[Federal Register Volume 65, Number 154 (Wednesday, August 9, 2000)]
[Notices]
[Pages 48744-48767]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-20014]


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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 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 July 15, 2000, through July 28, 2000. The 
last biweekly notice was published on July 26, 2000.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By September 8, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). 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

[[Page 48745]]

or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: June 5, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.8 to change the Required 
Actions and Completion Times for the Ultimate Heat Sink (UHS) in the 
event the service water (SW) temperature exceeds the 97 deg.F 
surveillance acceptance limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The CP&L conclusion is 
in accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change 
provides Required Actions for the plant condition where SW 
temperature exceeds the TS limit. The SW system temperature is not 
assumed to be an initiating condition of any accident analysis 
evaluated in the safety analysis report (SAR). Therefore, the 
revised limitations for SW temperature to be in excess of the design 
limit does not involve an increase in the probability of an accident 
previously evaluated in the safety analysis report. The SW system 
supports operability of safety-related systems used to mitigate the 
consequences of an accident. Plant equipment has been analyzed and 
determined able to perform its safety-related function at [an] SW 
temperature of 99 deg.F. Performance of the containment has been 
analyzed in support of Amendment No. 187 to Technical Specifications 
assuming 100 deg.F service water temperature and the results were 
acceptable. The magnitude of any increase in SW temperature in 
excess of the TS limit is expected to be small based on historical 
data and experience for the UHS. An evaluation would be performed to 
assure required cooling capability. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated in the SAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The temperature of the SW 
when near or slightly above the design temperature does not 
introduce new failure mechanisms for systems, structures or 
components not already considered in the SAR. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not allow continued operation with the 
SW temperature above the design basis limit. The proposed change 
will allow continued operation provided the required cooling 
capacity is verified and periodic monitoring is invoked to verify 
the SW temperature remains less than or equal to 99 deg.F. Design 
margins are affected which are associated with systems, structures 
and components which are cooled by the SW system, and system 
temperature is an input assumption for mitigating the effects of a 
DBA [design-basis accident]. However, allowing SW temperature to 
exceed the surveillance acceptance limit, as long as required 
cooling is verified, will not significantly reduce the margin of 
safety associated with this proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602 .
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: May 31, 2000.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to delete the requirement to 
remove the Reactor Protection System (RPS) circuitry shorting links 
from TS Section 3/4.3.1, ``Reactor Protection System Instrumentation,'' 
3/4.9.2, ``Refueling Operations Instrumentation,'' and 3/4.10.3, 
``Shutdown Margin Demonstrations,'' and to increase the required 
signal-to-noise ratio for the source range monitor in (SRM) TS Sections 
3/4.3.7.6, ``Source Range Monitors,'' and 3/4.9.2.

[[Page 48746]]

    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:

    Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes to TS Section 3/4.3.1, 3/4.9.2, and 3/
4.10.3 will relocate the requirement that the shorting links be 
removed from the RPS circuitry prior to and during specified plant 
conditions. The removal or installation of the RPS circuitry 
shorting links does not have an effect on the probability of any 
accident previously evaluated. The proposed changes to TS Sections 
3/4.3.7.6 and 3/4.9.2 will increase the minimum signal-to-noise 
ratio from  2:1 to  20:1, when the SRM count 
rate is greater than or equal to 0.7 counts per second (cps) and 
less than 3 cps.
    The operation of the SRM does not have an effect on the 
probability of any accident previously evaluated. Thus, the 
probability of any accident previously evaluated is not increased.
    The proposed changes do not affect the integrity of the fuel 
cladding, reactor coolant system or secondary containment, because 
no credit is taken in the current accident analyses for removal of 
the RPS circuitry shorting links. Thus, the radiological 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not affect the assumed accident 
performance of any LaSalle County Station structure, system or 
component previously evaluated because accidents previously 
evaluated assumed that the RPS circuitry shorting links were 
installed and did not credit SRM operation. The proposed changes do 
not introduce any new modes of system operation or failure 
mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to TS Sections 3/4.3.1, 3/4.9.2, and 3/
4.10.3 will relocate the requirement that the shorting links be 
removed from the RPS circuitry prior to and during specified plant 
conditions. The removal of the RPS circuitry shorting links in 
Operations Condition 5, ``Refueling,'' modifies the RPS by 
reconfiguring the scram signal for the intermediate range monitors 
(IRMs) and average power range monitors (APRMs) to non-coincidental 
and enabling the SRM non-coincidental high flux scram signal. 
However, the SRM non-coincidental high flux scram signal is not 
credited in any Design Basis Accident (DBA) and the IRM and APRM 
one-out-of-two taken twice full scram provides the credited 
protection with respect to safety analysis.
    Refueling interlocks and shutdown margin requirements ensure 
that the reactor is maintained in a subcritical condition in 
Operational Condition 5. The refueling interlocks are required to be 
operable by TS Section 3/4.9.1, ``Reactor Mode Switch.'' The SRM, 
IRM, and APRM control rod withdrawal block interlocks are not 
affected by the removal or installation of the RPS circuitry 
shorting links. Although shutdown margin may not yet have been 
demonstrated in Operational Condition 5, shutdown margin 
calculations performed prior to altering the reactor core, along 
with procedural compliance for any Core Alterations, provides 
indication that shutdown margin is available.
    The proposed changes to relocate the description and function of 
the RPS circuitry shorting links to the UFSAR and be controlled in 
accordance with the requirements of 10 CFR 50.59, are consistent 
with the requirements of 10 CFR 50.36, ``Technical Specifications.'' 
The existing TS requirements to remove the RPS circuitry shorting 
links do not satisfy any of the four criteria of 10 CFR 50.36 for 
inclusion of a requirement into the TS. In accordance with NRC 
guidance, existing TS requirements that do not satisfy the criteria 
of 10 CFR 50.36 can be removed from the TS and relocated to other 
controlled documents, such as the UFSAR. Changes to the LaSalle 
County Station UFSAR are controlled in accordance with the 
requirements of 10 CFR 50.59.
    The proposed changes to TS Sections 3/4.3.7.6 and 3/4.9.2 will 
increase the statistical neutron monitoring confidence that the 
indicated signal is correct when the SRMs indicate in the range form 
0.7 cps to 3 cps. A SRM signal-to-noise ratio of  2:1 
provides a statistical neutron monitoring confidence of 95% that the 
indicated signal is correct with a minimum count rate of 3 cps. A 
study was performed which concluded that a SRM signal-to-noise ratio 
of 20:1 is required to provide a statistical neutron 
monitoring confidence of 95% that the indicated signal is correct at 
0.7 cps.
    Thus, the proposed changes do 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

 Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: December 27, 1999.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to increase the allowable out-of-
service times and surveillance test intervals for selected actuation 
instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed TS [technical specification] changes increases the 
Allowable Outage Times and Surveillance Test Intervals (AOT/STI) for 
actuation instrumentation based on analyses developed and approved 
by the Nuclear Regulatory Commission (NRC). TS requirements that 
govern operability or routine testing of plant instruments are not 
assumed to be initiators of any analyzed event because these 
instruments are intended to prevent, detect, or mitigate accidents. 
Therefore, these changes will not involve an increase in the 
probability of occurrence of an accident previously evaluated. 
Additionally, these changes will not increase the consequences of an 
accident previously evaluated because the proposed changes do not 
involve any physical changes to plant systems, structures or 
components (SSCs), or the manner in which these SSCs are operated. 
These changes will not alter the operation of equipment assumed to 
be available for the mitigation of accidents or transients by the 
plant safety analysis or licensing basis. As justified and approved 
in the AOT/STI licensing topical reports, the proposed changes 
establish or maintain adequate assurance that components are 
operable when necessary for the prevention or mitigation of 
accidents or transients and that plant variables are maintained 
within limits necessary to satisfy the assumptions for initial 
conditions in the safety analyses. The proposed changes establish or 
modify time limits allowable for operation with inoperable 
instrument channels based on analyses which have been approved by 
the NRC. Furthermore, there will be no change in the types or 
significant increase in the amounts of any effluents released 
offsite. For these reasons, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical changes to 
SSCs, or the manner in which these SSCs function. Therefore, these 
changes will not create the possibility of a

[[Page 48747]]

new or different kind of accident from any accident previously 
evaluated. The changes in methods governing normal plant operation 
are consistent with the current safety analysis assumptions. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes increase the STIs and AOTs for actuation 
instrumentation based on generic analyses completed by the Boiling 
Water Reactor Owners' Group (BWROG). The NRC has reviewed and 
approved the generic studies and has concurred with the BWROG that 
the proposed changes do not significantly affect the probability of 
failure or availability of the affected instrumentation systems. The 
analysis determined that there is no significant change in the 
availability and/or reliability of instrumentation as a result of 
the proposed changes in STIs and AOTs. Furthermore, the change to 
increase the frequency of the reactor protection system scram 
contactor testing has been shown to improve plant safety. ComEd has 
determined these studies are applicable to Quad Cities Nuclear Power 
Station, Units 1 and 2. The proposed changes to AOTs provide 
realistic times to complete required testing and maintenance actions 
without increasing the overall instrument failure frequency. 
Likewise, the extended STIs do not result in significant changes in 
the probability of instrument failure. Furthermore, the proposed 
changes will reduce the probability of test-induced plant transients 
and equipment failures. Therefore, it is concluded that the proposed 
changes will not result in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: December 30, 1999.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to (1) remove the Main Steam Line 
Radiation Monitor (MSLRM) scram and main steam line isolation 
functions, and (2) add a new requirement for the MSLRM mechanical 
vacuum pump trip function.
    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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This proposed change involves the removal of existing Main Steam 
Line Radiation Monitor (MSLRM) scram and the MSLRM MSL [main steam 
line] Valve closure signal. The purpose of the MSLRM reactor scram 
and the MSL isolation signal is to mitigate the radiological effects 
of a fuel element failure. These functions do not serve as 
initiators for any of the accidents evaluated in Chapter 15 of the 
Updated Final Safety Analysis Report (UFSAR). Removal of these 
functions will not increase the probability of any of the accidents 
previously evaluated.
    The radiological effects of a Control Rod Drop Accident (CRDA) 
have been evaluated for the Boiling Water Reactor Owners' Group 
(BWROG) by General Electric (GE) in Report NEDO-31400A, ``Safety 
Evaluation For Eliminating the Boiling Water Reactor Main Steam 
Isolation Valve Closure Function and Scram Function of the Main 
Steam Line Radiation Monitor.'' The GE report was evaluated by the 
NRC and found acceptable by letter dated May 15, 1991, ``Acceptance 
for Referencing of Licensing Topical Report NEDO-31400.'' The NRC 
Safety Evaluation Report accepting the GE report required licensees 
to demonstrate that the assumptions of the GE report analysis were 
bounding for their plants. ComEd has evaluated the GE analysis for 
applicability to Quad Cities Nuclear Power Station, Units 1 and 2.
    The GE analysis demonstrates that operation with the proposed 
change does not represent a significant increase in the consequences 
of a CRDA. Therefore, operation of Quad Cities Nuclear Power 
Station, Units 1 and 2, under the proposed change does not represent 
a significant increase in the probability or consequences of an 
accident previously evaluated. A site specific radiological 
evaluation was completed to confirm the applicability of the generic 
GE analysis to Quad Cities Nuclear Power Station.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change involves the removal of the existing MSLRM 
scram and the MSL Valve closure input from the MSL Tunnel High 
Radiation signal. Removal of these functions does not represent a 
change in operating parameters for Quad Cities Nuclear Power 
Station, Units 1 and 2. Removal of these functions does not add any 
additional hardware and does not represent any new failure modes. 
Operation of Quad Cities Nuclear Power Station, Units 1 and 2, under 
the proposed change does not create the possibility of a new or 
different type of accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change involves the elimination of the MSLRM scram 
and the MSL Valve closure input from the MSL Tunnel High Radiation 
signal. Operation under the proposed change will not change any 
plant operation parameters, nor any protective system setpoints 
other than removal of these functions. The GE report has 
demonstrated that the consequences of the CRDA without the MSLRM 
High scram and MSL Valve closure signal from the MSL Tunnel 
Radiation detector results in doses which are well within 10 CFR 
part 100, ``Reactor Site Criteria,'' limits. Therefore, the proposed 
change does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: June 5, 2000.
    Description of amendment request: The proposed amendment, which 
changes the Perry Nuclear Power Plant as described in the Updated 
Safety Analysis Report, modifies the circuitry to the Reactor Core 
Isolation Cooling (RCIC) System initiation logic. The proposed circuit 
modification will include a time delay to the main turbine and 
feedwater pump turbine trip signal associated with a RCIC system 
automatic initiation. The addition of this time delay will prevent 
potential main turbine and feedwater pump turbine trips that result in 
unnecessary reactor scrams from inadvertent RCIC initiations.
    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 Reactor Core Isolation Cooling (RCIC) initiation turbine 
trip circuit performs an operational protection of the main turbine 
for commercial and reliability purposes. The proposed modification 
slightly alters the methodology by which the turbine protective 
features are performed but they have no

[[Page 48748]]

influence on any of the accidents previously evaluated. The 
associated circuits do not interfere with higher priority protection 
systems.
    Installation of circuits associated with the proposed 
modification cannot initiate an accident, nor are they used to 
mitigate the consequences of any previously defined accident. Their 
function is to provide turbine protection that is separate and 
distinct from the turbine overspeed protection system. The circuits 
modified by this modification will still result in actions taken 
(auto or manual) that meet the bases for the present design. Also, 
this modification does not alter or adversely affect the turbine 
overspeed function in any manner.
    The proposed modification reduces the probability of occurrence 
of spurious turbine trips due to spurious RCIC initiation. 
Therefore, with the implementation of this modification, the 
boundaries of the accident analysis will be less challenged and 
result in fewer false scrams.
    The proposed modification provides assurance for compliance with 
the current licensing basis regarding dose limits of General Design 
Criteria (GDC) 19 of Appendix A to 10 CFR [Part] 50 and 10 CFR 
[Part] 100. The proposed modification ensures originally stated 
design criteria are met and therefore does not affect the precursors 
for accidents or transients analyzed in Chapter 15 of the Perry 
Nuclear Power Plant (PNPP) Updated Safety Analysis Report (USAR). 
With the proposed modification, the radiological consequences are 
the same as previously stated in the USAR. Therefore, the 
implementation of the proposed modification does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The USAR addresses accident analysis of the reactor based on 
events such as turbine trips, including spurious trips and turbine 
missiles. The present RCIC initiation turbine trip circuit is a 
potential contributor to spurious turbine trips. The addition of the 
time delay relay reduces this potential. A time delay relay failure 
that fails to trip the turbine would have the same effect on the 
turbine as the failure of the present trip circuit that has no time 
delay relay. The consequence of the failure of this circuit to 
protect the turbine remains unchanged with the addition of a time 
delay relay and is bounded by the existing accident analysis. The 
accident analysis for missile protection of those systems, 
structures, components required for the safe shutdown of the plant 
remain unchanged.
    The probability of external missile generation has not changed 
with implementation of the proposed modification. The Main Turbine 
casing and surrounding structures will not be changed by the 
proposed modification. The location of equipment important to safety 
as it relates to the turbine missiles will not be changed. Therefore 
the missile strike probability will not be increased by the 4\1/2\ 
minute time delay.
    The proposed modification provides assurance for compliance with 
the current licensing basis regarding dose limits of GDC 19 of 
Appendix A to 10 CFR [Part] 50 and 10 CFR [Part] 100. The proposed 
modification does not change the assumptions used in any accident 
analysis and no new or different kind of accident is created. The 
proposed modification ensures originally stated design criteria are 
met and therefore does not affect the precursors for accidents or 
transients analyzed in Chapter 15 of the PNPP USAR. Therefore, the 
implementation of the proposed modification does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety by which this modification is evaluated 
against is the design/criteria of the turbine overspeed protective 
system relative to the PNPP USAR, SER, GDC4 and Reg[ulatory] Guide 
1.115, [``Protection Against Low-Trajectory Turbine Missiles.''] The 
change in response time of the main turbine RCIC initiation trip 
circuit does not affect the margin of safety as reflected in these 
documents. There is no safety margin criteria associated with this 
circuit, as defined in the USAR or the bases for any Technical 
Specifications.
    Although there is no margin of safety associated with the 
turbine, the regulatory requirement for acceptance of the turbine 
for use at PNPP is based upon a calculated value of probability of 
external turbine missile interaction with safety related equipment.
    The barriers (Turbine casing and surrounding structures) and 
barrier interaction as previously analyzed will not be changed by 
this modification. The location of safety related equipment as it 
relates to the turbine missiles will not be changed. The probability 
of external missile generation has not changed with implementation 
of the proposed modification. Therefore, there is no reduction in 
the margin of safety by the proposed modification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 19, 2000.
    Description of amendment request: To revise the license: (1) to 
implement Siemens Power Corporation (SPC) high thermal performance 
(HTP) fuel assembly design in Cycle 17, (2) relocate shutdown margin 
(SDM) requirements in Modes 1 to 5 to the Core Operating Limits Report 
(COLR), (3) update the COLR methodologies listed in the Technical 
Specification (TS) Section 6.9.1.11, and (4) request relief from the 
SPC fuel assembly reconstitution restrictions for peripheral low power 
fuel assemblies. Applicable TS surveillance requirements are changed to 
be consistent with the proposed license amendment. Additionally, 
administrative changes are proposed to the boron concentration 
specifications related to the boration 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) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment would allow the implementation of HTP 
fuel design for Cycle 17. The design of this fuel will be evaluated 
to meet all the mechanical, neutronics and thermal-hydraulics 
requirements, and acceptance criteria based on the approved 
methodology. The relocation of shutdown margin to the COLR and other 
proposed changes have no adverse impact on the operation of the 
plant and have no relevance to the accident initiators. There are no 
changes to the plant configuration, and thus the frequency of 
occurrence of previously analyzed accidents is not affected by the 
proposed changes. The changes proposed to the fuel reconstitution 
methodology would not impact the design acceptance criteria for the 
reconstituted fuel assemblies.
    The proposed change for the relocation of shutdown margin to the 
COLR has no impact on current safety analyses and their 
consequences. Changes to the COLR limits will be controlled per 
Generic Letter 88-16 under the provisions of 10 CFR 50.59 and the 
requirements of TS 6.9.1.11.c. The application of the added 
methodology, which includes the approved HTP DNB [departure from 
nucleate boiling] correlation, would remain consistent with the 
design basis requirements and would not involve a significant 
increase in the consequences of design basis accidents. Other 
proposed TS and TS bases changes do not affect safety analysis 
results. The changes proposed to the fuel reconstitution methodology 
would not impact the safety analysis consequences as the changes are 
related to the non-limiting rod locations.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Use of the modified specification would not create the 
possibility of a new or different

[[Page 48749]]

kind of accident from any previously evaluated.
    The proposed amendment updates the list of approved methodology 
in TS 6.9.1.11, relocates shutdown margin requirements to the COLR 
and requests relief for fuel reconstitution requirements. None of 
these changes would create the possibility of a new kind of accident 
since the reload analysis with these changes would continue to meet 
all applicable design limits. There is no change to plant 
configuration, systems or components which would create new failure 
modes. The modes of operation of the plant would remain unchanged.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The proposed changes have no significant adverse impact on the 
safety analysis. As such, these changes would continue to provide 
margin to the acceptance criteria for specified acceptable fuel 
design limits (SAFDL), 10 CFR 50.46(b) requirements, primary and 
secondary overpressurization, peak containment pressure, potential 
radioactive releases, and existing limiting conditions for 
operation. The future use of updated approved methodologies will 
follow all design basis requirements to ensure that a safety margin 
to the acceptance criteria would continue to remain available for 
full power operation of St. Lucie Unit 1.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company (FPL), Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 19, 2000.
    Description of amendment request: The amendment would revise the 
St. Lucie Unit 1 Technical Specifications (TS) to require laboratory 
testing of activated charcoal samples for applicable engineered safety 
feature ventilation systems using the ASTM D3803-1989 protocol. In 
addition the proposed changes revise the TS test criteria for methyl 
iodide removal efficiency to be consistent with the guidance of NRC 
Generic Letter (GL) 99-02. The affected Unit 1 TS are the shield 
building ventilation system (SBVS), TS 4.6.6.1; control room emergency 
ventilation system (CREVS), TS 4.7.7.1; emergency core cooling system 
(ECCS) area ventilation system, TS 4.7.8.1; and fuel pool ventilation 
system--fuel storage, TS 4.9.12.
    The July 19, 2000, application is a complete replacement of the 
proposed Unit 1 TS amendment previously submitted by FPL letter L-99-
241 on November 17, 1999. The NRC staff had previously published a 
Federal Register notice on January 12, 2000 (Vol. 65, page 1923), 
regarding the proposed amendments for St. Lucie Units 1 and 2, but 
subsequently, issued the licence amendment for St. Lucie, Unit 2 only, 
on February 17, 2000. This revised amendment request increases the TS-
required removal efficiency of the Unit 1 SBVS, ECCS area ventilation 
system, and CREVS charcoal adsorbers to 97.5% when tested in accordance 
with ASTM D3803-1989 at 30 deg.C, 70% relative humidity. The revised 
testing requirements align the TS acceptance criteria and methodology 
with the Unit 1 accident analysis assumptions and GL 99-02 
recommendations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated. The new charcoal testing protocol is performed offsite on 
samples extracted from the safety related ventilation systems. 
Therefore, there is no impact on any accident initiator and results 
in no changes in the probability. The proposed testing protocol is 
more conservative than previous tests; therefore, the efficiency of 
charcoal for the affected safety related systems would not be 
overestimated. With the new testing protocol, more conservative 
testing results are expected since the temperature at which testing 
is performed is lower and the charcoal retention capability is more 
consistent with actual accident conditions. The proposed change thus 
ensures that the charcoal in service will comply with the 
penetration requirements to meet the design basis accident 
conditions.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed charcoal testing protocol only affects 
surveillance testing requirements for safety related ventilation 
systems. The functions of these systems remain unchanged and 
unaffected. No new system interactions have been introduced by the 
proposed amendment, which would create a new or different type of 
accident than previously analyzed. No physical changes are being 
made to any structure, system, or component. The operation of the 
facility will not be altered by the proposed amendment. The systems 
involved are not initiators of any accidents as previously 
evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie Unit 1 
systems. Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment does not involve a reduction in the 
margin of safety. The margin of safety of the Technical 
Specifications, its Bases, the Updated Final Safety Analysis Report, 
the Safety Evaluation Report or in any other design document has 
been increased by the use of a safety factor of two for the TS 
affected by the proposed amendment. The change provided in this 
proposed amendment is related to introducing an improved testing 
protocol for the activated charcoal in safety related ventilation 
systems. The change consists of testing the charcoal with a new 
testing protocol, higher efficiencies, and with lower test 
temperatures to more closely reflect accident conditions and to 
eliminate potential overestimation of charcoal efficiency.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

[[Page 48750]]

Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and 
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: June 21, 2000.
    Description of amendment request: The proposed amendments would 
relocate Technical Specification Surveillance Requirement (SR) 
4.8.1.1.2.e.1 to a licensee controlled maintenance program that will be 
incorporated by reference into the next revision of each unit's Updated 
Final Safety Analysis Report (UFSAR). SR 4.8.1.1.2.e.1 requires that 
the emergency diesel generator (EDG) be inspected in accordance with 
procedures prepared in conjunction with its manufacturer's 
recommendations for this class of standby service, at least once every 
18 months during shutdown. Upon relocation to the licensee controlled 
maintenance program the requirement to perform the EDG inspections 
every 18 months during shutdown will be eliminated. These amendments, 
in combination with the previously submitted EDG risk informed allowed 
outage time extension to 14 days, allows the EDG maintenance to be 
performed in Modes 1 and 2. The licensee stated that approval of these 
amendments is expected to reduce the complexity of activities performed 
during refueling outages and, consequently, reduce human errors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no changes to the emergency diesel generator (EDG) 
maintenance program. The actual EDG maintenance program is 
unaffected.
    The only substantive change allows the periodic EDG inspection 
to be performed in any operational mode instead of only during 
shutdown. By FPL Letter L-99-228, dated November 17, 1999, FPL has 
previously submitted a request for a risk informed EDG allowed 
outage time (AOT) extension from 3 days to 14 days. An evaluation of 
the impact on plant risk as expressed by the change in core damage 
frequency (CDF), the incremental conditional core damage probability 
(ICCDP), the change in large early release frequency (LERF), and the 
incremental conditional large early release probability (ICLERP) was 
provided as part of the EDG AOT extension submittal (L-99-228). The 
EDG downtime (hours/train/year) assumed in the EDG AOT extension 
risk assessment includes the out-of-service time that would be 
incurred due to performing the proposed EDG inspections and 
overhauls in Modes 1 and 2 instead of during shutdown. The risk 
assessment for the proposed EDG AOT extension bounds the risk for 
this change.
    NRC Regulatory Guide (RG) 1.177, An Approach for Plant-Specific 
Risk-Informed Decision making: Technical Specifications, states that 
an ICCDP of 5.0E-07 and an ICLERP of 5.0E-08 is considered small for 
a single AOT change. Both the ICCDP and ICLERP for the proposed EDG 
AOT extension and these proposed changes are below the RG 1.177 
specified values and are thus considered small.
    NRC RG 1.174, An Approach for Using Probabilistic Risk 
Assessment in Decisions on Plant Specific Changes to the Licensing 
Basis, discusses acceptance criteria for changes in CDF and LERF. A 
change in CDF of 1E-06 with a total CDF of 1E-04/year and a change 
in LERF of 1E-07 with a total LERF of 1E-05 are considered very 
small. The changes in CDF and LERF for the EDG AOT extension and 
these proposed changes are below the RG 1.174 criteria and are thus 
considered very small.
    The removal of the Mode restrictions from the maintenance 
program are bounded by the risk assessment for the EDG AOT extension 
and therefore do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The use of the modified specifications cannot create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
implementation of this administrative change since the proposed 
changes do not involve the addition or modification of equipment, 
nor do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components remain unchanged by the proposed 
amendments. Therefore, these changes do not involve a significant 
reduction in the margin of safety. When the full scope of plant risk 
is considered, the risks incurred by performing either corrective or 
preventive EDG maintenance during power operation will be 
substantially offset by plant benefits associated with avoiding 
unnecessary plant transitions and/or reducing risks during shutdown 
operations.
    Based on the above, we have determined that the proposed 
amendments do not (1) involve a significant increase in the 
probability or consequences of an accident previously evaluated, (2) 
create the probability of a new or different kind of accident from 
any previously evaluated, or (3) involve a significant reduction in 
a margin of safety; and therefore does not involve a significant 
hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: May 22, 2000.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to incorporate the 
requirements specified in the American Society of Mechanical Engineers 
(ASME), Section XI, Subsection IWL, as modified and supplemented by the 
requirements in Section 50.55a(b)(2)(viii), Examination of concrete 
containments. In this regard, TS Section 3.6.1.6, ``Limiting Condition 
for Operation,'' will be revised to conform to IWL tendon lift-off 
force requirements, and TS Sections 4.6.1.6.1, 4.6.1.6.2, and 4.6.1.6.3 
will be revised to conform to containment tendon and containment 
surface inspection requirements specified in ASME Section XI, 
Subsection IWL, 1992 Edition with the 1992 Addenda, and 10 CFR 
50.55a(b)(2)(viii).
    The NRC Final Rule (61 FR 41303), dated August 8, 1996, requires 
implementation of the revised requirements for containment examination 
by September 9, 2001. FPL is planning to perform the containment tendon 
surveillance for Turkey Point Units 3 and 4 in March 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 48751]]

    Approval and implementation of this amendment will have no 
effect on the probability or consequences of accident previously 
evaluated. The containment is not an accident initiating system or 
structure; therefore, there will be no impact on any accident 
probabilities by the approval of this amendment. The containment 
examination requirements in the proposed amendments are identical, 
equivalent, or more rigorous than previous requirements. The 
containment serves an important function to mitigate consequences of 
postulated accidents evaluated and the examinations proposed in this 
amendment will not result in a reduction in the capability of the 
containment to meet its intended design function. Additionally, the 
proposed changes to the Technical Specifications reflect the 
adoption of ASME Section XI Subsection IWL containment inservice 
inspections required by 10 CFR 55a(b)(2).
    Based on the above, it is concluded that the proposed amendments 
do not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not alter the design, physical 
configuration, or modes of operation of the plant. No changes are 
being made to the plant that would introduce any new accident causal 
mechanisms. The proposed Technical Specification changes do not 
impact any plant systems that are accident initiators, since the 
containment functions primarily as an accident mitigator and the 
functional requirements of the containment structure are not 
changed. No new accident causal mechanisms are created as a result 
of NRC approval of the proposed amendments request. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments 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, including the 
performance of the containment. The containment is capable of 
performing as intended, and its function is verified by visual 
examination, post-tensioning system examinations, and leakage rate 
testing. The containment examination requirements in the proposed 
amendments are identical, equivalent, or more rigorous than previous 
requirements. As such, the ability of the containment to perform its 
design function will not be impaired by the implementation of the 
proposed amendments request. Therefore, operation of the facility in 
accordance with the proposed amendments would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: July 7, 2000.
    Description of amendment request: The proposed amendments would 
revise the pressure-temperature (P/T) limits specified in Technical 
Specification (TS) 3.4.9.1 and Figures 3.4-2, 3.4-3 and 3.4-4 to extend 
their service period to a maximum of 32 effective full power years. 
Also, the proposed amendments will revise TS 3.4.9.3, Cold Overpressure 
Mitigation System (COMS) setpoints and its associated Surveillance 
Requirements 4.4.9.3.1a and 4.4.9.3.1d. COMS is the Westinghouse 
version of Low Temperature Overpressure Protection. Additionally, the 
licensee's submittal requested two exemptions from the requirements of 
10 CFR 50.60 based on the American Society of Mechanical Engineers 
(ASME) Section XI, Code Cases N-588, ``Alternative to Reference Flaw 
Orientation of Appendix G for Circumferential Welds in Reactor Vessels, 
Section XI, Division 1'' and N-641, ``Alternative Pressure Temperature 
Relationship and Low Temperature Overpressure Protection (LTOP) System 
Requirements, Section XI, Division 1.'' The exemption requests will be 
evaluated separately from the proposed license amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probability of occurrence of an accident previously 
evaluated for Turkey Point is not altered by the proposed amendment 
to the Technical Specifications. Each accident in the Turkey Point 
UFSAR [Updated Final Safety Analysis Report] was examined with 
respect to the changes to the proposed Pressure-Temperature (P/T) 
limit curves and associated Cold Overpressure Mitigation System 
(COMS) setpoint limitations.
    The proposed changes do not impact the integrity of the reactor 
coolant system pressure boundary (i.e., no change in operating 
pressure, materials, seismic loading, etc.) and therefore does not 
increase the potential for the occurrence of a loss of coolant 
accident (LOCA). The changes do not modify the reactor coolant 
system pressure boundary, nor make any physical changes to the 
facility design, material, or construction standards. The 
probability of any design basis accident (DBA) is not affected by 
this change, nor are the consequences of any DBA affected by this 
change. The proposed P/T limit curves and COMS setpoint limit are 
not considered to be an initiator or contributor to any accident 
currently evaluated in the Turkey Point UFSAR.
    The curves and setpoint limit were generated in accordance with 
approved NRC and ASME methodology. Code Cases N-588 and N-641 have 
ASME Code Committee approval.
    Delaying performance of two of the COMS surveillances (PORV 
[power operated relief valve] Channel Operational Test and the 
backup nitrogen supply verification) until 12 hours after decreasing 
the RCS cold leg temperature to 275 deg.F during cooldown 
was also evaluated with respect to the plant accident analyses. The 
change was determined to not represent a significant increase in the 
probability or consequences of an accident because a) the likelihood 
of a low temperature overpressure event occurring concurrently with 
a loss of the redundant instrument air system is sufficiently small, 
and b) the existing procedural controls will effectively prevent 
challenges to the COMS.
    Additionally, delaying these surveillances for 12 hours will 
allow the operators to focus their attention on transitioning the 
plant to RHR [residual heat removal] cooling. Given the timing 
sequence of the RHR system entry point to the COMS enable 
temperature, the time extension is considered to be a prudent and 
safety focused change to the method of performing a plant cooldown. 
The proposed time extension is also consistent with the operational 
flexibility currently provided in NUREG-1431, Standard Technical 
Specifications for Westinghouse Plants.
    Based on the above, it is concluded that the proposed amendment 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not create a new accident scenario. The 
requirements for the P/T limit curves and low temperature 
overpressure protection have been in place for some time. The 
fundamental approach follows approved ASME and Westinghouse topical 
report methodology. The proposed curves reflect the change in 
material properties acknowledged and managed by regulation and an 
upgrade in technology, which has been approved by ASME.

[[Page 48752]]

    Delaying performance of two of the COMS surveillances (PORV 
Channel Operational Test and the backup nitrogen supply 
verification) until 12 hours after decreasing the RCS cold leg 
temperature to 275 deg.F during cooldown was also 
evaluated with respect to the plant accident analyses. The change 
was determined to not represent a significant increase in the 
probability or consequences of an accident because a) the likelihood 
of a low temperature overpressure event occurring concurrently with 
a loss of the redundant instrument air system is sufficiently small, 
and b) the existing procedural controls will effectively prevent 
challenges to the COMS.
    Additionally, delaying these surveillances for 12 hours is 
consistent with the operational flexibility currently provided in 
NUREG-1431, Standard Technical Specifications for Westinghouse 
Plants.
    Since no new failure modes are associated with the proposed 
changes, the activity does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The Technical Specifications for P/T limit curves and COMS 
setpoints are expiring and must be updated. The COMS setpoint is 
revised to incorporate additional margin in the instrument 
uncertainty. Conservative ASME code methods including safety factors 
have been used. The material properties used are from a much larger 
database than in past submittals. This results in many more 
datapoints available for the limiting weld metal than in past 
submittals. A new master curve of irradiated and unirradiated 
materials data has been developed for Turkey Point which shows that 
these curves and associated setpoints are conservative and represent 
an increase to the margin of safety. The new setpoint limit should 
reduce the possibility of an inadvertent PORV actuation. They should 
also reduce the potential for reactor coolant pump impeller 
cavitation or seal damage when the pumps are operated during low 
temperature conditions in the RCS. Changing the COMS surveillances 
to allow completion up to 12 hours after decreasing RCS temperature 
to 275 deg.F during cooldown does not result in a 
reduction in the margin of safety. Acceptability is based on: 
consistency with NUREG-1431, Standard Technical Specifications 
Westinghouse Plants, COT [Channel Operational Test] Surveillance 
Requirements; the inherent reliability and redundancy of the Turkey 
Point Instrument Air System; and the existing procedural controls 
established to prevent challenges to the LTOP System. The proposed 
amendments will not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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: May 30, 2000.
    Description of amendment requests: The proposed amendments would 
make changes to several Technical Specifications (TSs) to reflect 
implementation of the revised 10 CFR Part 20, ``Standards for 
Protection Against Radiation.'' In addition, the licensee proposed to 
revise TS 6.8.4.a.7 to maintain existing instantaneous dose rate 
limitations in the Offsite Dose Calculation Manual. Also, the licensee 
proposed a revision to the requirements governing the annual tabulation 
of radiation exposures.
    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?
    The proposed changes do not physically alter any plant 
structures, systems, or components (SSCs), and do not affect or 
create new accident initiators or precursors for any accident 
evaluated in the Updated Final Safety Analysis Report. Therefore, 
the probability of an accident previously evaluated is unchanged.
    The proposed changes do not affect the types or amounts of 
radionuclides released following an accident, or the initiation and 
duration of their release. The changes are administrative in nature. 
Therefore, the consequences of an accident previously evaluated are 
not increased.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not physically alter any SSC and do not 
affect or create new accident initiators or precursors. The accident 
analysis assumptions and results are unchanged. No new failures or 
interactions have been created.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    10 CFR 20.1301, Appendix I to 10 CFR 50, and 40 CFR 190 
establish the controls and limitations on total effective dose 
equivalent to individual members of the public from effluents 
discharged to unrestricted areas. The proposed changes maintain 
established limits for radioactive liquid effluents established in 
10 CFR Part 20 and limits for radioactive gaseous effluents 
established in the ODCM. I&M continues to comply with limits 
specified in 10 CFR 20.1301, Appendix I to 10 CFR 50, and 40 CFR 
190. Since compliance with these regulatory requirements has not 
been compromised, the proposed changes do not involve a significant 
reduction in the margin of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that the proposed amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: June 20, 2000.
    Description of amendment request: The following technical 
specification (TS) changes are being proposed to provide flexibility of 
operation. These changes include: (1) the ability to have a standby 
Safety Injection (SI) pump available during Reactor Coolant System 
(RCS) reduced inventory conditions with the RCS pressure boundary 
intact; (2) realigning a footnote to clarify the allowance of an 
inoperable SI pump to be energized for testing or filling accumulators; 
(3) allowance for an additional charging pump to be made capable of 
injection during pump-swap operations; (4) recognition that a 
substantial vent area exists for cold overpressure protection when the 
reactor vessel head is on, and the studs are fully detensioned; (5) 
limit maneuvering the plant beyond Hot Shutdown when one charging pump 
is operable; and (6) establishes a new value for the open permissive 
interlock associated with the Residual Heat Removal System suction 
isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against

[[Page 48753]]

the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect plant systems such that their 
function in the control of radiological consequences is adversely 
affected. The proposed changes do not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or manner in which structures, systems, and components 
(SSCs) perform their intended safety function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Since there are no changes to previous accident analyses, the 
radiological consequences associated with these analyses remain 
unchanged; therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accidents previously evaluated.
    The proposed changes do not result in a change to the design 
basis of any plant SSC. All equipment important to safety will 
operate as designed. The proposed TS changes in conjunction with 
administrative controls will provide adequate control measures to 
ensure component integrity is not challenged. The proposed changes 
do not cause the initiation of any accident nor create any new 
failure mechanisms. The changes do not result in any event 
previously deemed incredible being made credible. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not adversely affect equipment design or 
operation and there are no changes being made to the TS-required 
safety limits or safety system settings that would adversely affect 
plant safety. The proposed TS changes in conjunction with 
administrative controls will provide adequate control measures to 
ensure component integrity is not challenged. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    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.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: June 20, 2000.
    Description of amendment request: The licensee proposes revising 
the Technical Specifications (TS) by removing the prescriptive 
requirement for determining the reactor coolant system flow rate by 
precision heat balance in Surveillance Requirement 4.2.5.3 and 
incorporating a time limit for completion of the surveillance 
requirement. The change would also revise TS Table 2.2-1 to reflect the 
allowed calibration tolerance of the protection racks and note that the 
Trip Setpoint for Functional Unit 12, Reactor Coolant Flow-Low reactor 
trip is based on an indicated value rather than a measured value.
    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 involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design, conditions, and configuration of 
the facility or the manner in which the plant is operated. The 
proposed changes do not alter or prevent the ability of structures, 
systems, and components (SSCs) to perform their intended function to 
mitigate the consequences of an initiating event within the 
acceptance limits assumed in the Updated Final Safety Analysis 
Report (UFSAR).
    Determination of RCS [Reactor Coolant System] total flow rate by 
elbow tap P measurement will not subject the reactor core 
to conditions adverse to nuclear safety. The proposed change does 
not affect the source term; containment isolation or radiological 
release assumptions used in evaluating the radiological consequences 
of an accident previously evaluated in the Seabrook Station UFSAR. 
The initial conditions for all accident scenarios modeled are the 
same. Therefore, the consequences of an accident occurring remain 
unchanged.
    The evaluation for use of elbow tap P measurement 
determined that sufficient margin exists to account for all 
reasonable instrument uncertainties, therefore no changes to 
installed equipment or hardware in the plant are required. Though 
the calibration process of the elbow tap P transmitters has 
changed, i.e., normalization to previously performed precision RCS 
flow calorimetrics for Cycles 1 and 2 instead of normalization to a 
precision RCS flow calorimetric each cycle, this has been accounted 
for by the addition of instrument uncertainties usually considered 
to be zeroed out by normalization performed each cycle. Accounting 
for the additional instrument uncertainties yields a flow 
uncertainty that is slightly less (2.3 percent) than the current NRC 
[Nuclear Regulatory Commission] licensed value (2.4 percent), thus 
no change is required to the nominal reactor trip setpoint for RCS 
flow. The proposed change has no adverse affect on component or 
system interactions. Therefore, the proposed changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    4. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the design, conditions and 
configuration of the facility or the manner in which the plant is 
operated and maintained in a state of readiness. Existing system and 
component redundancy is not being changed by the proposed changes. 
Though the calibration process of the elbow tap P 
transmitters has changed, i.e., normalization to previously 
performed precision RCS flow calorimetrics for Cycles 1 and 2 
instead of normalization to a precision RCS flow calorimetric each 
cycle, this has been accounted for by the addition of instrument 
uncertainties usually considered to be zeroed out by normalization 
performed each cycle. Accounting for the additional instrument 
uncertainties yields a flow uncertainty that is slightly less than 
the current NRC licensed value, thus no change is required to the 
nominal reactor trip setpoint for RCS flow. The proposed change has 
no adverse affect on component or system interactions. The time of 
reactor trip remains the same. Therefore, since there are no changes 
to the design, conditions, configuration of the facility, or the 
manner in which the plant is operated and maintained in a state of 
readiness, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not adversely affect equipment design or 
operation and there are no changes being made to the Technical 
Specification required safety limits or safety system settings that 
would adversely affect plant safety. The additional instrument 
uncertainties resulting from use of elbow tap P 
transmitters without the requirement to normalize to a precision RCS 
flow calorimetric each cycle have been accounted for and no change 
in the nominal Trip Setpoint is required. The calculated instrument 
uncertainty is 2.3 percent flow. This uncertainty is slightly less 
than the current licensed value of 2.4 percent flow. The time of 
reactor trip, as modeled in the various safety analyses, is 
maintained. 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

[[Page 48754]]

involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: February 22, 2000
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) are associated with radiological 
effluent. The proposed changes will relocate selected radiological 
effluent TSs and the associated Bases to the Millstone Radiological 
Effluent Monitoring and Offsite Dose Calculation Manual in accordance 
with the Nuclear Regulatory Commission's (NRC) Generic Letter 89-01.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed changes and has concluded that 
they do not involve a Significant Hazards Consideration (SHC). The 
basis for this conclusion is that the three criteria of 10 CFR 
50.92(c) are not compromised. The proposed changes do not involve an 
SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The purpose of the Radiological Liquid and Gaseous Effluent 
Monitoring Instrumentation is to monitor routine radioactive 
releases. [This] instrumentation provide[s] a surveillance of 
potential release points and initiates automatic alarm and trip 
functions which will terminate the release prior to exceeding the 
limits of 10 CFR Part 20 (1993 version). Relocation of Technical 
Specification 3.3.3.9, ``Radioactive Liquid Effluent Monitoring 
Instrumentation,'' and Technical Specification 3.3.3.10, 
``Radioactive Gaseous Effluent Monitoring Instrumentation,'' to the 
Radiological Effluent Monitoring and Offsite Dose Calculation Manual 
(REMODCM) does not imply any reduction in its importance in 
monitoring routine radioactive releases. These instruments are 
neither used for, nor capable of, detecting a significant abnormal 
degradation of the reactor coolant pressure boundary before a design 
basis accident, nor do they function as a primary success path to 
mitigate events which assume a failure of or a challenge to the 
integrity of fission product barriers. These monitors are not an 
active design feature needed to preclude analyzed accidents or 
transients. Therefore, this change will not significantly increase 
the probability or consequences of an accident previously evaluated.
    Technical Specification 3.11.1.1 ensure[s] the concentration of 
radioactive materials released in liquid waste effluents from the 
site will be less than the concentration levels specified in 10 CFR 
Part 20 (1993 version), Appendix B, Table II. Technical 
Specification 3.11.1.2 ensures the dose or dose commitment from 
radioactive materials released in liquid waste effluents will not 
exceed the requirements of Sections II.A, III.A and IV.A of Appendix 
I, 10 CFR Part 50. Technical Specification 3.11.2.1 ensures the dose 
rate from gaseous effluents released from all units on site will be 
less than dose limits specified in 10 CFR Part 20 (1993 version), 
Appendix B, Table II. Technical Specification 3.11.2.2 ensures the 
dose from noble gases released in gaseous effluents will not exceed 
the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 
CFR Part 50. Technical Specification 3.11.2.3 implements the 
requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR 
Part 50. Technical Specification 3.11.3 ensures the reporting 
requirements of 40 CFR 190 are met. Relocation of these Technical 
Specifications to REMODCM does not imply any reduction in its 
importance in ensuring that the regulatory limits are met. The 
instrumentation covered by these Technical Specifications [is] 
neither used for, nor capable of, detecting a significant abnormal 
degradation of the reactor coolant pressure boundary before a design 
basis accident, nor [does it] function as a primary success path to 
mitigate events which assume a failure of or a challenge to the 
integrity of fission product barriers. [This] instrumentation [is] 
not an active design feature needed to preclude analyzed accidents 
or transients. Therefore, this change will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    As a result of the relocation of the Radiological Effluent 
Technical Specifications (RETS) to the REMODCM, there are no 
Technical Specifications remaining that use definitions 1.31 and 
1.26, ``Radiological Effluent Monitoring and Offsite Dose 
Calculation Manual (REMODCM),'' of Unit Nos. 2 and 3 respectively. 
The guidelines and procedures addressing the use of radioactive 
waste treatment systems are covered by Specifications 6.15 and 6.13 
of unit Nos. 2 and 3 respectively, which describes the REMODCM. 
Therefore, definitions 1.33 and 1.25, ``Radioactive Waste Treatment 
Systems,'' of Unit Nos. 2 and 3 respectively are no longer needed. 
In addition, there are no Specifications that use this phrase in the 
context of a defined term. These changes do not impact the 
assumptions used in any accident analysis, affect plant equipment, 
plant configuration, or the way the plant is operated. Therefore, 
this change will not significantly increase the probability or 
consequences of an accident previously evaluated.
    Replacing Technical Specification 6.9.1.6 of Millstone Unit No. 
2 with Technical Specifications 6.9.1.6a and 6.9.1.6b and revising 
Technical Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No. 3 
will provide descriptions which satisfy the requirements of parts 10 
CFR 50.36a and 10 CFR 50, Appendix I, Sections IV.B.1, IV.B.2, 
IV.B.3, and IV.C. These changes are consistent with NUREG-1432 and 
NUREG-1431. These changes do not impact the assumptions used in any 
accident analysis, affect plant equipment, plant configuration, or 
the way the plant is operated. Therefore, this change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The description of the REMODCM contained in Technical 
Specifications 6.15 and 6.13 of Millstone Unit Nos. 2 and 3 
respectively will be modified to be consistent with the guidance of 
GL 89-01, and with NUREG-1432 and NUREG-1431. Additional minor 
changes have been made to be consistent with the proposed changes to 
Technical Specification 6.9.1.6 of Millstone Unit No. 2 and 
Technical Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No. 
3. These changes do not impact the assumptions used in any accident 
analysis, affect plant equipment, plant configuration, or the way 
the plant is operated. Therefore, this change will not significantly 
increase the probability consequences of an accident previously 
evaluated.
    Adding Technical Specifications 6.20 and 6.15, Radiological 
Effluent Controls Program, to Millstone Unit Nos. 2 and 3 
respectively, and 6.21 and 6.16, Radiological Environmental 
Monitoring Program, to Millstone Unit Nos. 2 and 3 respectively is 
consistent with the guidance contained in Generic Letter 89-01 for 
the relocation of the Radiological Effluents Technical 
Specifications and with NUREG-1432 and NUREG-1431. Additional minor 
changes have been made to be consistent with the version of 10 CFR 
20, Appendix B, Table II, Column 1 which is being used by Millstone 
Unit Nos. 2 and 3, namely the 1993 version. These changes do not 
impact the assumptions used in any accident analysis, affect plant 
equipment, plant configuration, or the way the plant is operated. 
Therefore, this change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    The following proposed changes are administrative in nature. 
Therefore, these changes will not significantly increase the 
probability or consequences of an accident previously evaluated.
     Revise Index Pages of Unit Nos. 2 and 3 Technical 
Specifications to reflect the proposed changes to relocate the RETS 
to the REMODCM.
     Address additional changes to the Millstone Unit No. 2 
Technical Specifications to resolve issues not related to 
transferring the RETS to the REMODCM.
     Relocate to the associated Bases sections.
    The proposed changes do not alter how any structure, system, or 
component functions. There will be no effect on equipment important 
to safety. The proposed changes have no effect on any of the design 
basis accidents previously evaluated. Therefore, this License 
Amendment Request

[[Page 48755]]

does not impact the probability of an accident previously evaluated, 
nor does it involve a significant increase in the consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Relocation of Technical Specifications 3.3.3.9, 3.3.3.10, 
3.11.1.1, 3.11.1.2, 3.11.2.1, 3.11.2.2, 3.11.2.3, and 3.11.3 to 
REMODCM does not imply any reduction in its importance in monitoring 
and ensuring that the regulatory limits are met. As a result of the 
relocation of the RETS to the REMODCM, there are no Technical 
Specifications remaining that use definitions 1.31 and 1.26. 
Additionally, the guidelines and procedures addressing the use of 
radioactive waste treatment systems which are covered by 
Specifications 6.15 and 6.13 remove the need for definitions 1.33 
and 1.25 of Unit Nos. 2 and 3 respectively. Replacing Technical 
Specification 6.9.1.6 of Millstone Unit No. 2 with Technical 
Specifications 6.9.1.6a and 6.9.1.6b and revising Technical 
Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No. 3 will 
provide descriptions which satisfy the requirements of parts 10 CFR 
50.36a and 10 CFR 50, Appendix I, Sections IV.B.1, IV.B.2, IV.B.3, 
and IV.C. Modifying the description of the REMODCM contained in 
Technical Specifications 6.15 and 6.13 of Millstone Unit Nos. 2 and 
3 respectively and adding Technical Specifications 6.20, 6.21 and 
6.15, 6.16 to Millstone Unit Nos. 2 and 3 respectively is consistent 
with the guidance contained in Generic Letter 89-01 for the 
relocation of the Radiological Effluents Technical Specifications 
and with NUREG-1432 and NUREG-1431.
    The proposed changes do not affect any of the assumptions used 
in the accident analysis, nor do they affect any operability 
requirements for equipment important to plant safety. Therefore, the 
proposed changes will not result in a significant reduction in the 
margin of safety as defined in the Bases for Technical 
Specifications covered in this License Amendment Request.
    As described above, this License Amendment Request does not 
involve a significant increase in the probability of an accident 
previously evaluated, does not involve a significant increase in the 
consequences of an accident previously evaluated, does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, and does not result in a significant 
reduction in a margin of safety. Therefore, NNECO has concluded that 
the proposed changes do not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: June 26, 2000.
    Description of amendment request: The proposed changes to 
Millstone, Unit 3, Technical Specifications (TS) revise TS Section 
1.13, Definitions, ``Engineered Safety Features Response Time'', TS 
Section 1.28, ``Reactor Trip System Response Time,'' TS Section 3.3.1, 
``Instrumentation--Reactor Trip System Instrumentation,'' and TS 
Section 3.3.2, ``Instrumentation--Engineered Safety Features Actuation 
System Instrumentation'' 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 Nuclear Regulatory Commission.
    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.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same RTS 
[Reactor Trip System] and ESFAS [Emergency Safety Features Actuation 
System] instrumentation is being used; the time response 
allocations/modeling assumptions in the Chapter 15 analyses are 
still the same; 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 SAR [Safety Evaluation Report]. Therefore, 
there will be no significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change does not alter the performance of the pressure and 
differential pressure transmitters, Process Protection racks, 
Nuclear Instrumentation, and Logic Systems used in the plant 
protection systems. These sensors and systems will still have 
response time verified by test before being placed in operational 
service. Changing the method of periodically verifying instrument 
response for these sensors and systems (assuring equipment 
operability) from time response testing or calibration and channel 
checks will not create any new accident initiators or scenarios. 
Periodic surveillance of these sensors and systems will continue and 
may be used to (a) detect significant degradation in the sensor 
responses characteristic, and (b) other degradation that could cause 
the response time characteristic to exceed the total allowance. The 
total time response allowance for each function bounds all 
degradation that cannot be detected by periodic surveillance. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    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 differential pressure 
sensors, the Process Protection racks, Nuclear Instrumentation, and 
Logic Systems is modified to allow use of actual test data or 
engineering data. The method of verification still provides 
assurance that the total system response is within that defined in 
the safety analysis, since calibration tests will continue to be 
performed and may be used to detect any degradation which (a) might 
significantly affect sensor response time, or (b) might cause the 
response time to exceed the total allowance. The total system time 
response allowance for each function bounds all degradation that 
cannot be detected by periodic surveillance. Based on the above, it 
is concluded that the proposed license amendment request does not 
result in a significant reduction in margin with respect to plant 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

[[Page 48756]]

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

    Date of application for amendments: May 31, 2000.
    Description of amendment request: The proposed amendments would 
revise the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, 
Technical Specifications (TSs) Surveillance Requirement (SR) 3.6.1.3.11 
to allow a representative sample of reactor instrumentation line excess 
flow check valves (EFCVs) to be tested every 24 months, instead of 
testing each EFCV every 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The current SR frequency requires each reactor instrumentation 
line EFCV to be tested every 24 months. The EFCVs at PBAPS, Units 2 
and 3 are designed to not close accidentally during normal 
operation, but will close automatically in the event of a line break 
downstream of the valve. The proposed changes would allow a reduced 
number of EFCVs to be tested each operating cycle. Since the EFCVs 
are an accident mitigation feature, their postulated failure to 
isolate cannot initiate previously evaluated accidents. In addition, 
since the proposed changes will only change the surveillance 
frequency, there can be no increase in the probability of occurrence 
of an accident as a result of this proposed change.
    The postulated break of an instrument line attached to the 
reactor coolant pressure boundary is discussed and evaluated in the 
Updated Final Safety Analysis Report (UFSAR), Section 5.2.3.5. The 
proposed change will continue to verify the operability of the EFCVs 
to perform their mitigating functions. Industry operating experience 
as documented in the Boiling Water Reactors Owners Group (BWROG) 
Report B21-00658-01 provides supporting evidence that the reduced 
testing frequency will not affect the high reliability of these 
valves. The radiation dose consequences of such a break are not 
impacted by this proposed change. Therefore, the proposed TS changes 
do not involve a significant increase in the consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes would allow a reduced number of EFCVs to be 
tested each operating cycle. No other changes in requirements are 
being proposed. The changes are not a physical alteration of the 
plant and will not alter the operation of the structures, systems 
and components as described in the UFSAR. Therefore, a new or 
different kind of accident will not be created.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety. The consequences of an unisolable 
rupture of an instrument line has been previously evaluated in the 
PBAPS, Units 2 and 3 UFSAR, Section 5.2.3.5. That evaluation assumed 
a continuous discharge of reactor water for the duration of the 
detection and cooldown sequence. The integrity and functional 
performance of the secondary containment and standby gas treatment 
system are not impaired by this event, and the calculated potential 
offsite exposures are substantially below the guidelines of 10 CFR 
Part 100. Therefore, a failure of an EFCV, though not expected as a 
result of this TS change, is bounded by the previous evaluation of 
an instrument line break. Since the proposed changes are only 
affecting the surveillance frequency, the accident analyses are 
unaffected and this change does not involve a significant reduction 
in the margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 7, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point Nuclear Generating Unit No. 3 (IP3) Technical 
Specifications (TSs) would require either the Operations Manager or the 
Assistant Operations Manager to hold a Senior Reactor Operator (SRO) 
license. The proposed amendment would also remove the title of ``Shift 
Manager'' from the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. This change allows either the Operations Manager or 
Assistant Operations Manager to be SRO licensed. This is an 
administrative change. The Operations department will still have an 
SRO licensed individual overseeing the operating crews. Therefore, 
there will be no increase in the probability or consequences of an 
evaluated accident. This is consistent with the qualifications 
required to be a manager in TS 6.3.1.
    The change also deletes the title of Shift Manager. At IP3, 
``Shift Manager'' is the NYPA [New York Power Authority] specific 
title for the person meeting the requirements of 10 CFR 
50.54(m)(2)(ii) as the SRO assigned responsibility for overall plant 
operation. This requirement is redundant to 10 CFR 50.54(m)(2)(ii) 
and TS section 6.2.2 requirements for an SRO and therefore removal 
is an administrative change with no increase in the probability or 
consequences of an accident.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The change allows either the Operations Manager or Assistant 
Operations Manager to hold the SRO license. The proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated since they do not affect 
plant configuration or plant design. The Operations Manager and the 
Assistant Operations Manager are still required to maintain a 
knowledge of IP3 plant design and operations through job position 
requirements.
    The change also deletes the title of Shift Manager. At IP3, 
``Shift Manager'' is the NYPA specific title for the person meeting 
the requirements of 10 CFR 50.54(m)(2)(ii) as the SRO assigned 
responsibility for overall plant operation. This requirement is 
redundant to 10 CFR 50.54(m)(2)(ii) and TS section 6.2.2 
requirements and it is therefore an administrative change that 
cannot create the possibility of a new or different accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The change allows either the Operations Manager or Assistant 
Operations Manager to hold the SRO License. The proposed amendment 
does not involve a significant reduction in a margin of safety 
because the Operations Manager and/or the Assistant Operations 
Manager is still required to maintain a current SRO license. 
Administrative Controls ensure that shift activities are directed by 
an individual holding an SRO license. Technical Specification 6.3.1 
ensure that the Operations Manager will be a knowledgeable and 
qualified individual.
    The change also deletes the title of Shift Manager. At IP3, 
``Shift Manager'' is the NYPA specific title for the person meeting 
the requirements of 10 CFR 50.54(m)(2)(ii) as the SRO assigned 
responsibility for overall plant operation. This requirement is 
redundant to 10 CFR 50.54(m)(2)(ii) therefore the change has no 
effect on requirements and cannot offset the margin of safety.

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

[[Page 48757]]

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.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: July 21, 2000.
    Description of amendment request: The proposed amendment would 
delete the requirement to have the Control Room Emergency Air Treatment 
System (CREATS) Actuation Instrumentation and CREATS operable in Modes 
5 and 6 except during core alterations and fuel movement.
    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:

Evaluation of More Restrictive Changes

    The more restrictive changes (which is a conservative 
characterization, as these changes are implied by the current 
specifications) associated with amending the Applicability section 
for LCO [limiting condition for operation] 3.3.6 and LCO 3.7.9, and 
Condition C of LCO 3.3.6 and Condition D and F of LCO 3.7.9, to 
include ``during CORE ALTERATIONS'', do not involve a significant 
hazards consideration as discussed below:
    (1) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The changes add 
a conservative Mode of Applicability for the Control Room Emergency 
Air Treatment System (CREATS) and CREATS actuation instrumentation. 
This does not increase the probability of an accident previously 
evaluated since the CREATS and CREATS actuation instrumentation 
themselves are not accident initiators. The proposed changes are 
consistent with the guidance of NUREG-1431 and provide assurance 
that the CREATS is in the conservative mode of operation for a 
response to an accident. Therefore, the probability or consequences 
of an accident previously evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
change for a new mode of applicability does not of itself involve a 
physical alteration of the plant or change in the methods governing 
normal plant operation. The change only involves a conservative 
increase in the requirement of when the CREATS and CREATS actuation 
instrumentation are operable. Therefore, the possibility for a new 
or different kind of accident from any accident previously evaluated 
are not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed change requires the CREATS and CREATS actuation 
instrumentation to be in the conservative mode of operation for a 
response to an accident. The change adds conservatism as determined 
by the guidance of NUREG-1431. Therefore, this change does not 
involve a significant reduction in a margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
create the possibility of a new or different kind of accident from 
any accident previously evaluated, or involve a significant 
reduction in a margin of safety. Therefore, it is concluded that the 
proposed changes meet the requirements of 10 CFR 50.92(c) and do not 
involve a significant hazards consideration.

Evaluation of Less Restrictive Changes

    The less restrictive changes associated with amending the 
applicability sections for LCO 3.3.6 and LCO 3.7.9, and Condition C 
of LCO 3.3.6 and Condition D and F of LCO 3.7.9, to delete Modes 5 
and 6 from these sections do not involve a significant hazards 
consideration as discussed below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The changes are 
the result of an analysis performed of the control room dose 
consequences which could occur as the result of a potential waste 
gas decay tank failure. This does not increase the probability of an 
accident previously evaluated since the Control Room Emergency Air 
Treatment System (CREATS) and CREATS actuation instrumentation 
themselves are not accident initiators. The results of the analysis 
show that if no credit is taken for the CREATS, the control room 
doses remain well within the limits specified in 10 CFR 50, Appendix 
A, GDC [General Design Criteria] 19 and the guidance provided by the 
NRC in NUREG-0737 Section ll.B.2, Dose Rate Criteria, and NUREG-0800 
Section 6.4, Control Room Habitability Program. The proposed Mode of 
Applicability change is consistent with the guidance of NUREG-1431 
which allows plant-specific changes with respect to Modes 5 and 6. 
Therefore, the probability or consequences of an accident previously 
evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes associated with the modes of applicability for the CREATS 
and CREATS actuation instrumentation are not of themselves nor do 
they affect potential accident initiators. Therefore, the 
possibility for a new or different kind of accident from any 
accident previously evaluated are not created.
    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes remove the requirements for the control 
room ventilation system, which has been shown by analysis to not be 
required to meet regulatory limits. The changes are consistent with 
the guidance of NUREG-1431. Therefore, these changes do not involve 
a significant reduction in a margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
create the possibility of a new or different kind of accident from 
any accident previously evaluated, or involve a significant 
reduction in a margin of safety. Therefore, it is concluded that the 
proposed changes meet the requirements of 10 CFR 50.92(c) and do not 
involve a significant hazards consideration.
    The less restrictive change associated with amending the 
Required Action and Completion Time of Condition C of LCO 3.3.6 and 
Condition F of LCO 3.7.9 to remove a required action, do not involve 
a significant hazards consideration as discussed below:
    (1) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes to remove a required action of restoring equipment to 
operable status do not affect the probability of an accident as the 
Control Room Emergency Air Treatment System (CREATS) and CREATS 
actuation instrumentation, in and of themselves, have no failure 
modes or effects which are precursors to accidents. The proposed 
changes do not introduce any new failure modes or effects to any 
other system or component which is a precursor to an accident. The 
remaining Required Actions within the referenced Conditions place 
the plant outside of the Mode of Applicability for these systems. 
Therefore, the probability or consequences of an accident previously 
evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The changes do 
not of themselves involve a physical alteration of the plant or 
change in the methods governing normal plant operation. The proposed 
changes create no new functional interactions with existing plant 
equipment nor do they introduce any new failure modes or mechanisms 
which could lead to reactor core damage or fission product release. 
Therefore, because the changes do not affect any system that can act 
as an accident precursor, the possibility for a new or different 
kind of accident from any accident previously evaluated are not 
created.

[[Page 48758]]

    (3) Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes remove requirements for restoring 
systems which are no longer required. The changes are consistent 
with the guidance of NUREG-1431. Therefore, these changes do 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.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Section Chief: Marsha K. Gamberoni.

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: June 29, 2000.
    Description of amendment request: The proposed amendment would 
change the Farley Nuclear Plant, Units 1 and 2, design bases described 
in the Final Safety Analysis Report. The change adds a description of 
the methodology Southern Nuclear Operating Company uses to determine 
what systems and components need to be protected from tornado missiles.
    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.
    Proposed for NRC review and approval are changes to the Farley 
Nuclear Plant (FNP) Final Safety Analysis Report (FSAR) which in 
essence constitute a license amendment to incorporate use of an 
NRC[-]approved methodology to assess the need for additional 
positive (physical) tornado missile protection of specific features 
at FNP. The FSAR changes will reflect use of the Electric Power 
Research Institute (EPRI) Topical Report ``Tornado Missile Risk 
Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted 
in the NRC Safety Evaluation Report on this topic dated October 26, 
1983, the current licensing criteria governing tornado missile 
protection are contained in Standard Review Plan (SRP) Sections 
3.5.1.4 and 3.5.2. These criteria generally specify that safety-
related systems be provided positive tornado missile protection 
(barriers) from the maximum credible tornado threat. However, SRP 
Section 3.5.1.4 includes acceptance criteria permitting relaxation 
of the above deterministic guidance, if it can be demonstrated that 
the probability of damage to unprotected essential safety-related 
features is sufficiently small.
    As permitted in NRC Standard Review Plan (NUREG-0800) sections, 
the combined probability will be maintained below an allowable 
level, i.e., an acceptance criterion threshold, which reflects an 
extremely low probability of occurrence. The FNP approach assumes 
that if the probability calculation result for the total plant 
identifies that the probability of a combination of tornado missiles 
striking and damaging a portion of an important system or component 
is greater than or equal to 10-\6\ then installation of 
unique missile barriers would be needed to lower the total combined 
probability below the acceptance criterion of 10-\6\.
    With respect to the probability of occurrence or the 
consequences of an accident previously evaluated in the FSAR, the 
possibility of a tornado reaching the FNP site and causing damage to 
plant structures, systems and components is a design basis event 
considered in the [FSAR]. The changes being proposed do not affect 
the probability that the natural phenomenon (a tornado) will reach 
the plant, but from a licensing basis perspective they do affect the 
probability that missiles generated by the winds of the tornado 
might strike and damage certain plant systems or components. There 
are a limited number of safety-related components that could 
theoretically be struck and consequently damaged by tornado-
generated missiles. The probability of tornado-generated missile 
strikes on ``important'' systems and components (as discussed in 
Regulatory Guide 1.117) is what is to be analyzed using the 
probability methods discussed above. The combined probability of 
damage will be maintained below an extremely low acceptance 
criterion to ensure overall plant safety. The proposed change is not 
considered to constitute a significant increase in the probability 
of occurrence or the consequences of an accident, due to the 
extremely low probability of damage due to tornado-generated 
missiles and thus an extremely low probability of a radiological 
release. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of 
previously evaluated accidents.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility of a tornado reaching the FNP site is a design 
basis event considered in the [FSAR]. This change involves 
recognition of the acceptability of performing tornado missile 
probability calculations in accordance with established regulatory 
guidance. The change therefore deals with an established design 
basis event (the tornado). Therefore, the proposed change would not 
contribute to the possibility of a new or different kind of accident 
from those previously analyzed. The probability and consequences of 
such a design basis event are addressed in Question 1 above. Based 
on the above discussions, the proposed change will not create the 
possibility of a new or different kind of accident than those 
previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    The existing licensing basis for FNP with respect to the design 
basis event of a tornado reaching the plant, generating missiles and 
directing them toward safety-related systems and components is to 
provide positive missile barriers for all safety-related systems and 
components. With the change, it will be recognized that there is an 
extremely low probability, below an established acceptance limit, 
that a limited subset of the ``important'' systems and components 
could be struck and consequently damaged. The change from protecting 
all safety-related systems and components to ensuring an extremely 
low probability of occurrence of tornado-generated missile strikes 
and consequential damage on portions of important systems and 
components is not considered to constitute a significant decrease in 
the margin of safety due to that extremely low probability. 
Therefore, the changes associated with this license amendment 
request do not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: L. Raghavan, Acting.

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

    Date of amendment request: May 16, 2000.
    Description of amendments request: Amend Technical Specification 
(TS) 4.8.1.1.2 to revise the emergency diesel generator fuel oil 
surveillance requirements to adopt more current industry standards.
    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.

[[Page 48759]]

    The probability of occurrence or the consequences for an 
accident is not increased by this request. The proposal to establish 
a Diesel Fuel Oil Program and specifying the ASTM [American Society 
for Testing and Materials] standards in the TS Bases does not modify 
the manner in which the plant is operated. Deletion of the portion 
of the surveillance requirement (SR) that specifies the use of 
sodium hypochlorite solution in cleaning of the fuel oil storage 
tanks, and the deletion of the SR to perform a pressure test of 
those portions of the diesel fuel oil system designed as Section 
III, subsection ND of the ASME [American Society of Mechanical 
Engineers] Code do not alter the way any structure, system, or 
component functions and does not modify the manner in which the 
plant is operated.
    This request will ensure that the fuel oil continues to be 
properly evaluated to ensure that the fuel oil will not degrade the 
ability of the D/G [diesel generator] to perform its intended 
function. The fuel oil storage tanks will be cleaned at the required 
frequency. The deletion of the SR to perform a pressure test of 
those portions of the diesel fuel oil system designed to Section 
III, subsection ND of the ASME Code, removes potential confusion 
about testing of the fuel oil system since no portion of the system 
is designed to Section III, subsection ND of the ASME Code. 
Therefore, these changes will not change or impact previously 
evaluated accidents and the D/Gs ability to perform their intended 
function.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are procedural in nature concerning fuel 
oil testing, cleaning chemical to be used on the fuel oil storage 
tanks, and deletion of the pressure test of those portions of the 
diesel fuel oil system designed as Section III, subsection ND of the 
ASME Code. The possibility for an accident or malfunction of a 
different type than any evaluated previously in SQN's [Sequoyah's] 
Final Safety Analysis Report are not created. The proposal does not 
alter the way any structure, system, or component functions and does 
not modify the manner in which the plant is operated. The fuel oil 
quality will not be reduced and will not result in a decrease in D/G 
operability. The fuel oil storage tanks will be cleaned at the 
required frequency. Therefore, the possibility of a new or different 
kind of accident previously evaluated is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes are procedural in nature concerning fuel 
oil testing, cleaning chemical to be used on the fuel oil storage 
tanks, and deletion of the pressure test of the diesel fuel oil 
system. The margin of safety has not been reduced since the change 
in test methodologies are NRC approved and will continue to ensure 
the quality of the fuel oil. Also, deletion of the portion of the SR 
that specifies the use of sodium hypochlorite does not change the 
requirement to clean the fuel oil storage tanks. ASME Code 
requirements will continue to be met. Therefore, the proposed 
changes do not involve a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: July 10, 2000 (TS 00-08).
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) as follows:

Part A--Channel Operational Test (COT) 12 Hour Limit

    Channel operational tests (COTs) are performed for the Power 
Range and Intermediate Range neutron monitors in accordance with 
Reactor Trip System (RTS) Surveillance Requirements (SRs) 3.3.1.7 
and 3.3.1.8. While the unit is in Modes 1 or 2, SR 3.3.1.7 is 
performed for the Power Range monitors every 92 days. SR 3.3.1.8 is 
performed for the Intermediate Range monitors prior to startup of 
the reactor and at various points during power escalation or 
reduction. In addition, SR 3.1.10.1 currently requires that a COT be 
performed on the Power Range and Intermediate Range neutron monitors 
within 12 hours prior to initiation of a physics test, even though 
SR 3.3.1.7 and SR 3.3.1.8 have been performed on the required 
frequency.
    TVA proposes to eliminate the 12 hour requirement for the 
testing required by SR 3.1.10.1 so that the testing performed for SR 
3.3.1.7 and SR 3.3.1.8 can be used to satisfy SR 3.1.10.1. This 
issue was addressed by Technical Specification Task Force (TSTF) 
Traveler 108. The proposed amendment revises SR 3.1.10.1 to 
implement the portion of the approved TSTF 108 applicable to Watts 
Bar.

Part B--Trip System Logic for Physics Testing TSTF Traveler 315

    During the performance of physics testing one power range 
channel is used to provide input to the reactivity computer. In 
preparation for the test, the fuses to the electronics drawer for 
the channel are removed and the channel is placed in a tripped 
condition and results in the NIS trip logic being in a one-out-of-
three logic status. Therefore, any spurious signals received on one 
channel will result in a reactor trip. The changes proposed by TSTF-
315 allows the fuses to remain in the NIS channel that is connected 
to the reactivity computer and avoid tripping the bistables 
associated with the NIS channel. This configuration results in the 
channel being in a bypassed state and places the overall logic in a 
two-out-of-three logic status. The advantage of this configuration 
is that a single spurious signal would not result in a reactor trip. 
The proposed amendment does not deviate from the version of TSTF-315 
that was approved by NRC on June 29, 1999.

    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

Part A--Channel Operational Test (COT) 12 Hour Limit

    The proposed amendment removes the requirement to perform an 
additional Channel Operational Test (COT) on the Intermediate and 
Power Range functions within 12 hours of performing a physics test. 
The Intermediate and Power Range instrumentation is determined to be 
OPERABLE by periodic surveillance requirements which must be 
confirmed to be within frequency prior to making the reactor 
critical. A COT for the Intermediate or Power Range instrumentation 
is not a precursor to, or assumed to be an initiator of any analyzed 
accident. Therefore, this change does not involve a significant 
increase in the probability of an accident previously evaluated.
    Regarding a significant increase in the consequences of an 
accident, several factors must be considered. First the physics 
tests are performed in accordance with the Technical Specifications 
in Mode 2. Therefore, the power level of the reactor is limited to 5 
percent or less. Along with this, the reactor trip function of the 
intermediate range detectors will be unaffected by the proposed 
amendment and therefore, will be available to mitigate a reactivity 
transient at low power. Further, the trip setpoint for the power 
range monitors are decreased during startup of the reactor from the 
normal 109% setpoint to a value less than or equal to 85%. This 
setpoint reduction provides an additional measure to limit a 
reactivity excursion. Considering these factors, the proposed change 
will not involve a significant increase in the consequences of an 
accident previously evaluated.

Part B--Trip System Logic for Physics Testing

    During the performance of physics testing one power range 
channel is used to provide input to the reactivity computer. In 
preparation for the test, the fuses to the electronics drawer for 
the channel are removed and the channel is placed in a tripped 
condition and results in the NIS trip logic being in a one-out-of-
three logic status. Therefore, any spurious signals received on one 
channel will result in a reactor trip. The changes proposed by TSTF-
315 allows the

[[Page 48760]]

fuses to remain in the NIS channel that is connected to the 
reactivity computer. This configuration results in the channel being 
in a bypassed state and places the overall logic in a two-out-of-
three logic status. The advantage of this configuration is that a 
single spurious signal will not result in a reactor trip. In 
addition, the physics tests required by LCO 3.1.10 are performed 
while the reactor is in Mode 2. Therefore, the thermal power of the 
reactor is restricted to 5 percent or less. Neutron flux, which is 
monitored by the NIS, is only one of several RTS variables which may 
initiate a reactor trip in Mode 2. The other variables include 
reactor coolant temperature, pressurizer pressure and steam 
generator water level. These variables are unaffected by the 
proposed amendment. Considering this, the low thermal power level of 
the reactor, and a potential reduction in unnecessary plant 
transients due to the one-out-of-three logic, the proposed amendment 
will not significantly impact the safe operation of the plant. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    B. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Part A--Channel Operational Test (COT) 12 Hour Limit

    The proposed amendment is not based on a change in the design or 
configuration of the plant. Also, the proposed amendment does not 
change the manner in which the plant is operated. The amendment 
deletes the requirement for the performance of a COT for the 
Intermediate and Power Range instrumentation within 12 hours of 
starting a physics test. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident than 
any previously evaluated.

Part B--Trip System Logic for Physics Testing

    The NIS provides indication, alarm, control, and trip signals 
along with the capability to monitor neutron flux over the complete 
range from reactor shutdown to 120 percent full power. The system 
also generates permissive and level trip signals, which are then 
coupled to the logic matrices of the RTS. This interface either 
allows power changes based upon proper functioning of the next range 
of measurement instrumentation or shuts down the reactor as unsafe 
operating limits are approached. The changes in the operation of the 
NIS proposed by this amendment for TSTF-315, do not inhibit the 
capabilities of the system to initiate a reactor trip, if required. 
Therefore, the proposed amendment will not create the possibility of 
a new or different kind of accident.
    C. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.

Part A--Channel Operational Test (COT) 12 Hour Limit

    As stated previously, the proposed change deletes the 
requirement to perform an additional COT for the Intermediate and 
Power Range functions within 12 hours of the start of physics test. 
The Intermediate and Power Range instrumentation channels are 
determined to be operable by meeting the requirements of the 
periodic surveillances. These surveillance requirements are not 
affected by the proposed amendment. Since the equipment will be 
determined to be operable by periodic surveillances, the performance 
of the a surveillance prior to the initiation of a physics test does 
not provide any additional assurance that the functions are more 
reliable. Considering this, the proposed amendment does not 
significantly reduce the margin of safety.

Part B--Trip System Logic for Physics Testing

    During the low power physics testing, implementation of the 
proposed amendment will result in one power range channel being in a 
bypassed state. In this configuration, there will be three available 
channels with a two-out-of-three logic required to actuate the 
neutron flux trip function. As required by LCO 3.1.10, the testing 
will be performed while the reactor is in Mode 2 and therefore, 
restricted by the Technical Specifications to a power level of less 
than or equal to 5 percent.
    There are two power range control functions, rod control and 
steam generator level control. At the 5 percent or less power level, 
rod control is in manual and is not affected by the testing 
configuration. Steam generator level control is not affected since 
its input from the NIS channel connected to the Reactivity Computer 
is placed in bypass when establishing the test configuration. 
Therefore, an assumed failure affecting these control functions does 
not have to be considered for the testing configuration. Also while 
in this configuration, an assumed single failure will not prevent 
the power range monitors from actuating as designed.
    The reactor trip function of the intermediate range detectors 
will be unaffected by the proposed amendment and therefore, will be 
available to mitigate a reactivity transient at low power. Further, 
the trip setpoint for the power range monitors are decreased during 
startup of the reactor from the normal 109% setpoint to a value less 
than or equal to 85%. This setpoint reduction provides an additional 
measure to limit a reactivity excursion.
    Based on the preceding, TVA concludes that there is no significant 
reduction in the margin of safety due to the implementation of the 
proposed amendment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: June 22, 2000.
    Description of amendment request: The proposed amendments to the 
Technical Specification Figures 3.4-2, 3.4-3, and associated Bases 
would extend the cumulative core burnup applicability limits for the 
reactor coolant system pressure-temperature (P/T) operating limits, Low 
Temperature Overpressure Protection System (LTOPS) setpoints, and LTOPS 
enable temperature (T enable). Implementation of American Society of 
Mechanical Engineers (ASME) Section XI Code Cases N-640 and N-514 will 
require exemptions from the requirements of 10 CFR 50, Appendix G.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated[?]
    The proposed changes extend the cumulative core burnup 
applicability of the existing North Anna Units 1 and 2 P/T limits, 
LTOPS setpoints, and T enable values. No changes to plant systems, 
structures, or components are proposed, and no new allowable 
operating modes are established, The P/T limits, LTOPS setpoints, 
and T enable values do not contribute to the probability of 
occurrence or consequences of accidents previously analyzed. The 
revised licensing basis analyses utilize acceptable analytical 
methods, and continue to demonstrate that established accident 
analysis acceptance criteria are met. Therefore, there is no 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated[?]
    The proposed changes extend the cumulative core burnup 
applicability of the existing North Anna Units 1 and 2 P/T limits, 
LTOPS setpoints, and T enable values. No changes to plant systems, 
structures, or components are proposed, and no new allowable 
operating modes are established. Therefore, the proposed changes do 
not create the possibility of any accident or malfunction of a 
different type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety[?]
    The proposed revised analysis bases use the ASME Section XI code 
Case N-640 K1c stress intensity formulation and a plant

[[Page 48761]]

specific application of the analysis methodology which supports ASME 
Section XI Code Case N-514. These analysis features are less 
restrictive than those associated with the existing analyses, but 
are conservative with respect to [those] established by ASME Section 
XI margins. The proposed revised analyses support continued use of 
the existing North Anna Units 1 and 2 Technical specification P/T 
limit curves, LTOPS setpoints, LTOPS enable temperatures for North 
Anna Units 1 and 2 cumulative core burnups up to 32.3 effective full 
power years (EFPY) and 34.3 EFPY, respectively. The analyses 
demonstrate that established analysis acceptance criteria continue 
to be met. Specifically, the existing P/T limit curves, LTOPS 
setpoints, and LTOPS T enable values provide acceptable margin to 
vessel fracture under both normal operation and LTOPS design basis 
(mass addition and heat addition) accident conditions. Therefore, 
the proposed changes do not result in a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: L. Raghavan, Acting.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: June 22, 2000.
    Description of amendment request: The proposed changes would modify 
Facility Operating Licenses NPF4 and NPF-7, along with the associated 
Bases, to permit the elimination of the assumed increase in the rod 
control cluster assembly (RCCA) drop time resulting from a concurrent 
trip and seismic event, when determining if the measured rod drop times 
meet the Technical Specifications limit of 2.7 seconds.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated[?]
    Elimination of the assumed increase in the RCCA drop time 
resulting from a concurrent trip and seismic event when determining 
if the measured rod drop times, including measurement uncertainties, 
meet the accident analysis limit[,] does not contribute to the 
probability of previously analyzed accidents. The proposed change 
will not alter the limiting results of the safety analyses presented 
in Chapter 15 of the UFSAR [Updated Final Safety Analysis Report]. 
Although the proposed change eliminates an accident consideration 
that is currently addressed in the UFSAR accident analyses (i.e. any 
Chapter 15 accident with the effects of a concurrent seismic 
occurrence reflected in the RCCA drop time), there is no significant 
increae in the probability or consequences of any accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated[?]
    There are no modifications to the plant as a result of the 
changes. No new accident or event initiators are created by 
eliminating the assumed increase in the RCCA drop time resulting 
from a concurrent trip and seismic event. The proposed change will 
not alter the ability of the reactor protection and control system 
to perform their design functions or to meet the applicable criteria 
set forth in the IEEE [Institute of Electrical and Electronics 
Engineers] and ANSI [American National Standards Institute] 
standards and in 10 CFR 50 Appendix A. Therefore, the proposed 
changes do not create the possibility of any accident or malfunction 
of a different type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety[?]
    The proposed change will not alter the limiting results of the 
safety analyses presented in Chapter 15 of the UFSAR. Elimination of 
the assumed increase in the RCCA drop time resulting from a 
concurrent trip and seismic event when determining if the measured 
rod drop times, including measurement uncertainties, [meet] the 
accident analysis limit maintains adequate safety margin in the 
safety analysis. Therefore, the proposed change does not 
significantly reduce 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.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: L. Raghavan, Acting.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: March 29, 2000.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 3.19 and TS 4.1. The change would reflect 
two redundant trains of bottled air for the main control room (MCR), 
include remedial action statements for one train and two trains 
inoperable, eliminate the extension of 8 hours to 24 hours currently 
permitted by TS 3.19.B, add requirements for an inoperable control room 
pressure boundary, and include additional surveillance testing 
requirements. The TS 3.19 Basis and TS 4.1 Basis would be revised for 
consistency with the respective TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed TS change includes train specific requirements, 
adds requirements for an inoperable control room pressure boundary, 
imposes additional surveillance testing requirements for the MCR 
bottled air system, and is consistent with the existing accident 
analyses. We have reviewed the proposed TS change relative to the 
requirements of 10 CFR 50.92 and determined that a significant 
hazards consideration is not involved. Specifically, operation of 
Surry Power Station with the proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not involve a physical modification and 
does not modify the design or operation of the MCR bottled air 
system or the plant. Since the MCR bottled air system functions to 
respond to--not prevent--an accident, the probability of occurrence 
of an accident is not affected. The elimination of the currently 
allowed extension of the remedial action time, the addition of train 
specific requirements and inoperable boundary requirements, and the 
imposition of additional surveillance testing requirements serve to 
ensure no increase in the consequences of an accident. Therefore, 
the proposed change does not significantly increase the probability 
of occurrence or the consequences of any previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not involve a physical modification and 
does not affect the design or operation of the MCR bottled air 
system or the plant. Consequently, no new or unique operational 
modes or accident precursors are introduced. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed change does not involve a physical modification and 
does not modify the design or operation of the MCR bottled air 
system or the plant. The elimination of

[[Page 48762]]

the currently allowed extension of the remedial action time, the 
addition of train specific requirements and inoperable boundary 
requirements, and the imposition of additional surveillance testing 
requirements serve to ensure the bottled air system's ability to 
pressurize the main control room for one hour following a design 
basis accident, which is consistent with the existing accident 
analyses. Therefore, the proposed change does not result in a 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Acting Section Chief: L. Raghavan.

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.

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: July 13, 2000.
    Description of amendments request: Amend Technical Specification 
3.7.5.c to allow an increase in the average essential raw cooling water 
supply header temperature from 84.5 deg.F to 87 deg.F until September 
30, 2000.
    Date of publication of individual notice in the Federal Register: 
July 20, 2000 (65 FR 45113).
    Expiration date of individual notice: August 3, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: August 23, 1999, as supplemented 
January 8, 2000.
    Brief description of amendment: The amendment deletes certain 
license conditions that are obsolete and no longer apply.
    Date of issuance: July 24, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 130.
    Facility Operating License No. NPF-62: The amendment revised the 
License.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51346). The January 8, 2000, submittal identified an additional 
license condition that was no longer applicable and thus did not change 
the scope of the action noticed or alter the initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 24, 2000.
    No significant hazards consideration comments received: No.
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina
    Date of application for amendment: August 2, 1999, as supplemented 
April 7 and July 5, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification 6.2.2.e, ``Administrative Controls--Unit Staff.'' The 
license requirements for operations management have been modified.
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000.
    Amendment No.: 99.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46426). The supplemental letters dated April 7 and July 5, 2000, 
contained clarifying information only, and did not change the initial 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: April 12, 2000, as supplemented 
June 2, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.4.9.2, ``Pressure/Temperature (P-T) Limits--
Reactor Coolant System,'' and TS 3/4.4.9.4, ``Overpressure Protection 
System,'' and the associated Bases. Specifically, the amendment 
incorporates results of the Reactor Vessel Surveillance Program capsule 
analysis and an exemption from 10 CFR 50.60(a), based on American 
Society of Mechanical Engineers Code Case N-640.

[[Page 48763]]

    Date of issuance: July 28, 2000.
    Effective date: July 28, 2000.
    Amendment No. 100.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25762). The supplemental letter dated June 2, 2000, contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 29, 1999, as supplemented by 
letters dated August 24, 1999, January 27, 2000, May 22, 2000, and May 
31, 2000.
    On June 14, 2000, the Commission published in the Federal Register 
(FR) Notice of consideration of issuance of amendment to facility 
operating license, proposed no significant hazards consideration 
determination, and opportunity for a hearing (65 FR 37425). In this 
finding, incorrect reference is made to supplements dated August 8, 
1999, and March 29, 2000. No supplements from the licensee with these 
dates are related to this amendment.
    Brief description of amendment: The amendment modifies Technical 
Specification 3.8.1.1 and associated Bases by extending the Emergency 
Diesel Generator (EDG) allowed outage time from 72 hours to ten days. 
In the supplemental letter dated May 22, 2000, an alternate source for 
the onsite power system during the EDG maintenance outage, by way of a 
temporary EDG (TEDG), was added. The application dated July 29, 1999, 
did not include the TEDG.
    Date of issuance: July 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37425). This notice is based on the supplement dated May 22, 2000, and 
supercedes the notice dated February 9, 2000 (65 FR 6406), which is 
based on the licensee's letter dated July 29, 1999. The May 31, 2000, 
supplement did not expand the scope of the application as noticed or 
change the proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 21, 2000.
    No significant hazards consideration comments received: No

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: March 6, 2000.
    Brief description of amendment: Revised the Improved Technical 
Specification Action Condition and Surveillance Requirement related to 
the diesel-driven emergency feedwater pump (EFW-3) required lube oil 
volume.
    Date of issuance: July 17, 2000.
    Effective date: July 17, 2000.
    Amendment No.: 192.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21036).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Dade County, Florida

    Date of application for amendments: November 30, 1999, as 
supplemented March 8, May 15, and July 5, 2000.
    Brief description of amendments: The proposed amendments would 
revise the Technical Specifications to allow the use of credit for 
soluble boron in the spent fuel pool criticality analyses. In addition, 
a revised criticality analysis for the fresh fuel storage racks will be 
used to update the licensing bases.
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000.
    Amendment Nos.: 206 and 200.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25765). The May 15, and July 5, 2000, submittals provided clarifying 
information that did not change the scope of the original request or 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Dade County, Florida

    Date of application for amendments: April 27, 2000.
    Brief description of amendments: Incorporate references to the NRC 
safety evaluations supporting exemptions granted for the Thermo-Lag 
Upgrade project. In addition, the amendments modify Technical 
Specification Section 6.0, Administrative Controls, Section 4.7.6.g, to 
include page 3/4 7-21 which was inadvertently excluded from the 
previous submittal and amendment.
    Date of issuance: July 20, 2000.
    Effective date: July 20, 2000.
    Amendment Nos.: 207 and 201.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications and the Operating Licenses.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34746).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 20, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 15, 1999, as supplemented 
December 22, 1999, and February 24, 2000.
    Brief description of amendment: The amendment editorially revised 
the Technical Specifications to enhance clarity.
    Date of Issuance: July 17, 2000.
    Effective date: July 17, 2000 and shall be implemented within 30 
days of issuance.
    Amendment No.: 211.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12293).
    The February 24, 2000, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 3, 1999, as supplemented on 
December 22, 1999.

[[Page 48764]]

    Brief description of amendment: The proposed amendment revised the 
Technical Specifications to permit continued plant operation with a 
maximum of two inoperable recirculation loops, provided certain 
conditions are met. Oyster Creek's Technical Specifications (TSs), 
Section 3.3.F.2 currently permit operation with 4 of the 5 
recirculation loops with certain constraints. If only 3 loops are 
operable, however, the TSs require plant shutdown within 12 hours. 
Analysis indicates that the plant may be safely operated at 90 percent 
power with three operable recirculation loops.
    Two definitions are added to Section 1 of the TSs to specify the 
difference between an idle recirculation loop and an isolated 
recirculation loop. These definitions have been incorporated into the 
specification to provide an explicit description of acceptable valve 
configurations. In addition, several paragraphs have been added to the 
Bases of Section 3.3 and one paragraph in the Bases of Section 3.10 has 
been modified. In each case the Bases section has been segmented from 
the specification, which affects the pagination of the Bases.
    Date of Issuance: July 27, 2000.
    Effective date: July 27, 2000 and shall be implemented within 30 
days of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25766). The December 22, 1999, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 11, 1997, as supplemented by 
letter dated May 8, 2000.
    Brief description of amendment: The amendment revised the technical 
specifications by adding a new limiting condition for operation (LCO) 
for an inoperable engineered safety features logic subsystem. In 
addition, administrative changes were made to either support the new 
LCO or clarify existing text.
    Date of issuance: July 25, 2000.
    Effective date: July 25, 2000, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 194.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6987). The May 8, 2000, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 25, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear 
Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: April 10, 2000.
    Brief description of amendment: This amendment modifies the 
requirements contained in the Salem Unit No. 2 Technical Specifications 
regarding the operation of the movable incore detector system and 
allows continued operation of Salem Unit No. 2 through the remainder of 
Cycle 11. The revision represents a one-time change to allow use of the 
movable incore detector system for measurement of core peaking factors 
with less than 75% and greater than or equal to 50% of the detector 
thimbles available. Public Service Electric and Gas Company submitted 
this request in response to degradation of the movable incore detector 
system.
    Date of issuance: July 25, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2000 (65 FR 
33378).
    The Commission received comments which were addressed in the NRC 
staff's Safety Evaluation dated July 25, 2000.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 25, 2000.
    No significant hazards consideration comments received: Yes.

Southern California Edison Company, et al., Docket Nos. 50-206, San 
Onofre Nuclear Generating Station (SONGS), Unit 1, San Diego County, 
California

    Date of application for amendment: December 2, 1999, as 
supplemented on May 16, 2000.
    Brief description of amendment: The amendment revised the SONGS 
Unit 1 Technical Specifications by revising the administrative controls 
to be consistent with the SONGS Unit 2 and 3 Technical Specification 
administrative controls including changes to the administrative control 
of working hours and working hour deviation approvals, position titles 
and responsibilities and organizational description reference, 
qualifications for a multi-discipline supervisor, quality assurance 
program control of review and audit and record retention procedures, 
high radiation area controls, description of the plant configuration 
for environmental protection, and environmental protection related 
document reporting. The amendment also incorporated changes related to 
certified fuel handlers and 10 CFR 50.54(x).
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000, to be implemented within 30 days of 
issuance.
    Amendment No.: 159.
    Facility Operating License No. DPR-13: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73096).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

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: December 17, 1999, as 
supplemented on June 30, 2000.
    Brief description of amendments: Revises License Condition to allow 
storage at the Sequoyah Nuclear Plant site of low-level radioactive 
waste generated at Watts Bar Nuclear Plant, Unit 1.
    Date of issuance: July 18, 2000.
    Effective date: July 18, 2000.
    Amendment Nos.: 257 and 248.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the Operating Licenses.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9012). The supplemental letter of June 30,

[[Page 48765]]

2000, did not change the initial No Significant Hazards Consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in an Environmental Assessment dated June 29, 2000 (65 FR 41739) and in 
a Safety Evaluation dated July 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 20, 1998, as 
supplemented July 19, 1999, and January 21, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to change the surveillance requirements for an 
inspection of the ice condenser flow channels that previously used a 
0.38 inch ice/frost buildup criterion to a criterion that limits flow 
blockage to the 15 percent value that was used in the accident 
analysis. Changes to the Bases were also made. Tennessee Valley 
Authority also indicated that its proposal is consistent with TS 
Traveler Form No. 336.
    Date of issuance: July 17, 2000.
    Effective date: July 17, 2000.
    Amendment No.: 25.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70093). The January 21, 2000, letter contained clarifying 
information that did not change the initial No Significant Hazards 
Consideration Determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 23, 2000.
    Brief description of amendment: The amendment relocates the 
specifications for reactor coolant conductivity and chloride 
concentration from the Technical Specifications to the Technical 
Requirements Manual.
    Date of Issuance: July 18, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment No.: 190.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37430).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 23, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications to increase the interval between Local Power Range 
Monitor calibrations from 1,000 equivalent full power hours to 2,000 
megawatt-days/ton.
    Date of Issuance: July 18, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment No.: 191.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37431).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 18, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 22, 2000.
    Brief description of amendment: The amendment removes the Technical 
Specifications surveillance requirement for visual inspection of 
suppression chamber coating integrity once each refueling outage.
    Date of Issuance: July 19, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 192.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37430).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 31, 2000, as supplemented by 
letter of July 7, 2000.
    Brief description of amendment: The amendment modifies the actions 
for Limiting Condition for Operation (LCO) 3.7.9, ``Ultimate Heat Sink 
(UHS),'' of the TSs. The new Action A for the LCO allows the plant to 
operate with the plant inlet water temperature of the UHS above 
90 deg.F, if the required lake water level is verified within 1 hour 
and once per 12 hours thereafter, but would require that the plant be 
shut down if the water temperature exceeded 94 deg.F. The amendment 
replaces the requirement to shut down the plant if the UHS water 
temperature exceeds 90 deg.F.
    Date of issuance: July 14, 2000.
    Effective date: July 14, 2000, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 134.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21040). The supplemental letter of July 7, 2000, had minor 
clarifications that are within the scope of the initial notice and does 
not alter the no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 14, 2000.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date

[[Page 48766]]

the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By September 8, 2000, 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 electronically from the ADAMS Public Library 
component on the NRC Web site, http://www.nrc.gov (the Electronic 
Reading Room). 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. Since the Commission has

[[Page 48767]]

made a final determination that the amendment involves no significant 
hazards consideration, if a hearing is requested, it will not stay the 
effectiveness of the amendment. Any hearing held would take place while 
the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: July 13, 2000, as supplemented 
by letters dated July 14 and 21, 2000.
    Brief description of amendment: The amendment permitted a one-time 
change to Technical Specification 4.4.5.0 and allowed alternate 
inspection scope and expansion criteria for steam generator tube 
inspections to be implemented during the mid-cycle outage scheduled for 
summer 2000.
    Date of issuance: July 26, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 217.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes.
    The NRC published a public notice of the proposed amendment, issued 
a proposed finding of no significant hazards consideration, and 
requested that any comments on the proposed no significant hazards 
consideration be provided to the staff by the close of business on July 
24, 2000. The notice was published in The Courier (in Russellville) and 
the Arkansas Democrat-Gazette (in Little Rock) from July 20 through 22, 
2000. No public comments were received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of Arkansas, and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated July 26, 2000.

    Dated at Rockville, Maryland, this 3rd day of August 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-20014 Filed 8-8-00; 8:45 am]
BILLING CODE 7590-01-P