse affect on previously 
evaluated accidents.
    The Continuous Control Rod Withdrawal Error (CWE) transient is 
terminated by the Rod Block Monitor (RBM) in the RUN Mode. The APRM 
Reduced High Flux Scram provides the primary STARTUP Mode protection 
in conjunction with the IRMs and limits the consequences of this 
transient. Therefore, elimination of the APRM downscale RUN Mode 
SCRAM function has no effect on the consequences of this transient.
    Adding a new surveillance to verify SRM/IRM/APRM will enhance 
neutron monitoring during startups and shutdowns and does not have 
an adverse affect on previously evaluated accidents.
    None of the proposed changes will affect any of the rod blocks 
or other precursor events to either the CRDA or CWE. Therefore, 
there is no change in the probability of any accident previously 
analyzed.
    Use of ECCS Function-specific AOTs, actions and relocation of 
Bus Power Monitors to a licensee controlled document is consistent 
with STS and does not have an adverse affect on previously evaluated 
accidents.
    In addition, VY concluded the editorial corrections and 
administrative changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
These changes do not alter the meaning or intent of any 
requirements.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    VY has determined that the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change will not alter the physical characteristics 
of any plant systems or components and all safety-related systems 
and components remain within their applicable design limits. Thus, 
system and component performance is not adversely affected by this 
change, thereby assuring that the design capabilities of those 
systems and components are not challenged in a manner not previously 
assessed so as to create the possibility of a new or different kind 
of accident. Editorial corrections and administrative changes do not 
alter the meaning or intent of any requirements.
    The addition of allowable out-of-service times (AOTs), ECCS 
function-specific actions and the increase in surveillance test 
intervals (STIs) does not alter the function of the SCRAM (RPS), 
ECCS, Isolation, Rod Block,

[[Page 56537]]

and Selected Instrument Systems nor involve any type of plan 
modification and no new modes of plant operation are involved with 
these changes. Therefore, operation in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Elimination of APRM downscale RUN Mode SCRAM function affects 
only the operations of neutron monitoring and protective systems 
(IRM and APRM) which provide indication and mitigation actions only. 
Operation of these systems does not create the possibility for new 
precursors (such as reactivity) which would introduce a new or 
different kind of accident from any accident previously evaluated.
    Additionally, the proposed changes do not affect the ability of 
those systems required to mitigate previously evaluated accidents 
during the modes they are credited.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety The NRC staff has 
reviewed and approved the generic studies contained in the GE 
Topical Reports (LTRs) and has concurred with the BWR Owners' Group 
that the proposed changes do not significantly affect the 
availability of the SCRAM (RPS), ECCS, Isolation, Rod Block, or 
Selected Instrument Systems. The proposed addition of allowable out-
of-service times (AOTs) for the instruments addressed in the LTRs 
provide reasonable time for making repairs and performing tests. The 
lack of sufficient AOTs in the current Technical Specifications (TS) 
creates a hurried atmosphere during repairs and tests that could 
cause an increased risk of error. In addition, placing an individual 
channel in a tripped condition because no AOT exists, as in the 
current TS, increases the potential of an inadvertent scram. The 
proposed AOTs provide realistic times to complete the required 
actions without increasing the overall instrument failure frequency. 
Use of ECCS Function-specific AOTs, actions and relocation of Bus 
Power Monitors to a licensee controlled document is consistent with 
STS and there is no significant reduction in the margin of safety.
    Editorial corrections and administrative changes do not alter 
the meaning or intent of any requirements. Therefore, there is no 
significant reduction in the margin of safety.
    The incorporation of extended surveillance test intervals (STIs) 
does not result in significant changes in the probability of 
instrument failure, as demonstrated by the LTRs. In addition, the TS 
calibration frequency has not changed, and therefore assurance 
exists that the setpoints will not be affected by drift.
    These changes, when coupled with the reduced probability of 
test-induced plant transients and equipment failures, result in an 
overall increase in the margin of safety.
    The only scram function that the UFSAR takes credit for in the 
mitigation of the limiting accident (control rod drop accident) is 
the APRM 120% power scram which is not affected by this change. Only 
the APRM Downscale RUN Mode SCRAM, for which the UFSAR takes no 
credit in the termination of any analyzed event, is removed by this 
change. Removal of the APRM Downscale RUN Mode SCRAM will avoid the 
need to operate the plant in a ``half scram'' condition with the 
potential for an inadvertent plant transient. For these reasons, the 
change does not involve a significant reduction in a margin of 
safety.
    The Continuous Control Rod Withdrawal Error (CWE) transient is 
terminated by the Rod Block Monitor (RBM) in the RUN Mode. When 
initiated from the STARTUP Mode, the consequences of a CWE are 
limited by the APRM Reduced High Flux scram in conjunction with the 
IRM scram function. Therefore eliminating the TS requirement for the 
APRM Downscale RUN Mode SCRAM will not reduce the margin of safety 
for this transient.
    Adding a new surveillance to verify SRM/IRM/APRM overlap will 
enhance neutron monitoring during startups and shutdown, and 
consequently does not involve a significant reduction in a margin of 
safety.
    On the basis of the above, VY has determined that operation of 
the facility in accordance with the proposed change does not involve 
a significant hazards consideration as defined in 10 CFR 50.92(c), 
in that it: (1) does not involve a significant increase in the 
probability or consequences of an accident previously evaluated; (2) 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated; and (3) does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 21, 1999.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.10.C, ``Diesel Fuel'' by 
increasing the minimum usable volume of diesel fuel in the diesel fuel 
oil storage tank (FOST). The specified minimum amount of diesel fuel is 
that quantity necessary to support diesel generator operation for a 
period of 7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The diesel generators are used to support mitigation of the 
consequences of an accident; however, they are not considered the 
initiator of any previously analyzed accident. This change does not 
challenge or degrade the performance of any safety system assumed to 
function in the accident analysis. Since this change simply 
increases the minimum volume of stored diesel generator fuel in the 
FOST, its impact is to enhance the long-term operation of diesel 
generators used to mitigate the consequences of accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    This change does not affect the design or mode of operation of 
any plant system, structure or component. No physical alteration of 
plant structures, systems or components is involved, and no new or 
different type of equipment will be installed. Thus, no new 
condition of operation is created. The change is conservative in 
that it results in a net increase in the minimum required diesel 
fuel oil stored in the FOST.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated for Vermont Yankee.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    The[ ] proposed change does not adversely affect a margin of 
safety because increasing the minimum required volume of fuel oil 
provides additional assurance of diesel generator availability and, 
therefore, maintains or increases the availability of the onsite 
power supply. Since this change simply increases the quantity of 
diesel fuel oil available for diesel generator operation, there is 
no reduction in any value, condition, or range of parameters used in 
any accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

[[Page 56538]]

    NRC Section Chief: James W. Clifford.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: September 21, 1999.
    Description of amendment request: The proposed amendment would 
extend the effective full implementation date by six months, from 
December 31, 1999, to June 30, 2000, for Amendment 120 issued March 22, 
1999. Amendment 120 approved a modification to the plant to increase 
the storage capacity of the spent fuel pool and increase the nominal 
fuel enrichment to 5 weight percent U-235. The extension is due to 
delays fabricating and installing the new spent fuel storage racks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is administrative in nature and does not 
significantly affect any system that is a contributor to initiating 
events for previously evaluated accidents. The proposed change does 
not significantly affect any system that is used to mitigate any 
previously evaluated accidents. Therefore, the proposed change does 
not involve any significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change is administrative in nature and does not 
alter the design, function, or operation of any plant component and 
does not install any new or different equipment. Therefore, a 
possibility of a new or different kind of accident from those 
previously analyzed has not been created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative in nature and does not 
involve a significant reduction in the margin of safety associated 
with the fuel cladding, reactor coolant boundary, containment, or 
any safety limit.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC 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.

Consolidated Edison Company of New York, Docket No. 50-003, Indian 
Point Nuclear Generating Station, Unit No. 1, Buchanan, New York

    Date of amendment request: July 20, 1999.
    Description of amendment request: The amendment would revise the 
Technical Specifications to change the senior reactor license 
requirement for the Operations Manager.
    Date of publication of individual notice in Federal Register: 
September 9, 1999 (64 FR 49027).
    Expiration date of individual notice: October 12, 1999.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York.

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

    Date of amendment request: September 24, 1999.
    Description of amendment request: The proposed amendment would 
revise current Technical Specification (TS) 3.6.1.8 by adding footnote 
``**'' to Action b. The footnote would allow continued operation of 
Fermi 2 with the leakage of penetration X-26 exceeding the limit in TS 
4.6.1.8.2, provided certain compensatory measures are taken. Operation 
would be allowed to continue until the next plant shutdown.
    Because the NRC staff issued the Fermi 2 improved standard TSs 
(ITS) on September 30, 1999, with implementation within 90 days, the 
licensee also provided a version of the TS amendment that would be 
compatible with the ITS. This version would add a new special 
operations TS, ITS 3.10.8, to address the compensatory actions and 
other requirements associated with penetration X-26.
    Date of publication of individual notice in Federal Register: 
October 1, 1999 (64 FR 53421).
    Expiration date of individual notice: Comment period expires 
October 15, 1999; Opportunity for hearing period expires November 1, 
1999.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

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

[[Page 56539]]

Evaluation and/or Environmental Assessment as indicated. All of these 
items are available for public inspection at the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
and at the local public document rooms for the particular facilities 
involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: May 20, 1999, as supplemented 
by letters dated September 8, 1999, September 16, 1999, and September 
20, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) Section 3.8.A, ``Containment Cooling Service Water 
System,'' (CCSW) to clarify that only one pump is required to support 
operability of the Control Room Emergency Ventilation System (CREVS).
    Date of issuance: October 1, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 174 and 170.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46426). The September 8, September 16, and September 20, 1999, 
submittals provided additional clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: June 15, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 4.7.D.6 by replacing the leakage limit of 11.5 
standard cubic feet per hour (scfh) for each main steam isolation valve 
(MSIV) with a limit of 46 scfh on the total combined leakage for the 
MSIVs of all four main steam lines.
    Date of issuance: October 1, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 175 and 171.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: May 19, 1999.
    Brief description of amendments: The amendments relocated Technical 
Specification 3/4.4.4, ``Chemistry,'' from the TS to the Updated Final 
Safety Analysis Report (UFSAR) and to an Administrative Technical 
Requirement that has been incorporated into the UFSAR by reference.
    Date of issuance: October 1, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 134 and 119.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38024).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.

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: May 11, 1999, as supplemented by 
letter dated July 13, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications by incorporating changes to the pressure-
temperature limits; the heatup, cooldown, and inservice test limits for 
the reactor coolant system to a maximum of 33 Effective Full Power 
Years; the low temperature overpressure protection system; and 
operational requirements for the reactor coolant pumps.
    Date of Issuance: October 1, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1-307; Unit 2-307; Unit 3-307.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32289).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of application for amendment: April 7, 1999, as supplemented 
by letters dated May 25, June 21, August 2, and August 30, 1999.
    Brief description of amendment: The amendment revises the minimum 
critical power ratio safety limits.
    Date of issuance: September 27, 1999.
    Effective date: September 27, 1999.
    Amendment No.: 158.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27329).
    The May 25, June 21, August 2 and August 30, 1999, supplemental 
letters provided additional clarifying information that did not expand 
the scope of the application as originally noticed and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of application for amendment: April 20, 1999, as supplemented 
by letter dated September 9, 1999.
    Brief description of amendment: The amendment revised Technical 
Specification 3.4.11, ``RCS Pressure and Temperature (PT) Limits,'' for 
32 effective full power years (EFPY) using the latest vessel beltline 
material and fluence data.
    Date of issuance: October 6, 1999.
    Effective date: October 6, 1999.

[[Page 56540]]

    Amendment No.: 159.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27330).
    The September 9, 1999, supplemental letter provided additional 
clarifying information, did not significantly expand the scope of the 
application as originally noticed and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: May 14, 1999, as supplemented by letters 
dated June 17, and September 7, 15, 17, and 24, 1999.
    Brief description of amendment: The amendment revises the Technical 
Specification requirements affecting the surveillance criteria for that 
portion of the once-through steam generator tubes regarded as a 
primary-to-secondary pressure boundary located within the upper 
tubesheet and impacted by a specific degradation mechanism, namely, 
outside diameter intergranular attack.
    Date of issuance: October 4, 1999.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the Unit 1 Cycle 15 refueling outage.
    Amendment No.: 202.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29709).
    The June 17, and September 7, 15, 17, and 24, 1999, letters 
provided clarifying and additional information that did not change the 
scope of the May 14, 1999, application and the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 2, 1998, as supplemented by letters 
dated July 7 and August 24, 1999.
    Brief description of amendment: The amendment changes the ACTION 
requirements for Technical Specification (TS) 3/4.3.2 for the Emergency 
Feedwater Actuation Signal (EFAS). This change revises the allowed 
outage time for a channel of EFAS to be in the tripped condition from 
``prior to entry into the applicable MODE(S) following the next COLD 
SHUTDOWN'' to the more restrictive time limit of 48 hours and adds a 
shutdown requirement. Additionally, the TS 3.0.4 exemption is removed 
from the ACTION statement for the tripped condition. Changes to TS 
Bases Section 3/4.3.2 are also included to support the changes.
    Date of issuance: October 6, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 154.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69339). The July 7 and August 24, 1999, letters provided additional 
information that did not change the scope of the July 2, 1998, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: February 23, 1999.
    Brief description of amendment: This amendment removes redundant 
boron concentration monitoring requirements specified for Modes 3 
through 6 contained in TS 3/4.1.2.9, ``Reactivity Control Systems-Boron 
Dilution.''
    Date of Issuance: October 4, 1999.
    Effective Date: October 4, 1999.
    Amendment No.: 104.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46440).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: May 5, 1999, as supplemented May 
21, May 28, August 20, and September 2, 1999.
    Brief description of amendment: Changes the Crystal River Unit 3 
Technical Specifications to allow an alternate repair criteria (ARC) 
for axial tube end crack-like indications in the upper and lower 
tubesheets of the Once-Through Steam Generators (OTSGs). The ARC will 
allow leaving OTSG tubes with axially oriented tube end cracks located 
within the clad region of the tube-to-tubesheet roll joint in service.
    Date of issuance: October 1, 1999.
    Effective date: October 1, 1999.
    Amendment No.: 188.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29710). The May 21, May 28, August 20, and September 2, 1999, 
supplements did not affect the original no significant hazards 
consideration determination, or expand the scope of the amendment 
request as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: December 29, 1998, as 
supplemented June 18, 1999.
    Brief description of amendment: Transfer of the license for Crystal 
River Unit 3, to the extent it is held by the City of Tallahassee, to 
Florida Power Corporation.
    Date of issuance: October 1, 1999.
    Effective date: October 1, 1999.
    Amendment No.: 189.
    Facility Operating License No. DPR-31: Amendment revised the 
License.

[[Page 56541]]

    Date of initial notice in Federal Register: February 26, 1999 (64 
FR 9544). The supplemental letter dated June 18, 1999, did not change 
the original proposed no significant hazards consideration 
determination, or expand the scope of the amendment request as 
originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: December 16, 1998.
    Brief description of amendment: The amendment relocates Technical 
Specification (TS) 3/4.7.10 ``Area Temperature Monitoring,'' and the 
associated TS Table 3.7-3, to the Technical Requirements Manual, which 
is referenced in the Seabrook Station Updated Final Safety Analysis 
Report and is the implementing manual for the TS improvement program 
referenced in Section 6.7 of the TSs.
    Date of issuance: October 1, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 63.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6700).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

North Atlantic Energy Service Corporation, et al., Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 27, 1998, as supplemented by 
letter dated June 17, 1998.
    Brief description of amendment: To revise Technical Specification 
(TS) 3.7.6.1, Control Room Emergency Makeup Air and Filtration, and TS 
3.7.6.2, Control Room Air Conditioning, to delete the restriction to 
suspend all operations involving positive reactivity changes during the 
plant conditions specified.
    Date of issuance: October 5, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 64.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19973). The June 17, 1998, supplement provided clarifying information 
and did not change the staff's proposed no significant hazards 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: March 31, 1999.
    Brief description of amendment: The amendment revised Sections 
2.10.4, 3.1, and Table 3-3 of the technical specifications to increase 
the minimum required reactor coolant system (RCS) flow rate and change 
surveillance requirements for RCS flow rate.
    Date of issuance: October 6, 1999.
    Effective date: October 6, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 193.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27322).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

PECO Energy Company, Public Service Electric and Gas Company Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: March 29, 1999, as supplemented 
July 21, 1999.
    Brief description of amendments: The amendments delete the 
surveillance requirement (SR) associated only with the refuel platform 
fuel grapple fully retracted position interlock input, which is 
currently required by the Peach Bottom Atomic Power Station, Units 2 
and 3, Technical Specification SR 3.9.1.1.
    Date of issuance: September 24, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendments Nos.: 229 and 232.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43774). The July 21, 1999, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County, 
Pennsylvania

    Date of application for amendment: July 12, 1999, and supplemented 
August 30, 1999.
    Brief description of amendment: The amendment changed the minimum 
critical power ratio safety limit and the approved methodologies 
referenced in the core operating limits report.
    Date of issuance: October 5, 1999.
    Effective date: As of date of issuance and shall be implemented 
prior to the start of Peach Bottom Atomic Power Station Unit No. 3, 
Cycle 13 operation.
    Amendment No.: 233.
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43777). The August 30, 1999, letter provided additional information but 
did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice.

[[Page 56542]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: September 10, 1998 (PCN-496), 
as supplemented July 19, 1999.
    Brief description of amendments: The amendments delete Technical 
Specification 3.6.7 relating to hydrogen recombiners.
    Date of issuance: October 7, 1999.
    Effective date: October 7, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--159; Unit 3--150.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43778).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: November 6, 1998.
    Brief Description of amendments: The amendments revise the TS 
nuclear instrumentation system (NIS) surveillance requirements. The 
revised TS changes require Southern Nuclear Company to adjust the NIS 
power range channels only when calorimetric-calculated power is greater 
than the power range indicated power by more than +2 percent rated 
thermal power. The proposed TS changes are for both the current TS and 
the improved TS.
    Date of issuance: October 1, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 144 and 135
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4160).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: April 13, 1999, as supplemented 
by letter dated August 26, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications (TS) to update Limiting Condition for Operation (LCO) 
3.0.4 and Surveillance Requirements (SR) 3.0.4 in the existing TS to be 
consistent with the versions of the LCO 3.0.4 and SR 3.0.4 as they 
appear in Revision 1 to NUREG-1431. The proposed change also adds the 
words ``or that are part of a shutdown of the unit,'' to LCO 3.0.4 to 
allow reactor shutdowns that are not necessarily required by other TS 
Required Actions.
    Date of issuance: September 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--108; Unit 2--86.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43779). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 30, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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: July 29, 1999.
    Brief description of amendments: The amendments revise TS Section 
3.1.7, ``Standby Liquid Control (SLC) System.'' The revision replaces 
``greater than the Region B limits,'' which could be misleading, with 
``within the Region B limits.''
    Date of issuance: September 24, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--217; Unit 2--158.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46449). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of application for amendment: July 28, 1999 (TS-398).
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to implement operability and surveillance 
requirements for the previously-installed Oscillation Power Range 
Monitor trip function.
    Date of issuance: September 27, 1999.
    Effective date: As of the date of issuance, to be implemented at 
the end of the Cycle 9 outage.
    Amendment No.: 221.
    Facility Operating License No. DPR-68: Amendment revises the TS.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46450). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: February 26, 1999 (TS 98-08).
    Brief description of amendments: The amendments relocate Sequoyah 
Nuclear

[[Page 56543]]

Plant Technical Specification (TS) 3.7.6, ``Flood Protection Plan,'' 
and its associated bases from the TS to the Technical Requirements 
Manual. Future changes to the Flood Protection Plan will be processed 
in accordance with 10 CFR 50.59.
    Date of issuance: October 6, 1999.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 247 and 238.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14286) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 6, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: July 20, 1999, as supplemented 
August 13, 1999.
    Brief description of amendment: The amendment modifies the 
operability requirements for the high pressure cooling systems--High 
Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling 
(RCIC), and Automatic Depressurization System (ADS)--and the safety and 
relief valves, and adds a time limitation for conducting operability 
testing of HPCI and RCIC.
    Date of Issuance: October 1, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 177
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1999 (64 FR 
47537)
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Power Station, Vernon, Vermont
    Date of application for amendment: June 29, 1999 Brief description 
of amendment: The amendment revises the leak rate requirements for the 
main steam line isolation valves. Specifically, a total allowable 
leakage rate for the sum of the four main steam lines is established 
that is equal to four times the current allowable individual main steam 
line isolation valve leakage rate. The allowable individual main steam 
line isolation valve leakage rate is revised to be one half of the 
allowable total leakage rate.
    Date of Issuance: October 1, 1999.
    Effective date: 10/01/99, and shall be implemented within 30 days.
    Amendment No.: 178
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40909).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 1, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

    For the Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 13th day of October, 1999.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-27210 Filed 10-19-99; 8:45 am]
BILLING CODE 7590-01-P