[Federal Register Volume 66, Number 35 (Wednesday, February 21, 2001)]
[Notices]
[Pages 11050-11070]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-4228]


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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 January 29, 2001, through February 9, 2001. 
The last biweekly notice was published on February 7, 2001.

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 Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, 11555 Rockville Pike, Room O-1F15, Rockville, Maryland. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By March 23, 2001, 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, 11555 Rockville 
Pike, Room O-1F15, Rockville, Maryland, 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

[[Page 11051]]

the proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rules and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, 11555 Rockville Pike, Room O-1F15, Rockville, Maryland, 
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).

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: January 15, 2001.
    Description of amendment request: The proposed amendment revises 
the Technical Specification (TS) Design Features Section 5.4.2(f), 
``Spent Fuel Storage,'' to remove the existing TS fuel assembly U\235\ 
loading criterion for fuel assemblies stored in the spent fuel storage 
pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change has no effect on the normal 
operating, design basis accident, or transient analyses applicable 
to the TMI [Three Mile Island] Unit 1 fuel storage requirements. 
Other existing TMI Unit 1 Technical Specification provisions ensure 
sub-criticality for normal and postulated accident conditions.
    Therefore, the proposed change 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. Fuel assembly U\235\ loading is not an initial condition 
of a design basis accident or transient that either assumes the 
failure of or presents a challenge to the integrity of a fission 
product barrier. Discussion of fuel assembly U\235\ loading in the 
TMI Unit 1 UFSAR [Updated Final Safety Analysis Report] ensures that 
changes to fuel designs that increase fuel reactivity relative to 
design assumptions for fuel storage are evaluated in accordance with 
the requirements of 10 CFR 50.59.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. The proposed change does not affect existing 
TMI Unit 1 Technical Specification requirements controlling maximum 
fuel enrichment, allowable enrichment vs. burnup, soluble boron 
requirements, storage rack spacing, allowable rack locations for 
fuel assembly storage or sub-criticality requirements for normal and 
accident conditions. These existing Technical Specification 
requirements ensure that the current margin of safety is not 
reduced. The fuel assembly U\235\ loading criterion does not 
represent an input parameter or limiting design condition for any 
supporting design basis analyses applicable to the TMI Unit 1 spent 
fuel storage requirements.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.


[[Page 11052]]



Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina

    Date of amendment request: January 17, 2001.
    Description of amendment request: The proposed amendments would 
relax Surveillance Requirement 3.6.1.3.7 by allowing a ``representative 
sample'' of excess flow check valves to be tested every 24 months, such 
that each excess flow check valve will be tested at least once every 10 
years.
    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 license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The current surveillance requirement frequency requires each 
reactor instrumentation line excess flow check valve to be tested 
every 24 months. The excess flow check valves at BSEP are designed 
to close automatically in the event of a line break downstream of 
the valve. The proposed change allows a reduction in the number of 
excess flow check valves to be tested every 24 months to 
approximately 20 percent of the valves each operating cycle. 
Industry operating experience demonstrates a high level of 
reliability for these excess flow check valves. A failure of an 
excess flow check valve to isolate cannot initiate previously 
evaluated accidents. Therefore, there is no increase in the 
probability of occurrence of an accident as a result of this 
proposed change. The postulated failure of an excess flow check 
valve to isolate is bounded by the limiting analysis in the Updated 
Final Safety Analysis Report (UFSAR). For a postulated break of an 
instrument line upstream of an excess flow check valve, leakage from 
the line rupture would be minimized by the line size or the flow-
restricting orifice in the line. The rate and quantity of process 
fluid loss from an instrument line rupture is well within the 
capability of the reactor coolant make-up systems. The proposed 
change does not alter the design of the plants' instrument lines in 
any manner, and the integrity and functional performance of the 
secondary containment and Standby Gas Treatment system are not 
affected by this proposed change. The potential offsite radiological 
exposure associated with a postulated instrument line rupture 
upstream of an excess flow check valve is bounded by the main steam 
line break analysis and is substantially below the guidelines of 10 
CFR 100. Therefore, the proposed license amendments do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change allows a reduced number of excess flow check 
valves to be tested each operating cycle. No other change in 
requirements are being proposed. Industry operating experience 
demonstrates the high reliability of the excess flow check valves. 
The potential failure of an excess flow check valve to isolate is 
bounded by the main steam line break analysis. The proposed license 
amendments do not physically alter the plants and will not alter the 
operation of the structures, systems, and components described in 
the UFSAR. Therefore, a new or different kind of accident will not 
be created.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    Industry experience with excess flow check valves indicates that 
they have very low failure rates. The postulated failure of an 
excess flow check valve to isolate as a result of reduced testing is 
bounded by the limiting analysis in the UFSAR, which is the main 
steam line break analysis. For a postulated break of an instrument 
line upstream of an excess flow check valve, leakage from the line 
rupture would be minimized by the line size or the flow-restricting 
orifice in the line. The rate and quantity of process fluid loss 
from an instrument line rupture is well within the capability of the 
reactor coolant make-up systems. The proposed change does not alter 
the design of the plants' instrument line design in any manner, and 
the integrity and functional performance of the secondary 
containment and standby gas treatment system are not affected by 
this proposed change. Therefore, the proposed license amendments 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.
    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. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

    Date of amendment request: October 24, 2000.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to change the Westinghouse 
references for Best Estimate Large Break Loss of Coolant Accident 
(LOCA) analysis methodology. Reanalysis of large break LOCA transients, 
utilizing the NRC approved Westinghouse Best Estimate LOCA model 
WCOBRA/TRAC, was performed to demonstrate that 10 CFR 50.46 acceptance 
criteria are satisfied at uprated power conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No physical plant changes are being made as a result of using 
the Westinghouse Best Estimate Large Break LOCA analysis 
methodology. The proposed TS changes simply involve updating the 
references in TS 5.6.5.b, ``Core Operating Limits Report (COLR),'' 
to reference the Westinghouse Best Estimate Large Break LOCA 
analysis methodology (i.e., Westinghouse topical report, WCAP-12945-
P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, 
``Code Qualification Document for Best Estimate LOCA Analysis,'' 
March 1998). The plant conditions assumed in the analysis are 
bounded by the design conditions for all equipment in the plant; 
therefore, there will be no increase in the probability of a LOCA. 
The consequences of a LOCA are not being increased, since the 
analysis has shown that the Emergency Core Cooling System (ECCS) is 
designed such that its calculated cooling performance conforms to 
the criteria contained in 10 CFR 50.46, ``Acceptance criteria for 
emergency core cooling systems for light-water nuclear power 
reactors.'' Furthermore, the re-performance of the Large Break LOCA 
analysis has no effect on the performance of the ECCS equipment. No 
other accident consequence is potentially affected by this change.
    All systems will continue to be operated in accordance with 
current design requirements under the new analysis, therefore no new 
components or system interactions have been identified that could 
lead to an increase in the probability of any accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). No 
changes were required to the Reactor Protection System (RPS) or 
Engineered Safety Features (ESF) setpoints because of the new 
analysis methodology.
    Based on the analysis, it is concluded that the proposed TS 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind accident from any accident previously evaluated?
    There are no physical changes being made to the plant as a 
result of using the Westinghouse Best Estimate Large Break LOCA 
analysis methodology. No new modes of plant operation are being 
introduced. The

[[Page 11053]]

configuration, operation and accident response of the Byron Station 
and the Braidwood Station systems, structures or components are 
unchanged by utilization of the new analysis methodology. Analyses 
of transient events have confirmed that no transient event results 
in a new sequence of events that could lead to a new accident 
scenario. The parameters assumed in the analysis are within the 
design limits of existing plant equipment.
    In addition, employing the Westinghouse Best Estimate Large 
Break LOCA analysis methodology does not create any new failure 
modes that could lead to a different kind of accident. The design of 
all systems remains unchanged and no new equipment or systems have 
been installed which could potentially introduce new failure modes 
or accident sequences. No changes have been made to any RPS or ESF 
actuation setpoints.
    Based on this review, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed changes. Therefore, the 
proposed TS changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of Safety?
    It has been shown that the analytic technique used in the 
Westinghouse Best Estimate Large Break LOCA analysis methodology 
realistically describes the expected behavior of the Byron Station 
and Braidwood Station reactor system during a postulated LOCA. 
Uncertainties have been accounted for as required by 10 CFR 50.46. A 
sufficient number of LOCAs with different break sizes, different 
locations, and other variations in properties have been considered 
to provide assurance that the most severe postulated LOCAs have been 
evaluated. The analysis has demonstrated that there is a high 
probability that all acceptance criteria contained in 10 CFR 50.46, 
paragraph b, continues to be satisfied. Based on this review, the 
proposed TS changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

    Date of amendment request: November 7, 2000.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to extend the TS Surveillance Test 
Interval (STI) from a 92-day STI to an 18-month STI, for the Solid 
State Protection System (SSPS) slave relay types that meet the 
acceptance criteria for the reliability assessments performed in 
accordance with the methodology described in the NRC approved 
Westinghouse Electric Corporation Topical Reports.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes are consistent with the NRC approved 
Westinghouse Electric Corporation Topical Reports, WCAP-13877, 
``Reliability Assessment of Westinghouse Type AR Relays Used as SSPS 
Slave Relays,'' WCAP-13878, ``Reliability Assessment of Potter & 
Brumfield MDR Series Relays,'' and WCAP-13900, ``Extension of Slave 
Relay Surveillance Testing Intervals,'' that analyze extending the 
Solid State Protection System (SSPS) slave relay surveillance test 
interval (STI) for the Westinghouse Type AR slave relays and for the 
Potter & Brumfield MDR Series slave relays. The reliability 
assessment of the slave relays was comprised of a failure modes and 
effects analysis (FMEA) and an aging assessment of the slave relays. 
WCAP-13877 and WCAP-13878 verified that the Westinghouse Type AR and 
the Potter & Brumfield MDR Series slave relays are highly reliable 
and that degradation of the slave relays is sufficiently slow (i.e., 
the time to failure due to degradation is sufficiently long) that an 
18-month STI will adequately identify slave relay failures. A 92-day 
STI is no more likely to detect significant changes in the SSPS 
slave relays than an 18-month STI. The results demonstrate that 
extending the SSPS slave relay STI from 92 days to 18 months does 
not adversely affect the reliability of the SSPS slave relays 
utilized in Engineered Safety Features Actuation System (ESFAS) 
functions.
    The high reliability of these slave relays precludes the need 
for more frequent periodic surveillance testing to verify 
operability.
    As stated in WCAP-13877 and WCAP-13878, the overly conservative 
92-day STI can be extended to an 18-month STI without impact or 
consequence to slave relay reliability. In addition, the proposed 
changes will not adversely affect the ability of the SSPS to perform 
its safety function. The same ESFAS instrumentation is being used 
and the ESFAS reliability is being maintained with the proposed 
changes. Because the reliability of the slave relays used in the 
ESFAS applications is so high, elimination of the routine 
surveillance testing of the slave relays when the reactor is at 
power will have a positive impact on ESFAS availability and plant 
safety. The proposed changes will not modify any system interface 
and will not increase the likelihood of any accident initiator 
because such events are independent of the proposed changes. 
Therefore, the probability of an accident previously evaluated is 
not increased.
    The proposed changes will not modify, degrade, or prevent 
actions or alter any assumptions previously made in evaluating the 
radiological consequences of any accident described in the Updated 
Final Safety Analysis Report (UFSAR). The ESFAS instrumentation 
remains capable of performing its intended safety function of 
mitigation of consequences of accidents or transients. Therefore, 
the consequences of an 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.
    2. Does the proposed change create the possibility of a new or 
different kind accident from any accident previously evaluated?
    The proposed changes do not alter the performance of the ESFAS. 
The proposed changes only extend the STI, and no changes to the 
testing methodology or the way in which the slave relays are tested 
are being proposed. No new equipment is being installed, and no 
installed equipment is being operated in a new or different manner 
with the proposed changes. Extending the STI will maintain the 
reliability of the slave relays as demonstrated by the NRC approved 
FMEA and aging assessment, and may improve the reliability of the 
system by reducing potential test-induced degradation. As documented 
in WCAP-13877 and WCAP-13878, an STI of 92 days is no more likely to 
detect significant changes in the SSPS Type AR and MDR Series slave 
relays than a STI of 18 months.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes do not affect the total ESFAS response 
assumed in the safety analysis. The periodic slave relay functional 
verification is relaxed because of the demonstrated high reliability 
of the slave relays and their insensitivity to any short-term wear 
or aging effects. The Westinghouse Owners Group (WOG) program to 
extend the STI for the slave relays, as documented in the NRC 
approved WCAP-13877 and WCAP-13878, has concluded that the slave 
relays used in the SSPS are highly reliable and that the 
surveillance testing at a frequency of 18

[[Page 11054]]

months, instead of the 92-day STI currently required, does not 
significantly decrease any margin of safety assumed in the safety 
analysis. Plant safety will be improved by limiting the amount of 
on-line testing that will be performed, because on-line testing of 
the slave relays results in the removal of a train of equipment from 
service or manipulation of specific safety-related equipment which 
is then no longer able to perform its safety function if called upon 
until the surveillance test is completed. The proposed changes also 
act to improve plant safety by reducing equipment degradation and 
reducing unnecessary burden on the operating personnel. There are no 
changes in testing methodology or performance criteria.
    Therefore, 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 50.929c) 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 60609-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

    Date of amendment request: November 13, 2000.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to delete the ``Power Range Neutron 
Flux High Negative Rate,'' Trip Function from Reactor Trip System 
Instrumentation. The proposed change allows elimination of this 
unnecessary function and thereby reduces the potential for a transient. 
The proposed changes are consistent with the Westinghouse Topical 
report previously accepted by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of he issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The removal of the Power Range Neutron Flux High Negative Rate 
Trip (i.e., Negative Flux Rate Trip (NFRT)) Function does not 
increase the probability or consequences of reactor core damage 
accidents resulting from dropper Rod Cluster Control Assembly (RCCA) 
events previously analyzed. The safety functions of other safety 
related systems and components, which are related to mitigation of 
these events, have not been altered. All other primary Reactor Trip 
System (RTS) and Engineered Safety Features Actuation Systems 
(ESFAS) protection functions are not impacted by the elimination of 
the NFRT Function. The NFRT circuitry detects and responds to 
negative reactivity insertion due to RCCA misoperation events should 
they occur. Therefore, the NFRT Function is not assumed in the 
initiation of such events. Because the NFRT Function is being 
eliminated from the plant, it can no longer actuate and cause a 
transient. The consequences of accidents previously evaluated in the 
Updated Final Safety Analysis Report (UFSAR) are unaffected by the 
proposed changes because no change to any equipment response or 
accident mitigation scenario has resulted.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind accident from any accident previously evaluated?
    The deletion of the NFRT Function does not create the 
possibility of a new or different kind of accident than any accident 
previously evaluated in the UFSAR. No new accident scenarios, 
failure mechanisms, or limiting single failures are introduced as a 
result of the proposed changes. The proposed changes do not 
challenge the performance or integrity of any safety related 
systems. It has been demonstrated that the NFRT Function can be 
eliminated by the NRC approved methodology described in Westinghouse 
Topical Report WCAP-11394-P-A, ``Methodology for the Analysis of the 
Dropped Rod Event,'' dated January 1990. The Braidwood Station and 
the Byron Station cycle-specific analyses have confirmed that for a 
dropped RCCA(s) event, no direct reactor trip or automatic power 
reduction is required to meet the Departure From Nucleate Boiling 
(DNB) limits for this Condition II, ``Faults of Moderate 
Frequency,'' event. The NFRT Function is not credited either as a 
primary or backup mitigation feature for any other UFSAR event. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The margin of safety associated with the licensing basis 
acceptance criteria for any postulated accident is unchanged. It has 
been demonstrated that the NFRT Function can be eliminated by the 
NRC approved methodology described in WCAP 11394-P-A. The Braidwood 
Station and the Byron Station cycle-specific analyses have confirmed 
that for a dropped RCCA(s) event, DNB limits are not exceeded with 
the proposed changes. Conformance to our licensing basis acceptance 
criteria for Design Basis Accidents (DBAs) and transients with the 
deletion of the NFRT Function is demonstrated, and DNB limits are 
not exceeded. The proposed changes will have no adverse effect on 
the availability, operability, or performance of the safety related 
systems and components assumed to actuate in the event of a DBA or 
transient. Therefore, 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. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: February 29, 2000.
    Description of amendment request: The proposed amendment would 
reduce the number of safety valves required for overpressure protection 
at Dresden, Unit 2, by removing from Technical Specifications (TS) 
Section 3.6.E, the safety valve function of the Target Rock safety/
relief valve (SRV). The proposed amendment would move the safety valve 
lift pressure setpoints from TS Section 3.6.E to TS Section 4.6.E, 
remove the Target Rock SRV setpoints from TS, and change the number of 
safety valves from nine to eight. The proposed amendment would also 
remove footnote ``c'' of Unit 3, TS Section 4.6.E, since this footnote 
was only applicable to Unit 3, Cycle 15 which has been completed.
    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?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those

[[Page 11055]]

consequences. Limits have been established, consistent with Nuclear 
Regulatory Commission (NRC) approved methods to ensure that fuel 
performance during normal, transient, and accident conditions is 
acceptable. The proposed change to reduce the number of required 
safety valves from nine (9) to eight (8) does not affect the ability 
of plant systems to adequately mitigate the consequences of an 
accident previously evaluated.
    This conclusion was derived by evaluating all applicable 
analyses including thermal limit, American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel (B&PV) pressurization 
events, margin to unpiped safety valve, anticipated transient 
analysis without scram events, Loss Of Coolant Accident (LOCA), 
station blackout, and 10 CFR 50, Appendix R analyses. Therefore, 
there is no increase in the probability or consequences of an 
accident previously evaluated because the analyses supports 
operation without crediting the Target Rock Safety Relief Valve 
safety mode function.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The requested change has been previously evaluated by evaluating 
all applicable analyses including thermal limit, ASME B&PV 
pressurization events, margin to unpiped safety valve, anticipated 
transient analysis without scram events, station blackout, LOCA, and 
10 CFR 50, Appendix R analyses. The proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated because the analyses support operation 
without crediting the Target Rock safety relief valve safety 
function. No new failure modes will be introduced upon 
implementation of the proposed changes, therefore, the possibility 
of a new and different accident has not been created.
    Does the change involve a significant reduction in a margin of 
safety?
    Changing the required number of safety valves from nine (9) to 
eight (8) will not involve any reduction in margin of safety. This 
conclusion was derived by evaluating all existing analyses including 
thermal limit, ASME B&PV pressurization events, margin to unpiped 
safety valve, anticipated transient analysis without scram events, 
station blackout, LOCA, and 10 CFR 50, Appendix R analyses. The 
analyses previously evaluated remain valid, therefore, a significant 
reduction in the margin of safety does not exist.
    Therefore, based upon the above evaluation, ComEd has concluded 
that these changes do not constitute a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: February 14, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to correct various editorial 
errors and make other administrative changes. Specifically, the 
proposed amendment would make administrative changes that revise: (a) 
Tables 3.6-1 and 4.4-1 to correct listing and editorial errors, (b) TS 
3.8.B.10 to reflect the wording in 10 CFR 50.54(m)(2)(iv), (c) Figures 
3.10-2 through 3.10-6 to remove these figures, (d) Table 4.1-1 to 
reflect change in level indication components, (e) TS 4.19.B and 
6.14.1.1 to correct editorial errors, (f) TS 6.12.1 to reference the 
current sections of 10 CFR Part 20, (g) TS 6.12.1 to reflect an 
organizational title change, and (h) TS 6.13.2 to correct a 
typographical error.
    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) Changes To Tables 3.6-1 And 4.4-1 To Correct Listing And 
Editorial Errors
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed changes are administrative in nature. The 
changes involve correcting errors in Table 3.6-1 and additions to 
Tables 3.6-1 and 4.4-1 to reflect UFSAR [Updated Final Safety 
Analysis Report] Table 5.2-1 and the IST [inservice testing] 
Program. These changes do not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operation of the facility. The Limiting Safety System Settings and 
Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed changes would not involve 
a significant increase in the probability or in the consequences of 
an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative in nature. The 
safety analysis of the facility remains complete and accurate. There 
are no physical changes to the facility and the plant conditions for 
which the design basis accidents have been evaluated are still 
valid. The operating procedures and emergency procedures are 
unaffected. Consequently no new failure modes are introduced as a 
result of the proposed changes. Therefore, the proposed changes 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes are administrative in nature. Since 
there are no changes to the operation of the facility or the 
physical design, the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    (b) Change To Section 3.8.B.10 To Reflect The Wording In 10 CFR 
50.54(m)(2)(iv)
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The change 
involves updating Section 3.8.B.10 to reflect 10 CFR 
50.54(m)(2)(iv). This change does not affect possible initiating 
events for accidents previously evaluated or alter the configuration 
or operation of the facility. The Limiting Safety System Settings 
and Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed change would not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    (c) Deletion Of Figures 3.10-2 Through 3.10-6
    (1) Does the proposed license amendment involve a significant 
increase in the

[[Page 11056]]

probability or in the consequences of an accident previously 
evaluated?
    No. The proposed change is administrative in nature. The change 
involves the deletion of Figures 3.10-2, 3.10-3, 3.10-4, 3.10-5 and 
3.10-6. This change does not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operation of the facility. The Limiting Safety System Settings and 
Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed change would not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    (d) Change To Table 4.1-1 To Reflect Change In Level Indication 
Components
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. This change does not affect possible initiating events for 
accidents previously evaluated or operation of the facility. While 
the configuration of the facility has changed with installation of 
the new level sensors, the safety-related function of theses sensors 
remains unchanged (i.e., at a predetermined level of approximately 
35% of instrument span, a low level alarm will annunciate in the CCR 
[control room]). The Limiting Safety System Settings and Safety 
Limits specified in the current Technical Specifications remain 
unchanged. Therefore, the proposed change would not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The safety analysis of the facility remains complete and 
accurate. The plant conditions for which the design basis accidents 
have been evaluated are still valid. While the configuration of the 
facility has changed with installation of the new level sensors, the 
safety-related function of theses [sic] sensors remains unchanged 
(i.e., at a predetermined level of approximately 35% of instrument 
span, a low level alarm will annunciate in the CCR). Consequently no 
new failure modes are introduced as a result of the proposed change. 
Therefore, the proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. While the configuration of the facility has changed with 
installation of the new level sensors, the safety-related function 
of theses sensors remains unchanged (i.e., at a predetermined level 
of approximately 35% of instrument span, a low level alarm will 
annunciate in the CCR). Also, there are no changes to the operation 
of the facility. Thus the Updated Final Safety Analysis Report 
(UFSAR) design basis, accident assumptions, or Technical 
Specification Bases are not affected. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    (e) Change To Sections 4.19.B And 6.14.1.1 To Correct Editorial 
Errors
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed changes are administrative in nature. The 
change in Sections 4.19.B and 6.14.1.1 involve amending ``the 
Semiannual Radioactive Effluent Release Report'' to ``the Annual 
Radioactive Effluent Release Report.'' These changes do not affect 
possible initiating events for accidents previously evaluated or 
alter the configuration or operation of the facility. The Limiting 
Safety System Settings and Safety Limits specified in the current 
Technical Specifications remain unchanged. Therefore, the proposed 
changes would not involve a significant increase in the probability 
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative in nature. The 
safety analysis of the facility remains complete and accurate. There 
are no physical changes to the facility and the plant conditions for 
which the design basis accidents have been evaluated are still 
valid. The operating procedures and emergency procedures are 
unaffected. Consequently no new failure modes are introduced as a 
result of the proposed change. Therefore, the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes are administrative in nature. Since 
there are no changes to the operation of the facility or the 
physical design, the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    (f) Change To Section 6.12.1 To Reference The Current Sections 
Of 10 CFR [Part] 20
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The change 
involves updating Section 6.12.1 to reference 10 CFR 20.1601(a) and 
10 CFR 20.1601(b). This change does not affect possible initiating 
events for accidents previously evaluated or alter the configuration 
or operation of the facility. The Limiting Safety System Settings 
and Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed change would not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    (g) Change To Section 6.12.1 To Reflect An Organizational Title 
Change
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The change 
involves updating Section 6.12.1 to use the title ``Shift Manager'' 
instead of ``Senior Watch Supervisor.'' This change does not affect 
possible initiating events for accidents previously evaluated or 
alter the configuration or operation of the facility. The Limiting 
Safety System Settings and Safety Limits specified in the current 
Technical Specifications remain unchanged. Therefore, the proposed 
change would not involve a significant increase in the probability 
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of

[[Page 11057]]

accident from any accident previously evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    (h) Change To Section 6.13.2 To Correct A Typographical Error
    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed change is administrative in nature. The change 
involves updating Section 6.13.2 from ``DOR [Division of Operating 
Reactors] Guidelines of NUREG-0588'' to ``DOR Guidelines or NUREG-
0588.'' This change does not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operation of the facility. The Limiting Safety System Settings and 
Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed change would not involve a 
significant increase in the probability or in the consequences of an 
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change. Therefore, the proposed change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed change is administrative in nature. Since there 
are no changes to the operation of the facility or the physical 
design, the Updated Final Safety Analysis Report (UFSAR) design 
basis, accident assumptions, or Technical Specification Bases are 
not affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: December 11, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to provide editorial 
revisions, clarifications, and corrections. Specifically, the proposed 
amendment would: (1) Provide updated information and corrections to the 
TS cover page, table of contents, and list of figures, (2) revise TS 
4.5.E, ``Control Room Air Filtration System,'' to remove an incorrect 
system test description and provide consistent test values for system 
flow rate and filter efficiency, (3) revise TS 6.2.1.a, ``Facility 
Management and Technical Support,'' to reference the Quality Assurance 
Program Description as the location of the documentation rather than 
the Updated Final Safety Analysis Report, (4) revise TS 6.9.1.7, 
``Monthly Operating Report,'' to change the recipient of the Monthly 
Operating Report, and (5) correct the periodicity of the Radioactive 
Effluent Release Report from annual to semiannual in TS 6.15, ``Offsite 
Dose Calculation Manual'' and TS 6.16, ``Major Changes to Radioactive 
Liquid, Gaseous and Solid Waste Systems.'' In addition, the proposed 
change revises TS Figure 5.1-1B concerning the indicated vent location 
associated with Indian Point Unit 3 (IP3). The labels for the plant 
vent and the machine shop are reversed.
    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?
    The proposed changes consist of editorial changes, 
administrative changes, clarifications, and corrections to existing 
TSs. These changes do not involve a change to the design or 
operation of any plant system nor are any of the safety analyses 
affected as a result of these changes. Accordingly, the initiators 
of any accident as well as any system relied upon for the mitigation 
of an accident are not affected by the proposed changes. Therefore, 
there is no increase in the probability or in the consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not involve a change to the design or 
operation of any plant system. These changes include editorial 
changes, administrative changes, clarifications, and corrections of 
existing TSs and, therefore, do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes consist of editorial changes, 
administrative changes, and clarifications to existing TSs and do 
not involve changes to any 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: January 8, 2001.
    Description of amendment request: The proposed change revises the 
lower limit of the allowable containment internal pressure in Technical 
Specification (TS) 3.6.1.4, ``Containment Systems--Internal Pressure,'' 
from 14.375 pounds per square inch, absolute (psia) to 14.275 psia.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The proposed change revises the lower limit of the 
allowable containment internal pressure in TS 3.6.1.4 from 14.375 to 
14.275 psia. This change will allow

[[Page 11058]]

additional operating margin for the containment atmosphere purge 
(CAP) system during conditions of low atmospheric pressure. The 
containment minimum pressure parameter is not an accident initiator 
and does not affect the probability of any initiating event 
scenario. Although the TSs will allow a lower initial containment 
internal pressure, the current analyses for the associated design 
events are not affected since the lower pressure has already been 
conservatively included. The proposed change in initial containment 
internal pressure is bounded in the current design. Therefore, this 
proposed change does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. Will the operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change affects the TS allowed lower limit 
on containment internal pressure and consequently the atmospheric 
range in which the CAP system can be operated. The change in the 
lower limit on containment internal pressure is encompassed by 
current design analyses and does not result in a change of analyzed 
conditions or analyzed operating ranges.
    Based on the proposed TS change, CAP system operation will be 
allowed at a lower atmospheric pressure. This change does not change 
the function of the system or its method of operation. Although the 
initial atmospheric pressure at which the CAP system can be 
initiated is being lowered, this is within the current design of the 
CAP system and does not change the differential pressures at which 
it will be operated.
    Therefore, the proposed change[d] does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will the operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed change makes use of the initial 
containment pressure assumption values used in the current analyses 
to provide additional operating margin for the CAP system. The 
margin of safety that was inherent in the results of these safety 
analyses has been preserved. The associated analyses ensure the 
negative pressure differential associated with an inadvertent 
actuation of the containment spray system is acceptable, and ensure 
that the emergency core cooling system can satisfy its design safety 
function under worst case conditions. The calculated maximum 
differential pressure is 0.49 psid [pounds per square inch 
differential] which is within the design limit of 0.65 psid. The 
peak clad temperature for the worst case large break loss of coolant 
accident is 2177 deg.F which is within the acceptance criteria given 
in 10CFR50.46. Since the proposed change does not affect the initial 
containment pressure utilized in these analyses, the results of the 
analyses are unchanged. Therefore, there is no affect on any margin 
of safety associated with this parameter.
    Based on the above No Significant Hazards Consideration 
Determination, it is concluded that: (1) The proposed change does 
not constitute a significant hazards consideration as defined by 
10CFR50.92; (2) there is a reasonable assurance that the health and 
safety of the public will not be endangered by the proposed change; 
and (3) this action will not result in a condition which 
significantly alters the impact of the station on the environment as 
described in the NRC final environmental statement.

    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 1, 2000.
    Description of amendment request: The proposed amendments would 
revise the technical specifications to add a new requirement for the 
Main Steam Line Radiation Monitor 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 of occurrence or consequences of an accident previously 
evaluated?
    The addition of the MSLRM [Main Steam Line Radiation Monitor] 
automatic trip signal to the MVP [mechanical vacuum pump] has no 
adverse effect on safety. The addition of Surveillance Requirements 
(SRs) and the Limiting Condition for Operation (LCO) to our TS 
enhances current safety features of the plant by establishing 
controls for a required, and currently functional, safety feature. 
The automatic trip function of the MVP does not serve as an 
initiator for any accidents evaluated in Chapter 15, ``Accident and 
Transient Analysis,'' of the Updated Final Safety Analysis Report. 
Therefore, this change will not result in an increase of either the 
probability or consequences of an accident.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    These proposed changes involve the addition of the MVP trip 
input from the Main Steam Line Tunnel High Radiation signal. The 
addition of this function does not represent a change in operating 
parameters or equipment configuration for DNPS [Dresden Nuclear 
Power Station], Units 2 and 3. Operation of DNPS, Units 2 and 3, 
under the proposed changes 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?
    These proposed changes create a TS LCO and identify SRs for the 
MVP trip input from the MSLRM signal. Operation under the proposed 
change will not change any plant operation parameters, nor any 
protective system setpoints. The calculations of off site dose 
demonstrate that with the MVP trip instrumentation operating 
properly, the doses that result from a CRDA [control rod drop 
accident] with the MVP operating 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.

Exelon Generating Company, LLC (Exelon), Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: November 20, 2000.
    Description of amendment request: PECO Energy Company (PECO) 
proposed changes to the Technical Specifications (TSs) that would 
revise the heatup, cooldown, and inservice test Pressure-Temperature 
(P-T) limitations (TS Figure 3.4.6.1-1) of the Limerick Generating 
Station (LGS), Unit 2, Reactor Pressure Vessel (RPV) to a maximum of 32 
Effective Full Power Years (EFPY). In addition, the licensee proposed 
text changes to both Limiting Condition for Operation 3.4.6.1 and 
Surveillance Requirement 4.4.6.1.1 to delete the reference to the A' 
curve on TS Figure 3.4.6.1-1 since this curve will not be included in 
the proposed Figure 3.4.6.1-1. The licensee also proposed adding an 
intermediate hydrotest curve (Curve A22) to TS Figure 
3.4.6.1-1, which is valid to 22 EFPY. By letter dated January 30, 2001, 
Exelon stated that it has assumed responsibility, as of the date of the 
transfer, for the active items on the Limerick Units 1 and 2

[[Page 11059]]

dockets previously submitted by PECO, including the subject amendment 
request.
    Moreover, Exelon is revising its TS Bases Section B 3/4.4.6 to 
update several RPV material chemistry parameters. The licensee 
identified the need for these revisions during a Certified Material 
Test Report data search performed by General Electric Company during 
development of the proposed P-T curves.
    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 analysis is 
presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There are no physical changes to the plant being introduced by 
the proposed changes to the P-T curves. The proposed changes do not 
modify the reactor coolant pressure boundary, i.e., there are no 
changes in operating pressure, materials or seismic loading. The 
proposed changes do not adversely affect the integrity of the 
reactor coolant pressure boundary such that its function in the 
control of radiological consequences is affected. The proposed P-T 
curves were generated in accordance with the fracture toughness 
requirements of 10 CFR 50, Appendix G, and American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 
Section XI, Appendix G, in conjunction with ASME Code Case N-640. 
The proposed P-T curves were established in compliance with the 
methodology used to calculate the predicted irradiation effects on 
vessel beltline materials. Usage of these procedures provides 
compliance with the intent of 10 CFR 50, Appendix G, and provides 
margins of safety that ensure that failure of the reactor vessel 
will not occur. The proposed P-T curves prohibit operational 
conditions in which brittle fracture of reactor vessel materials is 
possible. Consequently, the primary coolant pressure boundary 
integrity will be maintained. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the P-T curves were generated in 
accordance with the fracture toughness requirements of 10 CFR 50, 
Appendix G, and ASME B&PV Code, Section XI, Appendix G, in 
conjunction with ASME Code Case N-640. Compliance with the proposed 
P-T curves will ensure that conditions in which brittle fracture of 
primary coolant pressure boundary materials are possible will be 
avoided. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any new failure mode 
from previously evaluated accidents. Further, the proposed changes 
to the P-T curves do not affect any activities or equipment, and are 
not assumed in any safety analysis to initiate nor mitigate any 
accident sequence. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes reflect an update of the P-T curves to 
extend the reactor pressure vessel operating limit to 32 Effective 
Full Power Years (EFPY). The revised curves are based on the latest 
ASME guidance. The revised P-T limits, which provide more 
operational flexibility than the current limits, were established in 
accordance with current regulations and the latest ASME Code 
information. No plant safety limits, set points, or design 
parameters are adversely affected by the proposed TS changes. These 
proposed changes maintain the relative margin of safety commensurate 
with that which existed at the time that the ASME B&PV Code, Section 
XI, Appendix G, was approved in 1974.
    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 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, Exelon Generating Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC (Exelon), Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: January 18, 2001.
    Description of amendment request: Exelon requested a Technical 
Specification (TS) change which will revise Surveillance Requirement 
(SR) 4.9.2.d.1 to clarify that ``shorting links'' do not need to be 
removed, if adequate shutdown margin has been demonstrated, when moving 
a control rod during Operational Condition 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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 change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This TS Change Request revises SR 4.9.2.d.1 to clarify that 
``shorting links'' do not need to be removed if adequate shutdown 
margin has been demonstrated when a control rod is withdrawn during 
Operational Condition 5. This revision ensures that the words and 
intent of the SR 4.9.2.d.1 match the words and intent of Limiting 
Condition for Operation (LCO) 3.9.2.d, and will improve the 
readability of the SR for plant operators. This change to SR 
4.9.2.d.1 will clarify that ``shorting links'' can remain installed 
if adequate shutdown margin has been demonstrated any time a control 
rod is withdrawn in Operational Condition 5. This revision does not 
impact any accident or transient events. There are no new initiators 
created by this revision. Additionally, this revision will not 
impact any existing analyses or requirements contained in the 
Updated Final Safety Analysis Report. No changes in the operation of 
the plant during power operation or refueling will occur as a result 
of this revision. Therefore, the proposed TS revision does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed TS revision does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS revision will not impact any physical changes to 
plant structures, systems, or components. The design, function, and 
reliability of the Reactor Protection System will not be impacted by 
this revision. This revision does not adversely impact any equipment 
which is required for the prevention or mitigation of accidents or 
transients. This revision ensures that the words and intent of the 
SR 4.9.2.d.1 match the words and intent of LCO 3.9.2.d, and will 
improve the readability of the SR for plant operators. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    This proposed revision to SR 4.9.2.d.1 does not affect any 
safety limits or analytical limits. There are also no changes to 
accident or transient core thermal hydraulic conditions, minimum 
combustible concentration limits, or fuel or reactor coolant 
boundary design limits, as a result of this proposed change. This 
revision ensures that the words and intent of the SR 4.9.2.d.1 match 
the words and intent of LCO 3.9.2.d. Therefore, the proposed change 
does not involve a significant reduction in a 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

[[Page 11060]]

Counsel, Exelon Generating Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC (Exelon), Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania
    Date of amendment request: January 18, 2001.
    Description of amendment request: Exelon requested a Technical 
Specification (TS) change which will revise the Units 1 and 2 TS Table 
1.2, ``Operational Conditions,'' to allow placing the reactor mode 
switch to the REFUEL position during Operational Conditions 3 and 4 to 
accommodate post maintenance and surveillance testing on control rod 
drives.
    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 change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision allows a single control rod to be 
withdrawn under control of the reactor mode switch REFUEL position 
and the one-rod-out interlock in Operational Conditions 3 and 4. 
This change does not affect any existing accident initiators. There 
is no change to the coupling integrity of the control rod during 
this accident. Although this change would allow an increase in the 
frequency of single control rod withdrawals in Operational 
Conditions 3 and 4, the probability of the previously analyzed 
accidents is not affected.
    The onsite and offsite radiological consequences of previously 
analyzed accidents in Operational Conditions 3 and 4 are not 
affected by this proposed change. This change does not affect any 
existing accident mitigators. The shutdown margin combined with the 
refueling interlocks prevent a rod withdrawal error while in 
refueling thereby preventing inadvertent criticality. There is no 
impact on the ability of the Reactor Protection System (RPS) 
circuitry to mitigate a Control Rod Drop Accident as described in 
the Safety Analysis Report, nor is there an increase in the number 
of fuel failures from this accident. As a result, the probability 
and consequences of previously analyzed accidents are not 
significantly increased.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    There are no new accident initiators created by the proposed 
revision to Table 1.2. Single control rods can be withdrawn in 
Operational Conditions 3 and 4 under the existing Technical 
Specifications to permit control rod recoupling. The proposed 
revision would expand this provision to other control rod 
maintenance and testing activities performed in Operational 
Conditions 3 and 4. The withdrawal of individual control rods in 
Operational Conditions 3 and 4 is a mode of operation permitted 
under limited circumstances by the existing TSs. The additional 
control rod maintenance and testing activities which could be 
performed in Operational Conditions 3 and 4, are already permitted 
by the existing TSs in Operational Conditions 1, 2, 4, and 5.
    Based on the above, this change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The TSs currently permit single control rod withdrawal for the 
purpose of control rod recoupling when in Operational Conditions 3 
or 4 if the one-rod-out interlock is Operable. This change allows 
additional activities for which a single control rod may be 
withdrawn when in Operational Conditions 3 or 4, with the same 
restriction that the one-rod-out interlock be Operable.
    The operability requirements for the one-rod-out interlock and 
the shutdown margin requirements of TS 3.1.1 ensure the reactor will 
be maintained subcritical during single control rod withdrawals. 
Therefore, this change will not involve a significant reduction in a 
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, Exelon Generation Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Exelon Generating Company, LLC (Exelon), Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania.

    Date of amendment request: February 1, 2001.
    Description of amendment request: Exelon proposed changes that 
would revise Technical Specification (TS) 2.1 to incorporate revised 
Safety Limit Minimum Critical Power Ratios due to the cycle-specific 
analysis performed by Global Nuclear Fuel for Limerick Generating 
Station, Unit 2, Cycle 7, which will include the use of the GE-14 fuel 
product line.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The derivation of the cycle specific Safety Limit Minimum 
Critical Power Ratios (SLMCPRs) for incorporation into the Technical 
Specifications (TS), and its use to determine cycle specific thermal 
limits, has been performed using the methodology discussed in 
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US, 
June, 2000, which incorporates Amendment 25. Amendment 25 was 
approved by the NRC [Nuclear Regulatory Commission] in a March 11, 
1999 safety evaluation report.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling. The GE-14 fuel is in compliance with 
Amendment 22 to ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing 
acceptance criteria. The probability of fuel damage will not be 
increased as a result of this change. Therefore, the proposed TS 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, calculated to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core if the limit is not violated. The new SLMCPRs are calculated 
using NRC approved methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. Additionally, the GE-14 fuel is in 
compliance with Amendment 22 to ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and 
U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which provides the 
fuel licensing acceptance criteria. The SLMCPR is not an accident 
initiator, and its revision will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of the proposed change to 
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs 
are calculated using

[[Page 11061]]

methodology discussed in ``General Electric Standard Application for 
Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, 
NEDE-24011-P-A-14-US, June, 2000, which incorporates Amendment 25. 
The SLMCPRs ensure that greater than 99.9% of all fuel rods in the 
core will avoid transition boiling if the limit is not violated when 
all uncertainties are considered, thereby preserving the fuel 
cladding integrity. Therefore, the proposed TS change will not 
involve a significant reduction in the margin of safety previously 
approved by the NRC.

    Based on the staff's review of the licensee's evaluation, 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, Exelon Generating Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: January 13, 2000.
    Description of amendment request: The proposed amendment would 
change the Kewaunee Nuclear Power Plant Technical Specification 3.6, 
``Containment.'' The proposed amendment would add limiting condition 
for operation and allowed outage times for containment penetrations and 
associated isolation devices to provide clear guidance. Also, the 
proposed amendment would provide additional information, clarification, 
and uniformity to the basis of the associated technical specification.
    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 Technical Specification [TS] change provides definition for 
the [Allowable Outage Time] AOT for a containment isolation valve 
and containment air lock. The original design and design basis of 
the plant is still maintained and the probability and consequences 
of previously evaluated accidents is unchanged. In our current 
Technical Specifications the allowed outage time for a safeguards 
480-volt bus is 24 hours. The basis for this outage time states:
    ``The intent of this TS is to provide assurance that at least 
one external source and one standby source of electrical power is 
always available to accomplish safe shutdown and containment 
isolation and to operate required engineered safety features 
equipment following an accident.''
    With one 480-volt safeguards bus out of service an associated 
motor operated containment isolation valve is also out of service. 
Since the 24-hour AOT is part of Kewaunee's original design basis, 
allowing the containment isolation valves to be out of service for 
24 hours does not increase the probability or consequences of an 
accident previously evaluated.
    A risk assessment of the probability of a -loss-of-coolant-
accident with a train of containment isolation failing during a 4-
hour verse a 24-hour time span was conducted. The probability of 
[loss-of-coolant accident] LOCA coincident with the failure of 
containment isolation occurring during a 4-hour period versus a 24-
hour period is shown on Figure 1[ in licensee's submittal]. This 
change in probability is considered insignificant.
    The proposed TS changes do not involve any physical or 
operational changes to structures, systems or components. The 
current safety analysis and design basis for the accident mitigation 
functions of the containment, the airlocks, and the containment 
isolation valves are maintained. On-site and off-site dose 
consequences remain unaffected.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The function of the containment vessel is to contain the 
radiologically hazardous material following a LOCA. By maintaining 
at least one containment isolation barrier intact the vessel can 
perform its function. This amendment still ensures that at least one 
barrier is intact or the leakage is evaluated not to exceed that 
which is already evaluated and allowed by current technical 
specification.
    The accidents considered are found in the Safety Analysis, 
Section 14 of the [Updated Safety Analysis Report] USAR. The 
proposed change does not involve a change to the plant design 
(structures, systems or components) or operation. No new failure 
mechanisms beyond those already considered in the current plant 
Safety Analysis are introduced. No new accident is introduced and no 
safety-related equipment or safety functions are altered. The 
proposed change does not affect any of the parameters or conditions 
that contribute to initiation of any accidents.
    3. Involve a significant reduction in a margin of safety.
    With one containment barrier intact during plant operation the 
isolation of containment is still ensured. The plant's original 
design basis addressed the inability of one of the two containment 
isolation valves to operate for a 24-hour period. As this AOT has 
been previously considered, there therefore is no reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: January 18, 2001.
    Description of amendment request: The proposed amendment would 
change the Kewaunee Nuclear Power Plant Technical Specification 3.10.m 
for reactor coolant minimum flow from the current value of 85,500 
gallons per minute (gpm) to 93,000 gpm due to the replacement of steam 
generators scheduled for the fall 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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The change in Reactor Coolant Minimum Flow value for TS 3.10.m 
proposed in this amendment request is needed to reflect operating 
characteristics of the new [Replacement Steam Generators] RSGs. 
Accident analyses affected by the RSGs have each been evaluated to 
establish that there is no significant change in the documented 
results (Attachment 3). These evaluations have shown that the 
proposed value for Reactor Coolant Minimum Flow is bounded by the 
Thermal Design Flow value used in the analyses and provides greater 
margin to safety analysis acceptance criteria (e.g., [Departure from 
Nucleate Boiling] DNB). All safety analysis acceptance criteria are 
satisfied. Since Reactor Coolant flow values for the RSG conform to 
the design bases and are bounded by the existing safety analyses, 
changing the technical specification within limits of the bounding 
accident analyses will not cause an 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.
    The proposed change is fully consistent with current plant 
design bases and does not adversely affect any fission product 
barrier, nor does it alter the safety function of safety related 
systems, structures, and components depended upon for accident 
prevention or mitigation. Thus, it does not create the possibility 
of a new or different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting

[[Page 11062]]

Safety System Setpoints, or Limiting Conditions for Operation are 
determined. It returns TS 3.10.m for Reactor Coolant Minimum Flow to 
a value slightly higher, thus more conservative, than the value 
specified for the [Original Steam Generators] OSG when new. It 
conforms to plant design bases, is consistent with current safety 
analyses, and limits actual plant operation. Analysis of the effect 
of the proposed Reactor Coolant Minimum Flow limitation on [Loss-of-
Coolant-Accident] LOCA and non-LOCA transients determined that all 
safety analysis acceptance criteria are satisfied at a [Thermal 
Design Flow] TDF that bounds the revised Reactor Coolant Minimum 
Flow and all [Kewaunee Nuclear Power Plant] KNPP safety requirements 
continue to be met. Therefore, the proposed change does not involve 
a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: February 5, 2001.
    Description of amendment request: The proposed amendment would 
change the Kewaunee Nuclear Power Plant Technical Specification 3.1.d.2 
to reduce the maximum allowable leakage of primary system reactor 
coolant to the secondary system from 500 gallons per day (gpd) through 
any one steam generator to 150 gpd through any one steam generator. In 
addition, the proposed amendment would remove reference to voltage 
based repair criteria.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The change in Leakage of Reactor Coolant value proposed by this 
request for [Technical Specification] TS 3.1.d.2 complies with 
[Nuclear Energy Institute] NEI 97-06, ``Steam Generator Program 
Guidelines.'' [Nuclear Management Company, LLC] NMC evaluated 
accident analyses affected by [steam generator] SG tube leakage and 
determined that this change continues to be bounded by the existing 
licensing and design basis. Design basis accidents and transients, 
including steam generator tube rupture (SGTR), were analyzed using 
Westinghouse Model 54F steam generator assumptions as part of steam 
generator replacement. These evaluations show that the proposed 150 
gpd [gallons per day] value for Leakage of Reactor Coolant is 
bounded by the larger value used in applicable existing design basis 
accident and transient analyses. The 150 gpd leak rate provides 
increased margin to acceptance criteria found in these analyses. All 
acceptance criteria are satisfied and SG primary to secondary 
leakage values for the [replacement steam generator] RSG conform to 
the existing design bases and are bounded by the existing safety 
analyses. Changing the technical specification within limits of the 
bounding accident analyses cannot change the probability or 
consequence of an accident previously evaluated. Removal of an 
allowance for voltage-based alternate repair criteria defaults to a 
more conservative repair criteria. Thus, nothing in this proposal 
will cause an increase in the probability or consequence of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The 150 gpd value proposed for maximum allowable Leakage of 
Reactor Coolant is consistent with current plant design bases and 
does not adversely affect any fission product barrier, nor does it 
alter the safety function of safety significant systems, structures 
and components or their roles in accident prevention or mitigation. 
The proposed value for maximum allowable leakage through any one 
steam generator is bounded by currently licensed design basis 
accident and transient analyses of record. Removal of a reference in 
the TS to voltage-based repair criteria leaves in its place a more 
conservative, more restrictive criteria for repair or plugging of 
steam generator tubes. Thus, this proposal does not create the 
possibility of a new or different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. It sets TS 3.1.d.2 for Leakage of Reactor 
Coolant to a lower, thus more conservative, value than that 
previously specified and approved for use by the NRC [Nuclear 
Regulatory Commission]. It conforms to plant design bases, is 
consistent with current safety analyses, and limits actual plant 
operation within analyzed and licensed boundaries. Analyses of 
applicable transients were performed using a primary to secondary 
leakage rate greater than the rate proposed by this request. All 
safety analysis acceptance criteria are satisfied at this value and 
all [Kewaunee Nuclear Power Plant] KNPP safety requirements continue 
to be met. The 150 gpd leak rate proposed by this amendment request 
is bounded by these analyses. Removal of reference to use of 
voltage-based repair criteria from TS 3.1.d.2 and its basis leaves 
an existing and more conservative repair criteria in place. Thus, 
changes proposed by this 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.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment requests: January 11, 2001.
    Description of amendment requests: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR

[[Page 11063]]

65018). The licensee affirmed the applicability of the following NSHC 
determination in its application dated January 11, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendment: January 18, 2001.
    Brief description of amendment request: The amendment request 
identifies an unreviewed safety question related to the planned 
replacement of the engineered safety features (ESF) transformers with 
new transformers having active automatic load tap changers (LTCs). 
Markups to the Final Safety Analysis Report (FSAR) were included in the 
application.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Based on the review of the modification details there is an 
insignificant increase in the probability of a malfunction of 
equipment important to safety, however there is no increase in the 
probability of an accident previously evaluated. The modification 
has no effect on the radiological consequences of accidents 
previously evaluated. Installation of the LTCs does not impact 
accident initiators though a failure mode has been identified that 
can increase the probability of malfunction, a risk study shows this 
risk is insignificant.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The overall effect of the malfunction of the LTC controllers 
would lead to a loss of the associated ESF bus which is not a new 
failure mode that can lead to a new or different kind of accident 
than previously evaluated. The LTC failure effects are limited to 
the associated ESF train, therefore this type of failure meets the 
definition of a single failure as defined in 10 CFR 50 Appendix A 
for operation under normal (Non T/S [technical specification] 
action) conditions. Additionally, during the 10 CFR 50.59 evaluation 
for the LTCs criteria (a)(2)(ii) with respect to accidents of a 
different type was not met.
    3. Does the proposed change involve a significant reduction in 
margin of safety?
    The installation of the replacement transformers with load tap 
changers will help assure the required minimum NB bus voltage 
established by Reference 7.10 [design calculations] under a wider 
variation of grid voltage.
    Current Technical Specification Bases for the offsite power 
distribution system are covered in sections B3.8.1-AC Sources-
Operating, B3.8.9-Distribution Systems-Operating, B3.8.2-AC Sources-
Shutdown, and B3.8.10-Distribution Systems-Shutdown. These bases 
ensure that sufficient power will be available to supply the safety-
related equipment required for: (1) The safe shutdown of the 
facility; and (2) The mitigation and control of accident conditions 
within the facility. The minimum specified independent and redundant 
AC power and distribution systems satisfy the requirements of 
General Design Criterion 17 of Appendix A to 10 CFR Part 50. The 
ACTIONS sections of the applicable Technical Specifications provide 
requirements specified for various levels of degradation of the 
power sources and provide restrictions upon continued facility 
operation commensurate with the

[[Page 11064]]

level of degradation. The Operability of the power sources are 
consistent with the initial condition assumptions of the safety 
analyses and are based upon maintaining at least one redundant set 
of onsite AC power sources and associated distribution systems 
operable during accident conditions coincident with an assumed loss 
of offsite power and single failure of the other onsite AC source.
    The installation of the transformers with automatic load tap 
changers reduces the possibility of the loss of the offsite power 
system due to the increased grid voltage variations as documented in 
the description of the change in section 4.1.4. Therefore, the 
installation of the transformers with load tap changers will not 
reduce 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Virginia Electric and Power Company, Docket No. 50-338, North Anna 
Power Station, Unit No. 1, Louisa County, Virginia

    Date of amendment request: January 9, 2001.
    Description of amendment request: The proposed administrative 
changes will remove obsolete license conditions from the Facility 
Operating License (FOL) and implement associated changes to the 
Technical Specifications (TS). These changes involve editorial 
revisions, relocation of license conditions, removal of redundant 
license conditions covered throughout the license, removal of expired 
license conditions, and removal of license conditions and TS associated 
with completed modifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--The proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to the North Anna Unit 1 Facility Operating 
License, NPF-4, is administrative (and in part editorial) in nature. 
The removal of license conditions regarding completed, no longer 
needed, and expired requirements has no impact on plant operations 
since these requirements no longer have meaningful applications. The 
renumbering and/or relocation within the FOL of various license 
conditions in this proposed administrative change does not alter the 
technical basis, requirements or the implementation of the affected 
items. The proposed change is within the current design and 
licensing bases of the facility. Since this change is administrative 
only and neither station operations nor design are affected by the 
change, it does not involve any significant increase in the 
probability or the consequences of any accident or malfunction of 
equipment important to safety previously evaluated.
    Criterion 2--The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change is administrative (and in part editorial) in 
nature. The license conditions that are being removed or relocated 
by this proposed change do not impact station operations or station 
equipment in any manner. The proposed change does not involve a 
physical alteration of the plant, nor a change in the methods used 
to respond to plant transients that has not been previously 
analyzed. No new or different equipment is being installed and no 
installed equipment is being removed or operated in a different 
manner. Consequently, no new failure modes are introduced and the 
proposed administrative change to the North Anna Unit 1 Facility 
Operating License does not create the possibility of a new or 
different kind of accident or malfunction of equipment important to 
safety from any previously evaluated.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed change is administrative (and in part editorial) in 
nature and neither station operations nor design are affected by the 
change. Since station operations are not affected by the proposed 
administrative change and no physical change is being made to the 
station, the change does not impact the condition, design, or 
performance of any station structure, system or component. 
Therefore, the proposed administrative change to the North Anna Unit 
1 Facility Operating License does not involve a significant 
reduction in any margin of safety described in the bases of the 
Technical Specifications.

    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: Maitri Banerjee, Acting.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit No. 2, Louisa County, Virginia

    Date of amendment request: January 9, 2001.
    Description of amendment request: The proposed administrative 
changes will remove obsolete license conditions from the Facility 
Operating License (FOL). These changes involve editorial revisions, 
relocation of license conditions, removal of expired license 
conditions, and removal of license conditions associated with completed 
modifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--The proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to the North Anna Unit 2 Facility Operating 
License, NPF-7, is administrative (and in part editorial) in nature. 
The removal of license conditions regarding completed, no longer 
needed, and expired requirements has no impact on plant operations 
since these requirements no longer have meaningful applications. The 
renumbering and/or relocation within the FOL of various license 
conditions in this proposed administrative change does not alter the 
technical basis, requirements or the implementation of the affected 
items. The proposed change is within the current design and 
licensing bases of the facility. Since this change is administrative 
only and neither station operations nor design are affected by the 
change, it does not involve any significant increase in the 
probability or the consequences of any accident or malfunction of 
equipment important to safety previously evaluated.
    Criterion 2--The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change is administrative (and in part editorial) in 
nature. The license conditions that are being removed or relocated 
by this proposed change do not impact station operations or station 
equipment in any manner. The proposed change does not involve a 
physical alteration of the plant, nor a change in the methods used 
to respond to plant transients that has not been previously 
analyzed. No new or different equipment is being installed and no 
installed equipment is being removed or operated in a different 
manner. Consequently, no new failure modes are introduced and the 
proposed administrative change to the North Anna Unit 2 Facility 
Operating License does not create the possibility of a new or 
different kind of accident or malfunction of equipment

[[Page 11065]]

important to safety from any previously evaluated.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed change is administrative (and in part editorial) in 
nature and neither station operations nor design are affected by the 
change. Since station operations are not affected by the proposed 
administrative change and no physical change is being made to the 
station, the change does not impact the condition, design, or 
performance of any station structure, system or component. 
Therefore, the proposed administrative change to the North Anna Unit 
2 Facility Operating License does not involve a significant 
reduction in any margin of safety described in the bases of the 
Technical Specifications.

    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: Maitri Banerjee, Acting.

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.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: January 8, 2001.
    Brief description of amendment request: The proposed amendment 
would revise the Technical Specifications (TS) to change the acceptance 
values for Core Spray subsystem flow contained in TS 4.5.1.b.1 from the 
current value of 6350 gallons per minute (gpm) to 6150 gpm.
    Date of publication of individual notice in Federal Register: 
January 22, 2001 (66 FR 6701).
    Expiration date of individual notice: February 21, 2001.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

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

    Date of application for amendments: June 16, 2000.
    Brief description of amendments: The amendments revise TS Table 
3.3.10-1, ``Post Accident Monitoring Instrumentation,'' to add the high 
pressure safety injection (HPSI) cold leg flow and HPSI hot leg flow 
instrumentation to the table.
    Date of issuance: February 8, 2001.
    Effective date: February 8, 2001, and shall be implemented within 
30 days of the date of issuance.
    Amendment Nos.: Unit 1--131 , Unit 2--131, Unit 3--131.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59220)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 08, 2001.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: September 14, 2000.
    Brief description of amendments: The amendments incorporate changes 
described below into the Technical Specifications for Culvert Cliffs 
Units 1 and 2. On September 9, 1996, a final rule amending 10 CFR 
50.55a was issued requiring owners to implement, by September 9, 2001, 
the requirements of the 1992 Edition through the 1992 Addenda of the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code Section XI, Subsections IWE and IWL, as modified and supplemented 
by 10 CFR 50.55a. Calvert Cliffs Nuclear Power Plant, Inc. has 
developed a program to effect the implementation of Subsections IWE and 
IWL.
    Date of issuance: January 30, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 240 and 214.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62384).

[[Page 11066]]

    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 19, 1999, as 
supplemented May 31, August 2, October 19, and November 21, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) by changing (1) the design features description of 
the fuel storage equipment and configuration to allow an increase in 
the spent fuel pool (SFP) storage capacity and (2) the description of 
the high-density spent fuel racks program to clarify that the 
surveillance program is applicable only to racks containing Boraflex as 
a neutron absorber. Specifically, the amendment revises the TSs for 
Fermi 2 to increase the capacity of the SFP from 2,414 to 4,608 fuel 
assemblies.
    Date of issuance: January 25, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications
    Date of initial in Federal Register March 13, 2000 (65 FR 13336)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated a January 25, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 20, 2000.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 5.5.7.d to decrease the maximum allowed pressure 
drops across control room emergency filtration (CREF) make-up and 
recirculation train filters and charcoal absorbers. The words ``(CREF 
only)'' are also removed from the TS.
    Date of issuance: February 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days
    Amendment No.: 142.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 

65340).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 30, 2000.
    Brief description of amendment: The amendment relocated the 
boration systems requirements from the Technical Specifications to the 
Technical Requirements Manual.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 229.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81916).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: November 19, 1999, as 
supplemented October 12, 2000.
    Brief description of amendment: The amendment changes the Technical 
Specification surveillance testing requirements of the charcoal 
adsorbers in the Standby Gas Treatment System and the Control Room 
Emergency Ventilation Air Supply System to meet the requested actions 
of Generic Letter 99-02.
    Date of issuance: February 5, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 269.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6410).
    The October 12, 2000, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 2001.
    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: May 22, 2000, as supplemented 
October 4, 2000.
    Brief description of amendments: Changed the Technical 
Specifications to incorporate that portion of the August 8, 1996, Final 
Amended Rule (61 FR 41303) related to revised requirement of inservice 
inspection of the containment post-tensioning system.
    Date of issuance: January 31, 2001.
    Effective date: January 31, 2001.
    Amendment Nos. 210 and 204.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48750). The October 4, 2000 letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2001.
    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: December 6, 2000.
    Brief description of amendments: The amendments delete Technical 
Specifications (TS) Section 6.8.4.d, ``Post Accident Sampling,'' for 
Turkey Point Units 3 and 4 and thereby eliminate the requirements to 
have and maintain the post-accident sampling system (PASS) for those 
units.
    Date of issuance: January 31, 2001.
    Effective date: January 31, 2001.
    Amendment Nos. 211 and 205.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81923).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

[[Page 11067]]

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of application for amendment: September 30, 2000, as 
supplemented November 22, and December 20, 2000.
    Brief description of amendment: The amendment would allow an 
extension of the steam generator tube inspection surveillance 
requirements of Technical Specification Surveillance Requirement 
4.4.5.3. Specifically, the licensee requested to extend the required 
inspection interval from 40 to 56 calendar months.
    Date of issuance: January 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 232.
    Facility Operating License No. DPR-74: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 ( 65 
FR 
62387).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 1, 2000.
    Brief description of amendment: The amendment approves changes to 
Technical Specifications (TSs) 3.3.3.2, ``Instrumentation, Movable 
Incore Detectors''; 3.3.3.3, ``Instrumentation, Seismic 
Instrumentation''; 3.3.3.4, ``Instrumentation, Meteorological 
Instrumentation''; 3.3.3.8, ``Loose-Part Detection System''; 3.3.4, 
``Turbine Overspeed Protection''; and Index Pages vi and vii. The 
changes relocate the requirements for the incore detectors, seismic 
instrumentation, meteorological instrumentation, loose-part detection 
system, and turbine overspeed protection system from the TSs to the 
Technical Requirements Manual. The Bases for these TSs have been 
modified to reflect the TS changes.
    Date of issuance: January 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 193.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71136).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2000 as supplemented 
January 5, 2001.
    Brief description of amendment: The amendment changes Technical 
Specifications (TSs) 3.8.1.1, ``Electrical Power Systems--A.C. 
Sources--Operating,'' and 3.8.1.2, ``Electrical Power Systems--A.C. 
Sources--Shutdown.'' The changes allow certain EDG surveillance 
requirements to be performed when the plant is operating instead of 
shut down as currently required. The index and Bases for these TSs are 
modified to reflect the changes.
    Date of issuance: February 2, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 194.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 
54087).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: November 18, 1999, as 
supplemented August 7, 2000.
    Brief description of amendment: The amendment to the Kewaunee 
Nuclear Power Plant Technical Specifications approves an increase in 
the allowable number of spent fuel assemblies in the spent fuel pools. 
The addition of the 215 storage locations in the new north canal pool 
will extend the full-core reserve capability until after the 2009 
outage, and increase the total capacity to 1,205 spent fuel assemblies.
    Date of issuance: January 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 150.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1 and December 
21, 2000 (65 FR 65347 and 65 FR 80471 respectively)
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated January 23, 2001.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: May 12, 2000, as supplemented 
by letter dated January 25, 2001.
    Brief description of amendments: These amendments authorize (1) a 
design upgrade of the refueling water purification (RWP) system to 
permit reclassification of this system from Design Class II/non-Seismic 
Category 1 to Design Class I/Seismic Category 1 to allow filtering of 
the refueling water storage tank (RWST) water while the RWST is 
required to be operable, and (2) the use of a temporary reverse osmosis 
skid mounted system to reduce RWST silica concentration levels while 
the RWST is required to be operable following upgrade of the RWP system 
to satisfy reactor coolant chemistry limits.
    Date of issuance: January 29, 2001.
    Effective date: January 29, 2001, and shall be implemented in the 
next periodic update to the FSAR Update, following upgrade of the 
refueling water purification system, in accordance with 10 CFR 
50.71(e).
    Amendment Nos.: Unit 1--144 ; Unit 2-143.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the FSAR Update.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43050).
    The January 25, 2001, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed,

[[Page 11068]]

and did not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 6, 2000 (PCN-274, 
Supplement 1).
    Brief description of amendments: The amendments revised the San 
Onofre, Units 2 and 3 Technical Specification 3.3.11, ``Post Accident 
Monitoring Instrumentation (PAMI),'' to extend the PAMI surveillance 
frequency from 18 to 24 months to accommodate a 24-month fuel cycle.
    Date of issuance: January 30, 2001.
    Effective date: January 30, 2001, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--176; Unit 3-167.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62391).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 6, 2000 (PCN-518).
    Brief description of amendments: The amendments revise TS 3.7.11, 
``Control Room Emergency Air Cleanup System (CREACUS),'' to establish 
actions to be taken for inoperable ventilation systems due to a 
degraded control room pressure boundary. The amendments allow up to 24 
hours to restore the pressure boundary to operable status when two 
ventilation trains are inoperable due to an inoperable pressure 
boundary in Modes 1, 2, 3, and 4. In addition, a limiting condition for 
operation note is added to allow the pressure boundary to be opened 
intermittently under administrative control without affecting CREACUS 
operability.
    Date of issuance: January 30, 2001.
    Effective date: January 30, 2001, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--177; Unit 3--168.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69066).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: November 10, 2000.
    Brief description of amendment: This amendment will allow: (a) the 
minimum fuel oil stored in the fuel oil storage tank (FOST) for each 
emergency diesel generator (EDG) to be raised from 47,100 gallons to 
48,500 gallons for Modes 1-4, and from 33,200 gallons to 42,500 gallons 
for Modes 5 and 6; and (b) the minimum fuel oil maintained in the day 
fuel tank for each EDG to be raised from 300 gallons to 360 gallons for 
Modes 1-6.
    Date of issuance: February 2, 2001.
    Effective date: February 2, 2001.
    Amendment No.: 150.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 20, 2000 (65 
FR 69795).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: October 9, 2000, supplemented December 
4, 2000.
    Brief Description of amendments: The amendments revise Technical 
Specification 5.5.14, ``Technical Specification (TS) Bases Control 
Program,'' to provide consistency with the changes in 10 CFR 50.59 
which were published in the Federal Register on October 4, 2000.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 148 and 140.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2000 (65 
FR 79907) The supplement dated December 4, 2000, provided clarifying 
information that did not change the scope of the October 4, 2000, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 14, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) Surveillance Requirements (SR) 3.8.1.9 
and 3.8.1.14 to reduce diesel generators loading requirements from 
 6800 kW and  7000 kW to  6500 kW and 
 7000 kW. These changes will make the above SRs consistent 
with SRs 3.8.1.3 and 3.8.1.13, which are in the current TSs. In 
addition, the proposed changes would correct a typographical error in 
Section 5.6.7, ``EDG Failure Report,'' in the Vogtle TS. This editorial 
change will correctly reference Regulatory Position C.4 of Regulatory 
Guide 1.9, Revision 3 instead of Regulatory Position C.5.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--117; Unit 2--95.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 54087).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

[[Page 11069]]

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: August 11, 2000 (TS-400) as 
supplemented by letter dated October 20, 2000.
    Brief description of amendments: The amendments revised the 
surveillance test requirements for excess flow check valves.
    Date of issuance: January 29, 2001.
    Effective date: January 29, 2001.
    Amendment Nos.: 268 and 228.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 54088). The October 20, 2000, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 29, 2001.
    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: December 7, 2000 (ET 00-0041).
    Brief description of amendment: The amendment changes Table 3.3.2-
1, ``Engineered Safety Feature Actuation System Instrumentation,'' of 
the Technical Specifications. The change adds Surveillance Requirement 
(SR) 3.3.2.10 for the following two engineered safety feature actuation 
system instrumentation in the table: item 6.f, loss of offsite power, 
and item 6.h, auxiliary feedwater pump suction transfer on suction 
pressure--low.
    Date of issuance: February 06, 2001.
    Effective date: February 06, 2001, and shall be implemented 
including the changes to the Bases for the response times, within 60 
days of the date of issuance.
    Amendment No.: 136.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81932).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 06, 2001.
    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 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, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852, 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 March 23, 2001, 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

[[Page 11070]]

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, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852, 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 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 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, 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 20555-001, 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 Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York
    Date of amendment request: December 19, 2000.
    Description of amendment request: The amendment revises the 
Technical Specifications to indicate that quadrant power tilt limits 
apply only when reactor power is greater than 50 percent.
    Date of issuance: December 20, 2000.
    Effective Date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
December 20, 2000.
    Attorney for licensee: Mr. John M. Fulton, Assistant General 
Counsel Entergy Nuclear Generating Co. Pilgrim Station, 600 Rocky Hill 
Road Plymouth, MA 02360.
    NRC Section Chief: Marsha Gamberoni.

    Dated at Rockville, Maryland, this 14th day of February 2001.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-4228 Filed 2-20-01; 8:45 am]
BILLING CODE 7590-01-U