[Federal Register Volume 65, Number 221 (Wednesday, November 15, 2000)]
[Notices]
[Pages 69058-69071]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-29250]


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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 October 23, 2000, through November 3, 2000. 
The last biweekly notice was published on November 1, 2000.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 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's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By December 15, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, 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). 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

[[Page 69059]]

the issue of law or fact to be raised or controverted. In addition, the 
petitioner shall provide a brief explanation of the bases of the 
contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. Petitioner 
must provide sufficient information to show that a genuine dispute 
exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, 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 20555-0001, 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, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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: November 22, 1999, as supplemented on 
September 11, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Sections 4.5.D, ``Containment Air 
Filtration System (CAFS),'' 4.5.E, ``Control Room Air Filtration System 
(CRAFS),'' 4.5.F, ``Fuel Storage Building Air Filtration System 
(FSBAFS),'' and 4.5.G, ``Post-accident Containment Venting System 
(PACVS),'' to address the testing requirements in Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
    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 in the consequences of an accident 
previously evaluated?
    No. The proposed change would revise Section 4.5 to incorporate 
current NRC [Nuclear Regulatory Commission] testing requirements 
which affect how the charcoal would be tested in the laboratory. 
These changes would 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 
would 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 would implement testing methodology for 
ventilation system charcoal in accordance with Generic Letter 99-02, 
but would not alter equipment performance criteria or standards. The 
safety analysis of the facility would remain complete and accurate, 
and would not be affected by the new charcoal testing requirements. 
There would be no physical changes to the facility and the plant 
conditions for which the design basis accidents have been evaluated 
would still be valid. The operating procedures and emergency 
procedures would be unaffected. Consequently no new failure modes 
would be 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. Since there would be no changes to the operation of the 
facility, to its physical design, or to the performance 
characteristics of any safety-related equipment, neither the Updated 
Final Safety Analysis Report (UFSAR) design basis, accident 
assumptions, nor Technical Specification bases would be affected. 
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 
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.

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

    Date of amendment request: September 5, 2000.
    Description of amendment request: The amendment revises Technical 
Specification 3.3.5.1, 3.3.6.1 and 3.3.6.2. The proposed changes would 
add notes to tables listing instrument channels that are common to, or 
support the operability of interrelated systems as governed by these 
technical specifications.
    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:


[[Page 69060]]


    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 impact on previously analyzed 
accidents or transients and has no affect on design, operation, 
capacity, or surveillance requirements of the affected 
instrumentation channels. The change provides branching notes to the 
Loss of Coolant Accident (LOCA) Time Delay Relay (TDR) Functions of 
LCO [limiting condition of operation] 3.3.5.1 from instrument 
channels of the primary and secondary containment isolation channels 
of LCO 3.3.6.1 and LCO 3.3.6.2 and the associated support features 
for the LOCA TDR function. Since these instruments affect multiple 
LCOs, this change will assure that operators implement the most 
restrictive Action and Completion Time when a channel becomes 
inoperable or is placed in the tripped condition. Providing this 
branching to the more restrictive Actions makes explicit what is 
currently required for Operability and has no impact on any 
previously evaluated accident.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not impact any operational or physical 
aspect of WNP-2. The change only makes explicit the LCOs affected by 
the primary and secondary containment isolation instruments and the 
associated supported features for the LOCA TDR function.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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 provides branching notes to the LOCA TDR 
channels of LCO 3.3.5.1 from instrument channels of the primary and 
secondary containment isolation channels of LCO 3.3.6.1 and LCO 
3.3.6.2 and provides notes for identifying associated support 
features for the LOCA TDR function. This change only makes explicit 
what is currently required for LCO 3.3.5.1 Functions 1c, 1d, 2c and 
2d instrument channel Operability. This change will make explicit 
the most restrictive Action when an instrument sensor or channel 
becomes inoperable or is placed in the tripped condition, thereby, 
maintaining the margin of safety in accordance with the Technical 
Specifications.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 3, 2000.
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 (CR-3) Improved Technical 
Specifications (ITS) 3.7.12, ``Control Room Emergency Ventilation 
System (CREVS),'' ITS 5.6.2.12, ``Ventilation Filter Testing Program 
(VFTP),'' ITS 3.3.16, ``Control Room Isolation--High Radiation,'' and 
ITS 3.7.18, ``Control Complex Cooling System.'' The proposed ITS 
changes are based on the results of revised public and control room 
dose calculations for CR-3 design basis radiological accidents using an 
alternative source term (AST).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
analyzed. The CR-3 Control Room Emergency Ventilation System (CREVS) 
and the Control Complex Habitability Envelope (CCHE) only function 
following the initiation of a design basis radiological accident. 
Therefore, the changes to the CREVS specification, the CREVS filter 
testing criteria, and the deletion of the requirement for control 
room isolation on high radiation proposed by this amendment will not 
increase the probability of any previously analyzed accident. The 
Control Complex Cooling System and Auxiliary Building Ventilation 
System are not initiators of any design basis accident. Therefore, 
the changes to the Control Complex Cooling System specification and 
the changes to the testing guidelines for the Auxiliary Building 
Ventilation System exhaust filters proposed by this amendment will 
not increase the probability of occurrence of any previously 
analyzed accident.
    Revised dose calculations, which take into account the changes 
proposed by this amendment and the use of an AST, have been 
performed for the CR-3 design basis radiological accidents. The 
results of these revised calculations indicate that public and 
control room doses will not exceed the limits specified by 10 CFR 
50.67 and Regulatory Guide 1.183. In addition, a comparison between 
results of the current public dose calculations and the revised 
public dose calculations indicate that the proposed changes will not 
result in a significant increase in predicted dose consequences for 
any of the analyzed accidents. Therefore, the proposed changes do 
not involve a significant increase in the consequences of any 
previously analyzed accident.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    Limiting the requirements for the Control Complex Cooling System 
and CREVS to be operable to Modes 1, 2, 3, and 4, and changing the 
Auxiliary Building Ventilation System exhaust filter testing 
guidelines do not result in changes to the design or operation of 
these systems. Although the other changes proposed by this amendment 
could affect the operation of the CREVS and CCHE following a design 
basis radiological accident, none of these changes can initiate a 
new or different kind of accident since they are only related to 
system capabilities that provide protection from accidents that have 
already occurred. Therefore the proposed changes do not create the 
possibility of a new or different kind of accident from those 
previously analyzed.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed changes to the control complex cooling 
specification do not affect the ability of the system to maintain 
control complex temperatures within safety-related equipment 
operability limits when the equipment is required. The results of 
revised control room dose calculations indicate that the proposed 
changes to the CREVS specification, the CREVS filter testing 
criteria, and removal of the CREVS actuation signal on high 
radiation will not affect the ability of the CREVS and CCHE to 
maintain control room doses less than required limits during design 
basis radiological accidents. The revised dose calculations also 
indicate that the Auxiliary Building Ventilation System exhaust 
filters are not required in order to maintain public or control room 
doses less than required limits; therefore the proposed changes to 
the testing requirements for these filters cannot adversely affect 
public or control room doses.
    Based on the above, the revised technical specifications meet 
the same intent as the currently approved specifications. Therefore, 
the proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042.

[[Page 69061]]

    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: June 7, 1999, as supplemented February 
4, 2000.
    Description of amendment request: The proposed amendment requests 
the staff to evaluate the integrity of the Kewaunee Reactor Pressure 
Vessel (RPV) circumferential beltline weld using a Master Curve-based 
methodology.
    The licensee submitted a request for exemptions to 10 CFR 50.61, 10 
CFR 50 Appendix G, and 10 CFR 50, Appendix H, to allow the use of the 
Master Curve-based methodology for calculating the RPV Reference 
Temperature for Pressurized Thermal Shock (RTPTS) based on 
the fracture toughness data from irradiated pre-cracked Charpy V-notch 
specimen testing of Kewaunee and Maine Yankee surveillance welds. The 
Master Curve methodology is based on American Society for Mechanical 
Engineers (ASME) Code Case N-629 and American Society for Testing and 
Materials Standard (ASTM) E-1921. In its submittals, the licensee also 
requested a revision of the facility's Pressure-Temperature (P/T) limit 
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, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Failure of a reactor vessel is not an accident that has been 
previously evaluated. Design provisions ensure that this is not a 
credible event. Since the potential consequences of a reactor vessel 
failure are so severe, industry and governmental agencies have 
worked together to ensure that failure will not occur. Compliance 
with 10 CFR 50.61, 10 CFR 50 Appendix G and H, and application of 
ASME Code Case N-514, ASME Code Case N-588, and the exemption 
requested in Attachment 1 ensures that failure of a reactor vessel 
will not occur. The proposed changes do not impact the capability of 
the reactor coolant pressure boundary piping (i.e., no change in 
operating pressure, materials, seismic loading, etc.) and therefore 
do not increase the potential for the occurrence of a LOCA.
    The LTOP setpoint, LTOP system enabling temperature, and revised 
P/T limits reflected in proposed Figures TS 3.1-1 and TS 3.1-2 
ensure that the Appendix G pressure/temperature limits are not 
exceeded, and therefore, ensure that RCS integrity is maintained. 
The changes do not modify the reactor coolant system pressure 
boundary, nor make any physical changes to the facility design, 
material, construction standards, or setpoints. The reactor coolant 
system full power operating pressure (2235 psig) is not being 
changed by this proposed amendment. The LTOP valve setpoint remains 
at 500 psig. The LTOP enabling temperature based on 
Figure TS 3.1-2 is 200 deg.F and is consistent with ASME Code Case 
N-514 guidance of RTNDT + 50 deg.F. The LTOP enabling 
temperature is not changed by this amendment. The allowable 
combination of Appendix G pressure and temperature for the cooldown 
limits is marginally greater than the current limits. The 
combination of slightly greater allowable Appendix G pressure and 
temperature limits and low enabling temperature produces an adequate 
operating window. An adequate operating window reduces the 
likelihood of inadvertently lifting the LTOP relief valve while 
maneuvering the plant through the knee of the P-T curve during 
startup and shutdown. The probability of an LTOP event occurring is 
independent of the pressure-temperature limits for the RCS pressure 
boundary and enabling temperature. Therefore, the probability of a 
LTOP event is not increased.
    The revised heatup and cooldown limit curves and corresponding 
LTOP enabling temperature were developed using test results from 
unirradiated and/or irradiated specimens that represent the KNPP 
reactor vessel beltline circumferential weld, closure head flange, 
and intermediate forging. The circumferential beltline weld and 
intermediate forging are the most limiting materials in the reactor 
coolant pressure boundary. These materials are limiting due to the 
effects of neutron irradiation which cause the flow properties to 
increase and the toughness to decrease. The circumferential beltline 
weld is the controlling material for evaluation of pressurized 
thermal shock. With NRC approval to use Code Case N-588 and the 
exemption requested in Attachment 1, the reactor vessel intermediate 
forging and head flange become the limiting and controlling 
materials for development of the Appendix G limit curves and 
corresponding LTOP system enabling temperature. 10 CFR 50, Appendix 
G states that the metal temperature of the closure flange regions 
must exceed the material unirradiated RTNDT by at least 
120 deg.F for normal operation and 90 deg.F for hydrostatic pressure 
tests and leak tests when the pressure exceeds 20 percent of the 
preservice hydrostatic test pressure. Fracture toughness, drop 
weight, and Charpy V-notch testing of the 1P3571 weld metal and drop 
weight, and Charpy V-notch testing of the intermediate forging 
material has been performed. The results of those tests have been 
used for derivation of the revised PTS assessment, the proposed 
Appendix G heatup and cooldown limit curves, and the corresponding 
LTOP system enabling temperature. The revised limit curves and 
corresponding LTOP enabling temperature have been developed using 
accepted engineering practices. The evaluations were performed in 
accordance with methods derived from the ASME Boiler and Pressure 
Vessel Code, criteria set forth in NRC Regulatory Standard Review 
Plan 5.3.2, and 10 CFR 50.61. The revised heatup and cooldown limit 
curves and corresponding LTOP enabling temperature ensures adequate 
fracture toughness for ferritic materials of the pressure-retaining 
components of the reactor coolant pressure boundary. These limit 
curves provide adequate margins of safety during any condition of 
normal operation, including anticipated operational occurrences and 
system hydrostatic tests, and low temperature overpressure 
protection [corresponding to isothermal events during low 
temperature operations (i.e., 200 deg.F)], thus ensuring 
the integrity of the reactor coolant pressure boundary.
    The changes do not adversely affect the integrity of the RCS 
such that its function in the control of radiological consequences 
is affected. Radiological off-site exposures from normal operation 
and operational transients, and faults of moderate frequency do not 
exceed the guidelines of 10 CFR 100. In addition, the changes do not 
affect any fission product barrier. The changes do not degrade or 
prevent the response of the LTOP relief valve or other safety-
related systems to previously evaluated accidents. In addition, the 
changes do not alter any assumption previously made in the 
radiological consequence evaluations nor affect the mitigation of 
the radiological consequences of an accident previously evaluated. 
Therefore, the consequences of an accident previously evaluated will 
not be increased.
    Thus, operation of KNPP in accordance with the PA [proposed 
amendment] does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the potential consequences of a reactor vessel failure are 
so severe, industry and governmental agencies have worked together 
to ensure that failure will not occur. Application of ASME Code Case 
N-514, ASME Code Case N-588, and the exemption requested in 
Attachment 1 ensures that failure of a reactor vessel will not 
occur. Therefore, a failure of the reactor vessel can still be 
considered incredible.
    The proposed heatup and cooldown limit curves have been 
constructed by combining the most conservative pressure-temperature 
limits derived by using material properties of the intermediate 
forging, closure head flange, and beltline circumferential weld to 
form a single set of composite curves. Use of the proposed curves, 
does not modify the reactor coolant system pressure boundary, nor 
make any physical changes to the LTOP setpoint or design. Proposed 
Figures TS 3.1-1 and TS 3.1-2 were prepared in accordance with 
regulatory and code requirements and were derived using conservative 
material property basis and neutron exposure projections thru 33 
EFPY. Therefore, the proposed heatup and cooldown curves and LTOP 
limits will continue to protect the reactor vessel from failure.
    The LTOP system enabling temperature and the proposed Appendix G 
pressure

[[Page 69062]]

temperature limitations were prepared using methods derived from the 
ASME Boiler and Pressure Vessel Code and the criteria set forth in 
NRC Regulatory Standard Review Plan 5.3.2. The changes do not cause 
the initiation of any accident nor create any new credible limiting 
failure for safety-related systems and components. The changes do 
not result in any event previously deemed incredible being made 
credible. As such, it does not create the possibility of an accident 
different than previously evaluated. The changes do not have any 
adverse effect on the ability of the safety-related systems to 
perform their intended safety functions.
    The proposed changes do not make physical changes to the plant 
or create new failure modes. Thus, the PA [proposed amendment] does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed Appendix G pressure temperature limitations and 
corresponding LTOP enabling temperature were prepared using methods 
derived from the ASME Boiler and Pressure Vessel Code, including 
ASME Code Cases N-514 , N-588, and N-629.
    Inherent conservatism in the P/T limits resulting from these 
documents is described in the Safety Evaluation.
    Alternative methodologies to the safety margins required by 
Appendix G to 10 CFR Part 50 have been developed by the ASME Working 
Group on Operating Plant Criteria. Three of these methodologies are 
contained in ASME Code Cases N-514, N-588, and N-629.
    Code Case N-514 provides criteria to determine pressure limits 
during LTOP events that avoid certain unnecessary operational 
restrictions, provide adequate margins against failure of the 
reactor pressure vessel, and reduce the potential for unnecessary 
activation of the relief valve used for LTOP. Specifically, the ASME 
Code Case N-514 allows determination of the setpoint for LTOP events 
such that the maximum pressure in the vessel would not exceed 110% 
of the P/T limits of the existing ASME Appendix G; and redefines the 
enabling temperature at a coolant temperature less than 200  deg.F 
or a reactor vessel metal temperature less than RTNDT + 
50  deg.F, whichever is greater. Code Case N-514, ``Low Temperature 
Overpressure Protection,'' has been approved by the ASME Code 
Committee but not yet approved for use in Regulatory Guides 1.147, 
1.85, or 1.84. The content of this Code Case has been incorporated 
into Appendix G of Section XI of the ASME Code and published in the 
1993 Addenda to Section XI. It is expected that the next revision of 
10 CFR 50.55a will endorse the 1993 Addenda and Appendix G of 
Section XI. Code Case N-514 is not in conflict with 10 CFR 50.61 and 
therefore has been used to establish the LTOP system enabling 
temperature; the provision for exceeding 110% of the Appendix G 
limits has not been incorporated in PA [proposed amendment] 160. The 
NRC previously approved use of Code Case N-514 for determination of 
the LTOP enabling temperature in Reference 6.
    Code Case N-588 provides benefits in terms of calculating 
pressure-temperature limits by revising the Section XI, Appendix G 
reference flaw orientation for circumferential welds in reactor 
vessels. The NRC previously approved use of Code Case N-588 for use 
at KNPP in references 4 and 5.
    In support of this PA [proposed amendment], WPSC used fracture 
toughness results representing the beltline weld metal that were 
irradiated to EOL and in excess of EOLE fluence. The fracture 
toughness results were analyzed as described under Case #6 in WCAP-
15075 and ASME Code Case N-629 for determining the EOL and EOLE 
indexing reference temperature values. Attachment 1 to this letter 
provides information to support NRC approval to use the weld metal 
fracture toughness results along with the methodology presented in 
WCAP-15075 for the KNPP PTS evaluation. The KNPP application of the 
methodology presented in WCAP-15075, identified as Case #6, 
incorporates the following additional margins beyond that 
recommended in ASTM E1921-97:
    (a) A delta value of 17  deg.F is added to T0 to 
ensure that the margin in the KNPP application is at least as 
conservative as the margin associated with the most limiting HSST-02 
plate material.
    (b) An additional margin of 18  deg.F has been added to the 
above 17  deg.F to be consistent with the ASME Code Case N-629, and 
align the KNPP lead plant application with current consensus of the 
technical community regarding the best use of fracture toughness 
based indexing reference temperature data.
    (c) A 2  value of 16  deg.F and 24  deg.F is added to 
account for RTTo measurement uncertainty for EOL and 
EOLE, respectively.
    (d) A value of (+)35  deg.F and (-)32  deg.F accounts for heat 
uncertainty between the KNPP and Maine Yankee surveillance capsule 
specimens for EOL and EOLE, respectively.
    Fracture toughness testing of irradiated 1P3571 weld metal, 
performed in accordance with ASTM E1921-97 and application of ASME 
Code Case N-629 along with the methods in WCAP-15075, indicate that 
the end of life indexing reference temperature is 234  deg.F. This 
fracture toughness generated EOL indexing reference temperature 
value includes a margin of 34  deg.F (18  deg.F + 16  deg.F). The 
fracture toughness generated indexing reference temperature value 
(234  deg.F) is lower than the ART value (277  deg.F) predicted by 
the Charpy V-notch and Drop Weight methodology. Both methodologies 
predict end of life indexing reference temperature values that are 
below the pressurized thermal shock screening criteria (300  deg.F).
    Use of the methodology set forth in the ASME Boiler and Pressure 
Vessel Code, NRC Regulatory Standard Review Plan 5.3.2., WCAP-15075, 
10 CFR 50.61, and 10 CFR 50 Appendices G and H ensures that proper 
limits and safety factors are maintained. Thus, the PA [proposed 
amendment] does not involve a significant reduction in the margin of 
safety.
    The revised heatup and cooldown limit curves and corresponding 
LTOP system enabling temperature were prepared using fracture 
toughness, drop weight and Charpy V-notch data for the beltline weld 
material; drop weight and Charpy V-notch data for the closure head 
flange and intermediated forging material; along with practices 
described herein and methods derived from the ASME Boiler and 
Pressure Vessel Code and 10 CFR 50.61. The safety factors and 
margins used in the development of the limit curves and LTOP system 
enabling temperature meet the criteria set forth by these documents. 
Application of low leakage core designs decreases the rate of shift 
in transition temperature from ductile to nonductile behavior. The 
revised limit curves and corresponding LTOP enabling temperature 
provide adequate margins of safety during any condition of normal 
operation, including anticipated operational occurrences and system 
hydrostatic tests, and low temperature overpressure protection 
[corresponding to isothermal events during low temperature 
operations (i.e., 200  deg.F)]. With the preparation of 
the revised limit curves in accordance with the latest criteria and 
guidance, this proposed amendment ensures that proper limits and 
safety factors are maintained.
    Thus, the proposed amendment does not involve a significant 
reduction in a margin of safety. Therefore, the proposed amendment 
does not represent a significant decrease in the margin of safety. 
As shown in Attachment 1 [in the proposed amendment], a loss of 
reactor vessel integrity is still incredible. Furthermore, the LTOP 
setpoint and enabling temperature will continue to protect the 
reactor coolant system during low temperature operation.

    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.

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

    Date of amendment request: September 7, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point Nuclear Generating Unit No. 3 (IP3) Technical 
Specifications (TSs) would reflect a modification planned for refueling 
outage (RO) 11, scheduled to begin in May of 2001. The modification 
will automatically close, on a safety injection signal, the existing 
main feedwater inlet isolation valves (MFIIVs) and the main feedwater 
low flow bypass inlet isolation valves.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 69063]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    Operation of the Indian Point 3 plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS change reflects a planned modification to 
automatically isolate main feedwater on a safety injection signal 
using the motor operated Main Feedwater Inlet Isolation Valves 
(MFIIVs) and MF [main feedwater] low flow bypass inlet isolation 
valves. These non-safety valves will be incorporated into the IST 
[inservice testing] program as augmented components and included in 
the Generic Letter 89-10 program for motor operated valves. The 
modification will not relocate the safety injection signal from the 
Main Boiler Feedpump Discharge Valves (MBFPDVs) but closure will no 
longer be assumed in analyses. The modification is based on current 
design function for the feedwater isolation following a main steam 
line break inside containment accomplished by MBFPDVs. The TS 
changes add a limiting condition for operation, required action 
statements with completion times and surveillance requirements that 
are the same as those previously approved for Westinghouse plants in 
the Standard Technical Specifications found in NUREG-1432. The plant 
core reload analysis will assume that the modification is complete 
(this eliminates the continued addition of the feedwater between the 
MFIIVS and associated bypass valves and the MBFPDVs) and demonstrate 
that a shutdown margin of 1.3% is acceptable and that no boron 
concentration needs to be assumed in the safety injection lines. The 
proposed changes cannot affect the probability of an accident 
occurring since they reflect a change in plant design consistent 
with current design which is not an accident initiator. The proposed 
changes cannot increase the consequences of postulated accidents 
since they reflect a change in plant design that will mitigate the 
effects of feedwater to a faulted steam generator for a main steam 
line break inside containment and restore past analytical 
assumptions regarding a 1.3% shutdown margin and no boron in the 
safety injection lines.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed TS change reflects a planned modification to 
automatically isolate main feedwater on a safety injection signal 
using the motor operated Main Feedwater Inlet Isolation Valves 
(MFIIVs) and MF low flow bypass inlet isolation valves. These non-
safety valves will be incorporated into the IST program as augmented 
components and included in the Generic Letter 89-10 program for 
motor operated valves. The modification will not relocate the safety 
injection signal from the Main Boiler Feedpump Discharge Valves 
(MBFPDVs) but closure will no longer be assumed in analyses. The 
modification is based on current design function for the feedwater 
isolation following a main steam line break inside containment 
accomplished by MBFPDVs. The TS changes add a limiting condition for 
operation, required action statements with completion times and 
surveillance requirements that are the same as those previously 
approved for Westinghouse plants in the Standard Technical 
Specifications found in NUREG-1432. The proposed TS changes do not 
create the possibility of a new or different type of accident from 
those previously evaluated since they reflect a design change that 
will accomplish the same feedwater isolation function as previously 
done by the MBFPDVs with no change to the manner in which the 
feedwater system operates.
    3. Involve a significant reduction in a margin of safety.
    The proposed TS change reflects a planned modification to 
automatically isolate main feedwater on a safety injection signal 
using the motor operated Main Feedwater Inlet Isolation Valves 
(MFIIVs) and MF low flow bypass inlet isolation valves. These non-
safety valves will be incorporated into the IST program as augmented 
components and included in the Generic Letter 89-10 program for 
motor operated valves. The modification will not relocate the safety 
injection signal from the Main Boiler Feedpump Discharge Valves 
(MBFPDVs) but closure will no longer be assumed in analyses. The 
modification is based on current design function for the feedwater 
isolation following a main steam line break inside containment 
accomplished by MBFPDVs. The TS changes add a limiting condition for 
operation, required action statements with completion times and 
surveillance requirements that are the same as those previously 
approved for Westinghouse plants in the Standard Technical 
Specifications found in NUREG-1432. The plant core reload analysis 
will assume that the modification is complete (this eliminates the 
continued addition of the feedwater between the MFIIVS and 
associated bypass valves and the MBFPDVs) and demonstrate that a 
shutdown margin of 1.3% is acceptable and that no boron 
concentration needs to be assumed in the safety injection lines. The 
proposed TS change cannot involve a significant reduction in the 
margin of safety since it is based upon a modification that will 
restore the margin of safety with respect to feedwater addition, 
shutdown margin and core boration for a main steam line break inside 
containment to the previously analyzed condition. This assumes that 
loading of the valves on the emergency diesel generators will not 
affect the emergency diesel generators margin.

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

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

    Date of amendment request: September 7, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point Nuclear Generating Unit No. 3 (IP3) Technical 
Specifications (TSs) would extend allowed outage times (AOTs) on a one-
time basis, before May 31, 2002, to allow for replacement of the 31 and 
32 station batteries while the plant is on line. The proposed amendment 
also removes an expired footnote regarding repairs to the 32 diesel 
fuel oil tank.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed License amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed AOT extension does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. During the replacement of the existing station 
batteries, a temporary battery will provide the same function as the 
Exide batteries being removed. Even though this temporary battery 
will not meet seismic, seismic interaction or security requirements, 
due to its location on the 53-ft elevation of the Turbine Building, 
it is qualified as safety related in all other respects. The 125 VDC 
EDS [electrical distribution system] is normally supplied by the 
associated 480 VAC bus through a Battery Charger. The essential 
function of 31, 32 and 33 station battery is to supply DC control 
power necessary to start and load the associated EDG [emergency 
diesel generator]. Once the EDGs are on line, the 125 VDC EDS will 
be supplied via the battery charger. However, the station batteries 
have been sized to carry shutdown loads for a period of two hours 
without battery terminal voltage falling below its minimum required 
voltage following a plant trip that includes a loss of all AC power. 
This provides additional assurance that the critical DC loads are 
available in the event of a loss of the battery charger. During the 
10-day AOT, when the temporary battery and the associated battery 
charger are supporting the 125 VDC bus, the ability of that ESF 
[engineered safety feature] DC power panel to mitigate an event/
accident remains unchanged except for its ability to cope with a 
seismic, seismic interaction or security event. However, the 
probability of these

[[Page 69064]]

types of events concurrent with the 10-day AOT is very small. During 
these types of events, one ESF DC power panel may be compromised, 
however IP3 has adequate 125 VDC power available in the form of two 
other ESF train DC power panels to mitigate all DBAs. The postulated 
loss of one ESF DC power panel is bounded by the loss of an entire 
ESF electrical train, a condition which the plant is currently 
evaluated to withstand. Based upon the above, the overall design, 
function and operation of the 125 VDC EDS and equipment has not been 
significantly modified by the proposed changes. The proposed changes 
do not affect accident initiators or precursors, nor do they alter 
the design assumptions for the systems or components used to 
mitigate the consequences of an accident as analyzed in Chapter 14 
of the IP3 USFAR [UFSAR] [updated final safety analysis report], 
except for one of the three trains of DC power. The remaining DC 
power trains can mitigate a DBA [design-basis accident]. Therefore, 
the proposed one-time AOT extension TS amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed License Amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. During the replacement of the existing station batteries, a 
temporary battery will provide the same function as the batteries 
being removed. Even though this temporary battery does not meet all 
design requirements of a seismic, seismic interaction or security 
event it possesses adequate capacity to fulfill the safety related 
requirements of supplying necessary power to the associated 125 VDC 
bus under most conditions. Because the temporary battery will 
perform like the station battery that is currently installed, and 
will be connected and used in the same way as a backup power supply 
to the DC bus, no new electrical or functional failure modes are 
created. The temporary battery will be located in the turbine 
building, which is non-seismic and a non-vital area. The temporary 
battery will not be placed into seismically mounted racks. Thus, a 
seismic failure of this temporary battery is possible. Since the 
temporary battery is located in the turbine building the potential 
for battery failure to initiate an accident is not present. The 
failure of the temporary battery cannot create a different response 
from any previously postulated accident. Due to the location of the 
main turbine-generator in relationship to the temporary battery, it 
is not likely that a turbine missile would strike the battery. 
Likewise, an unmitigated Steam Line Break accident outside the VC 
would be interrupted by successful closure of all MSIVs [main steam 
isolation valves] thereby leaving the battery and the associated DC 
bus intact and available. This MSIV closure would occur before any 
potential steam line break impacting the battery on the Turbine deck 
ensuring necessary DC power to the MSIVs when needed. Also, any 
affects of postulated severe weather on the turbine building have 
been evaluated and do not impede the ability of the remaining DC 
subsystems to perform their intended safety function. The remaining 
125 VDC EDS and its equipment will continue to perform the same 
function and be operated in the same fashion. The proposed changes 
do not introduce any new accident initiators or precursors, or any 
new design assumptions for those systems or components used to 
mitigate the consequences of an accident. Therefore, the possibility 
of a new or different kind of accident from any previously evaluated 
has not been created. Thus, the proposed one-time AOT extension TS 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed License Amendment involve a significant 
reduction in a margin of safety?
    No. During the replacement of the existing station batteries, a 
temporary safety related battery will perform the same function as 
the battery being removed. Even though this battery is not 
seismically mounted, not in a seismically qualified building, nor in 
a vital area of the plant it is qualified as a safety related 
battery in all other respects.
    This battery is virtually identical to the safety related 
station battery that is already installed. It possesses adequate 
capacity to fulfill the requirements of the associated 125 VDC bus. 
The proposed replacement activity will not prevent the plant from 
mitigating a DBA during events that result in the loss of the 
temporary battery. In these cases, the remaining DC power supporting 
the design mitigation capability will be maintained. Due to the 
limited duration of the activity, the very low probability of a 
seismic or other seismic interaction event over this limited AOT 
period and the planned implementing contingency actions, a 
significant reduction in the margin of safety will not result. The 
associated DC bus will always be supplied with both a temporary 
battery and a battery charger at all times. The inherent design 
conservatism of the 125 VDC system and its equipment has not been 
significantly altered; only the degree of redundancy is not fully 
qualified. The 125 VDC EDS and its equipment will continue to be 
operated with the same degree of conservatism. Accordingly, there is 
no significant reduction in the margin of safety.
    Therefore, based upon the above evaluation, the Authority has 
concluded that these changes involve no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, 
New York, New York 10019.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: September 7, 2000.
    Description of amendment request: The proposed amendment to the 
Indian Point Nuclear Generating Unit No. 3 Technical Specifications 
would extend the surveillance frequency from 720 hours to 1440 hours 
for the Fuel Storage Building Emergency Ventilation system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    Response: The proposed license amendment does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated. Extending the surveillance frequency from 720 
hours to 1440 hours for the Fuel Storage Building Emergency Ventilation 
(FSBEV) System charcoal and HEPA [High Efficiency Particulate 
Adsorbers] adsorbers does not involve any modifications to the plant, 
will not require changes to how the plant is operated nor will it 
affect the operation of the plant. Filter systems are not initiators of 
accidents, and therefore extending the filter surveillance frequency 
will not increase the probability of an accident. The way the filters 
perform will not be changed by extending the surveillance frequency. In 
addition, it is reasonable to expect satisfactory filter performance at 
this extended frequency based on past surveillance results. Hence, 
there is no change in the assumptions of an accident. Therefore, this 
change will not increase the consequences of an accident previously 
evaluated.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Extending the surveillance frequency from 720 
hours to 1440 hours for the FSBEV charcoal and HEPA adsorbers does not 
involve any modifications to the plant, will not require changes to how 
the plant is operated nor will it affect the operation of the plant. 
Therefore, extending the surveillance frequency will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: The proposed license amendment does not involve a 
significant reduction in a margin of safety. Extending the surveillance

[[Page 69065]]

frequency from 720 hours to 1440 hours for the FSBEV charcoal and HEPA 
adsorbers does not change the TS required methyl iodine efficiency 
removal requirement of >90% that ensures a safety factor of at least 2. 
This change is acceptable because it is reasonable to expect 
satisfactory filter performance at this extended frequency based on 
past surveillance results, hence it is reasonable to expect that the 
additional 720 hours before testing will not result in the safety 
factor being diminished. Thus, the proposed change would not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: September 26, 2000, as supplemented on 
October 6, 2000.
    Description of amendment request: The proposed change would amend 
the Salem Nuclear Generating Station (Salem) Unit Nos. 1 and 2 
Technical Specifications (TSs) to increase the as-found set point 
tolerance for the Pressurizer Safety Valves (PSV) from 1% 
to 3%; increase the as-found set point tolerance for the 
Main Steam Safety Valves (MSSV) from 1% to 3%; 
change the required action for inoperable MSSVs to require a reduction 
in power based upon the number of inoperable MSSVs, as opposed to the 
current requirement to reduce the Power Range Neutron Flux High trip 
setpoint; and remove specifications and references related to plant 
operation with three Reactor Coolant System loops. The associated TS 
Bases sections will also be amended to reflect the TS changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    Changing the pressurizer and main steam safety relief valve lift 
setpoint tolerance from 1% to 3% does not 
significantly increase the probability of any accident previously 
evaluated. The only events initiated by the opening of these safety 
valves are the accidental depressurization of the Reactor Coolant 
System and accidental depressurization of the Main Steam System. 
These events are a result of an inadvertent lifting of these valves 
and do not depend on the safety valve lift setpoint or tolerance. 
Therefore, the likelihood that either of these events will occur has 
not been increased.
    [Analyses associated with the limiting overpressurization 
transients (Loss of External Electrical Load and/or Turbine Trip, 
and Single Reactor Coolant Pump Locked Rotor) have been performed 
that demonstrate that increasing the Pressurizer Safety Valve and 
Main Steam Safety valve lift setpoint tolerance to 3% 
would result in primary and secondary side pressure responses less 
than the acceptance criteria of 110% of the design pressure. 
Therefore, since the proposed setpoint tolerance increase would not 
adversely impact current accident analysis assumptions, the proposed 
change would not result in an increase in consequences of an 
accident previously evaluated.]
    For operation with inoperable main steam safety valves, changing 
the required action from a reduction of the power range high neutron 
flux trip setpoint to a reduction of the allowable reactor power 
level will not increase the consequences of any accident. With 
inoperable Main Steam Safety Valves, the Loss of External Electrical 
Load and/or Turbine Trip event becomes limiting in terms of 
secondary side pressurization. The high flux trip does not provide 
any mitigation for this event. Other events limiting at power, that 
require the power range trip for mitigation, assume a safety 
analysis trip setpoint of 118% (based on a nominal trip setpoint of 
109%) regardless of the initial power level. Therefore, the proposed 
change does not impact any of the accident analysis assumptions.
    The current Salem licensing basis for the Spurious Activation of 
the Safety Injection System credits operator action to unblock a 
pressurizer Power Operated Relief Valve prior to the water solid 
pressurizer reaching the safety valve lift setpoint. The analyses 
that determined the time at which the safety valve would reach its 
pressure setpoint covered the -3% tolerance. Since this would 
conservatively result in the earliest opening time, there was no 
need to consider the positive side of the tolerance. The results of 
the analyses indicate that the allowable operator action time is 
sufficient, such that water relief occurs through the Power Operated 
Relief Valves and not through the Pressurizer Safety Valves. As such 
the consequences of this event have not changed as a result of the 
proposed change.
    Increasing the Main Steam Safety Valve lift setting tolerance 
may result in increased secondary side backpressure for the 
Auxiliary Feedwater Pumps. However, analyses have demonstrated that 
with the elevated backpressures that could result from increasing 
the Main Steam Safety Valve setpoint upper tolerance to +3%, the 
Auxiliary Feedwater Pumps would still provide [greater than the 
minimum] flow required to mitigate events in which normal feedwater 
is not available, a Loss of Normal Feedwater and a Loss of Offsite 
Power to Station Auxiliaries.
    In terms of radiological consequences, the current design and 
licensing basis analyses that include steaming through the Main 
Steam Safety Valves bound the proposed lift setpoint tolerance 
change.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposal will result in a change in the allowed Pressurizer 
Safety Valve and Main Steam Safety Valve lift setpoint tolerance 
range. No physical changes to these valves or to their nominal lift 
setpoint is required. These valves are assumed to malfunction only 
as the initiator for the accidental depressurization of the Reactor 
Coolant System or Main Steam System. An increased lift setpoint 
tolerance range does not change the assumption of these 
depressurization events nor create a new type of event.
    Requiring a reduction in reactor thermal power in the event of 
inoperable Main Steam Safety Valves is consistent with the analysis 
methodology. Initiation of any Salem UFSAR [Updated Final Safety 
Analysis Report] analyzed event at a power level less than full 
power is bounded by those events analyzed at full power, or 
specifically analyzed at the limiting power level, and does not 
constitute a new or different kind of accident. Also, no changes are 
being made to the power range high flux trip setpoint that will make 
it inconsistent with any analytical assumption.
    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 changes do not involve a significant reduction 
in the margin of safety.
    Analyses performed demonstrate that the proposed increase in the 
Pressurizer Safety Valve and Main Steam Safety Valve lift pressure 
setpoint tolerance from 1% to 3% will 
provide acceptable primary and secondary side pressure responses to 
the anticipated operational occurrences and design basis accidents. 
The limiting overpressurization transients, Loss of External 
Electrical Load and/or Turbine Trip, and Single Reactor Coolant Pump 
Locked Rotor, stay well within the acceptance criteria of 110% of 
the design pressure.
    For operation with inoperable Main Steam Safety Valves, 
requiring a reduction in reactor thermal power is consistent with 
the accident analysis. The current requirement to reduce the power 
range high neutron flux trip setpoint [does not reduce the] margin 
of safety since this trip does not provide any mitigation for the 
limiting secondary system pressurization event, Loss of External 
Electrical Load and/or Turbine Trip with inoperable Main Steam 
Safety Valves.

[[Page 69066]]

    The current licensing basis for the Spurious Activation of the 
Safety Injection System credits operator action to unblock a 
pressurizer Power Operated Relief Valve prior to the water solid 
pressurizer reaching the Pressurizer Safety Valve lift setpoint. As 
the Pressurizer Safety Valves are not designed for water relief, 
failure to unblock a Power Operated Relief Valve before reaching the 
Pressurizer Safety Valve lift setpoint would result in water relief 
and likely failure of the Pressurizer Safety Valve to reseat. This 
condition would escalate the Spurious Activation of the Safety 
Injection System (Condition II event) into a small break Loss Of 
Coolant Accident (Condition III event). The analyses that determined 
the time at which primary system pressure would reach the 
Pressurizer Safety Valve setpoint bound the -3% tolerance. The 
results of the analyses indicate that the allowable operator action 
time is sufficient, such that water relief occurs through the Power 
Operated Relief Valves and not through the Pressurizer Safety 
Valves. Since the Pressurizer Safety Valve would not fail due to 
water relief, there is no reduction in the margin of safety for this 
event.
    Increasing the Main Steam Safety Valve lift setting tolerance 
may result in increased secondary side backpressure for the 
Auxiliary Feedwater System. However, analyses have demonstrated that 
under degraded Auxiliary Feedwater Pump performance, and with 
secondary side backpressure corresponding to 103% of the lowest Main 
Steam Safety Valve setpoint, the Auxiliary Feedwater System can 
provide [greater than the minimum] flow required to mitigate those 
events where normal feedwater is not available, a Loss of Normal 
Feedwater and a Loss of Offsite Power to Station Auxiliaries.
    Therefore the proposed changes to the Technical Specifications 
do not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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: October 6, 2000 (PCN-518).
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear Generating Station, Units 2 
and 3, Technical Specification (TS) 3.7.11, ``Control Room Emergency 
Air Cleanup System (CREACUS)'' consistent with generic industry changes 
recently approved by the U.S. Nuclear Regulatory Commission (NRC) 
document Technical Specification Task Force (TSTF)-287. The proposed 
amendments would allow up to 24 hours to restore the Control Room 
Pressure Boundary (CRPB) to operable status when two CREACUS trains are 
inoperable due to an inoperable CRPB in MODE 1, 2, 3, or 4. In 
addition, a Limiting Condition for Operation note would be added to 
allow intermittent opening of the CRPB under administrative controls 
without entering the Actions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the facility in accordance with the proposed 
amendments does not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The Control Room Area Ventilation System and Control Room 
boundary are not assumed to be an initiator of any analyzed 
accident; they are provided to minimize doses to the control room 
operators during an accident. Therefore, these proposed changes have 
no impact on the probability of occurrence of any previously 
analyzed accident.
    The proposed changes also have no impact on offsite dose 
consequences. The control room ventilation system and control room 
boundary provide protection for control room personnel and do not 
mitigate radiological effluents released offsite. With the control 
room boundary inoperable and not pressurized, the accident analyses 
assume unfiltered air would enter the control room and operator 
doses would be significantly increased. Conservative accident 
analysis assumptions do not take credit for available compensatory 
measures to mitigate operator dose. Compensatory measures include 
the supply of protective clothing, and self contained breathing 
apparatus adequate for at least nine persons within the control room 
envelope.
    Additionally, for cases where the control room boundary is 
opened under administrative control, appropriate administrative 
measures ensure the boundary can be rapidly restored. Based on the 
compensatory measures available to the control room operator to 
minimize dose (to be consistent with the intent of General Design 
Criterion 19), the administrative controls required to rapidly 
restore an opened boundary, and considering the low probability of 
an event occurring in this short time period, the consequences are 
not considered to be significantly increased. Operators maintain the 
ability to mitigate a design basis event.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    No changes are being made to actual plant hardware which will 
result in any new accident causal mechanisms. Therefore, no new 
accident causal mechanisms will be generated.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    Response: No.
    Margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be degraded by the 
proposed changes. The Control Room Ventilation System and control 
room boundary provide a protected environment for the control room 
operators during analyzed events. The proposed change would allow 
the boundary to be degraded for a limited period of time. However, 
administrative controls would be in place to rapidly restore an 
opened boundary and existing compensatory measures (e.g., protective 
clothing and self contained breathing apparatus) would be 
implemented to minimize operator dose. Therefore, it is expected 
that operators would maintain the ability to mitigate design basis 
events and none of the fission product barriers would be affected by 
this change.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: October 6, 2000.
    Description of amendment request: The proposed amendment would 
amend each of the three units' Technical Specifications (TS) to adopt 
Technical Specifications Task Force (TSTF) change No. 318, Revision 0 
(TSTF-318). TSTF-318 provides a 7-day action period and completion time 
in the event

[[Page 69067]]

of inoperability of one of the two low pressure coolant injection 
(LPCI) pumps in each of the two emergency core cooling system (ECCS) 
divisions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed new Condition of one LPCI pump in each LPCI 
injection subsystem being inoperable is more reliable than the 
current Limiting Condition for Operation which allows 2 LPCI pumps 
in one ECCS subsystem to be inoperable for 7 days. Also, the LPCI 
mode of the Residual Heat Removal system is not assumed to be 
initiator of any analyzed event. Therefore, the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment or require any existing equipment 
to be operated in a manner different from the present design. The 
proposed change will not impose any new or eliminate any existing 
requirements. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses assumptions. The proposed 
new Condition for one LPCI pump in each LPCI injection subsystem 
represents a more reliable configuration than the existing LCO which 
allows two LPCI pumps in one ECCS subsystem to be inoperable for 7 
days. For these reasons, the proposed amendment 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET I0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: June 16, 2000, as supplemented by letter 
dated September 27, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7 and TS Tables 3.7-1, 3.7-2, 
3.7-3, and 4.1-1. The proposed changes would: (a) revise the 
surveillance frequency for Reactor Protection System and Engineered 
Safety Features Actuation System analog channels from monthly to 
quarterly; (b) decrease the frequency for most permissives to a 
refueling interval; (c) increase the time allowed to perform 
maintenance on an inoperable instrument channel; and (d) revise 
associated action statements consistent with NUREG-1431.
    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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed Reactor 
Protection System (RPS) and Engineered Safety Features Actuation 
System (ESFAS) Technical Specification changes for the Surry Units 1 
and 2 and determined that a significant hazards consideration is not 
involved. In support of this conclusion, the following evaluation is 
provided.
    Criterion 1--Operation of Surry Units 1 and 2 in accordance with 
the proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The determination that the results of the proposed changes 
remain within acceptable criteria was established in the SER(s) 
[Safety Evaluation Report(s)] prepared for WCAP-10271, WCAP-10271 
Supplement 1, WCAP-10271 Supplement 2, WCAP-10271 Supplement 2, 
Revision 1 and WCAP-14333 issued by letters dated February 21, 1985, 
February 22, 1989, April 30, 1998, and July 15, 1998.
    Implementation of the proposed changes is expected to result in 
an increase in total RPS and ESFAS yearly unavailability. The 
proposed changes have been shown to result in a small increase in 
the core damage frequency (CDF) due to the combined effects of 
increased RPS and ESFAS unavailability and reduced inadvertent 
reactor trips.
    The values determined by the WOG [Westinghouse Owners Group] and 
presented in the WCAP for the increase in CDF were verified by 
Brookhaven National Laboratory (BNL) as part of an audit and 
sensitivity analyses for the NRC Staff. Based on the small value of 
the increase compared to the range of uncertainty in the CDF, the 
increase is considered acceptable. The analysis performed by the WOG 
and presented in the WCAP included changes to the surveillance 
frequencies for the automatic actuation logic and actuation relays 
and the reactor trip and bypass breakers. The overall increase in 
the CDF, including the changes to the surveillance frequencies for 
the automatic actuation logic and actuation relays and the reactor 
trip and bypass breakers, was approximately 6 percent. However, even 
with this increase, the overall CDF remains lower than the NRC 
safety goal of 10 E-4/reactor year.
    Changes to surveillance test frequencies for the RPS and ESFAS 
interlocks do not represent a significant reduction in testing. The 
currently specified test interval for interlock channels allows the 
surveillance requirement to be satisfied by verifying that the 
permissive logic is in its required state using the annunciator 
status light. The surveillance as currently required only verifies 
the status of the permissive logic and does not address verification 
of channel setpoint or operability. The setpoint verification and 
channel operability is verified after a refueling shutdown. The 
definition of the channel check includes comparison of the channel 
status with other channels for the same parameter. The requirement 
to routinely verify permissive status is a different consideration 
than the availability of trip or actuation channels which are 
required to change state on the occurrence of an event and for which 
the function availability is more dependent on the surveillance 
interval. Therefore, the change in the interlock surveillance 
requirement to at least once every 18 months does not represent a 
significant change in channel surveillance and does not involve a 
significant increase in unavailability of the RPS and ESFAS.
    For the additional relaxations in WCAP-14333, the WOG evaluated 
the impact of the additional relaxation of allowed outage times and 
completion times, and action statements on core damage frequency. 
The change in core damage frequency is 3.1 percent for those plants 
with two out of three logic schemes that have not implemented the 
proposed surveillance test interval, allowed outage times, and 
completion times evaluated in WCAP-10271 and its supplements. This 
analysis calculates a significantly lower increase in core damage 
frequency than the WCAP-10271 analysis calculated. This can be 
attributed to more realistic maintenance intervals used in the 
current analysis and crediting the AMSAC [ATWS (anticipated 
transient without scram) mitigating system actuation circuitry] 
system as an alternative method of initiating the auxiliary 
feedwater pumps. Therefore, the overall increase in CDF is estimated 
to be 3.1% for the proposed changes per the generic Westinghouse 
analysis.
    The NRC performed an independent evaluation of the impact on 
core damage frequency (CDF) and large early release fraction (LERF). 
The results of the staff's review indicate that the increase in core 
damage frequency is small (approximately 3.2%) and the large early 
release fraction would increase by only 4 percent for 2 out of 3 
logic schemes that have not

[[Page 69068]]

implemented the proposed surveillance test interval, allowed outage 
times, and completion times evaluated in WCAP-10271 and its 
supplements. Further, the absolute values for CDF still remain 
within NRC safety goals.
    Therefore, the proposed changes do not result in a significant 
increase in the severity or consequences of an accident previously 
evaluated. Implementation of the proposed changes affects the 
probability of failure of the RPS and ESFAS but does not alter the 
manner in which protection is afforded or the manner in which 
limiting criteria are established.
    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 changes do not result in a change in the manner in 
which the RPS or ESFAS provide plant protection. No change is being 
made which alters the functioning of the RPS or ESFAS (other than in 
a test mode). Rather the likelihood or probability of the RPS or 
ESFAS functioning properly is affected as described above. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident as defined in the Safety Analysis 
Report.
    The proposed changes do not involve hardware changes. Some 
existing instrumentation is designed to be tested in bypass and 
current Technical Specifications allow testing in bypass. Testing in 
bypass is also recognized by IEEE [Institute of Electrical and 
Electronics Engineers] Standards. Therefore, testing in bypass has 
been previously approved and implementation of the proposed changes 
for testing in bypass does not create the possibility of a new or 
different kind of accident from any previously evaluated. 
Furthermore, since the other proposed changes do not alter the 
physical operation or functioning of the RPS or ESFAS, the 
possibility of a new or different kind of accident from any 
previously evaluated has not been created.
    Criterion 3--The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed changes do not alter the safety limits, limiting 
safety system setpoints or limiting conditions for operation. The 
RPS and ESFAS analog instrumentation remain operable to mitigate as 
assumed in the accident analysis. The impact of reduced testing 
other than as addressed above is to allow a longer time interval 
over which instrument uncertainties (e.g., drift) may act.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety by less frequent testing of the RPS 
and ESFAS analog instruments and will result in less inadvertent 
reactor trips and actuation of Engineered Safety Features 
components.
    This analysis demonstrates that the proposed amendment to the 
Surry Units 1 and 2 Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch.

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 electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

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

    Date of application for amendment: August 25, 2000, as supplemented 
September 21, October 14, and October 25, 2000.
    Brief description of amendment: The amendment revises the reactor 
vessel pressure-temperature limits.
    Date of issuance: October 31, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 134.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2000 (65 
FR 56598).
    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 October 31, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 27, 2000, as supplemented 
October 5, 2000.
    Brief description of amendment: The amendment revises the Safety 
Limit Minimum Critical Power Ratio.
    Date of issuance: November 3, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 135.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51348).
    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 November 3, 2000.
    No significant hazards consideration comments received: No.

[[Page 69069]]

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster 
Creek Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: March 7, as supplemented on 
April 21, June 14, and September 15, 2000.
    Brief description of amendment: The proposed amendment revised the 
Technical Specifications to revise the surveillance requirements from 
once per refueling interval for each excess flow check valve (EFCV) to 
testing a representative sample of EFCVs once per 24 months.
    Date of Issuance: October 25, 2000.
    Effective date: October 25, 2000 and shall be implemented within 30 
days of issuance.
    Amendment No.: 216.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51354).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 25, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 7, 2000, as supplemented 
June 14 and September 11, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.7.6, ``Control Room Emergency Filtration 
System,'' TS 3/4.7.7, ``Reactor Auxiliary Building Emergency Exhaust 
System,'' and TS 3/4.9.12, ``Fuel Handling Building Emergency Exhaust 
System.'' Specifically, these TS have been revised to provide an action 
when the Control Room Emergency Filtration System or Reactor Auxiliary 
Building Emergency Exhaust System ventilation boundary is inoperable, 
and a note that allows an applicable ventilation boundary to be open 
intermittently under administrative controls. The associated TS Bases 
are also being changed in accordance with the amendment. In addition, 
TS 3/4.3.3.1, ``Radiation Monitoring for Plant Operations,'' has been 
modified to provide consistency between the applicability of the 
Control Room Emergency Filtration System and the radiation monitors 
that initiate a control room isolation signal.
    Date of issuance: October 30, 2000.
    Effective date: October 30, 2000.
    Amendment No.: 102.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25762). The supplemental letters dated June 14 and September 11, 2000, 
contained clarifying information only, did not expand the application 
beyond the scope of the initial notice, and 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 October 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 21, 2000.
    Brief description of amendments: The amendments revised the 
condensate storage tank (CST) low-level setpoint to prevent entrainment 
of air in the high pressure coolant injection (HPCI) pump suction line 
when taking suction from the CST. The amendments also revised the 
surveillance requirements for the CST level instruments.
    Date of issuance: October 31, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 182 and 177.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15376).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2000.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: November 23, 1999, as 
supplemented by letter dated September 6, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.5.11--Ventilation Filter Testing Program, 
which provides the test requirements for charcoal filters, to assure 
compliance with the requirements of American Society for Testing and 
Materials D3803-1989.
    Date of issuance: November 2, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 196/177.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15377).
    The supplement dated September 6, 2000, provided clarifying 
information that did not change the scope of the November 23, 1999, 
application and initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 2000.
    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: August 10, 2000.
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow an alternate storage configuration of fuel 
assemblies adjacent to the walls within Region I of the spent fuel 
pool.
    Date of issuance: October 24, 2000.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 224.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 54086).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 7, 2000, as supplemented 
October 2 and 4, 2000.
    Brief description of amendments: The pressure-temperature limits 
specified in Technical Specification (TS) 3.4.9.1 and Figures 3.4-2, 
3.4-3 have been modified, Figure 3.4-4 deleted, and the Cold 
Overpressure Mitigation System (COMS) requirements have been changed. 
The COMS is the Westinghouse version of the Low Temperature 
Overpressure Protection System.
    Date of issuance: October 30, 2000.

[[Page 69070]]

    Effective date: October 30, 2000.
    Amendment Nos.: 208 and 202.
    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 
48751). The supplemental information provided on October 2 and 4, 2000, 
provided clarifying information only and did not affect the proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 2000.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: September 1, 2000.
    Brief description of amendments: The amendments clarify Technical 
Specification 3/4.4.4, ``Pressurizer,'' to reflect the current power 
supply to the pressurizer heaters and require two operable trains of 
pressurizer heaters during Modes 1, 2, and 3. In addition, the 
amendments revise the Bases for Technical Specification 3/4.4.4 to 
reflect these changes and clarify the purpose of the pressurizer 
heaters.
    Date of issuance: October 20, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 246 and 227.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56952).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 2000.
    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: April 19, 2000, as supplemented 
on August 31, 2000.
    Description of amendment request: The amendment implements a 
performance-based Containment Leakage Testing Program in accordance 
with 10 CFR Part 50, Appendix J, Option B as a substitute for the 
requirements of 10 CFR Part 50, Appendix J, Option A. The use of this 
option requires the implementation of a program based on Regulatory 
Guide 1.163, ``Performance-Based Containment Leak-Test Program,'' and 
modification of the Technical Specifications to reflect this program.
    Date of issuance: November 2, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 186.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51359).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 2000.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of application for amendment: December 31, 1999, as 
supplemented by letters dated January 18, July 7, September 22, and 29, 
and October 12, 2000.
    Brief description of amendment: The amendment revises Section 
2.C.(1) of Facility Operating License No. DPR-80 to authorize operation 
of Unit 1 at reactor core power levels not in excess of 3411 megawatts 
thermal (100 percent rated power). Unit 2 is already authorized to 
operate at that power level. This amendment also revises several 
sections within the Improved TS to reflect the increase in reactor 
power level.
    Date of issuance: October 26, 2000.
    Effective date: October 26, 2000.
    Amendment No.: Unit 1--143.
    Facility Operating License No. DPR-80: The amendments revised the 
Technical Specifications and operating license.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21037).
    The January 18, July 7, September 22, and 29, and October 12, 2000, 
supplemental letters provided additional clarifying information, did 
not expand the scope of the application as originally noticed, and did 
not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 26, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 27, 2000, as supplemented 
August 16, 2000, and September 29, 2000.
    Brief description of amendment: The amendment provides for the 
applicability of the current safety limit minimum critical power ratio 
(SLMCPR), TS Section 1.1.A, to cycles beyond Cycle 14. The change also 
updates the approved version of the topical report in TS Section 
6.9.A.4.b.1.
    Date of issuance: October 30, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 266.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51362).
    The August 16 and September 29, 2000, letters 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 October 30, 2000.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: September 8, 2000, as supplemented on 
October 2, 2000.
    Brief Description of amendment: The amendment revises surveillance 
requirements 3.4.11.1 and 3.4.11.4 to eliminate the requirement to 
cycle the Unit 2 pressurizer power-operated relief valve block valves 
during the remainder of operating cycle 14.
    Date of issuance: October 25, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: Unit 2--139.
    Facility Operating License No. NPF-8: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 10, 2000 (65 FR 
60223).
    Public comments requested as to proposed no significant hazards 
consideration: Yes.
    The notice provided an opportunity to submit comments on the 
Commission's

[[Page 69071]]

proposed no significant hazards consideration determination. No 
comments have been received. The notice also provided for an 
opportunity to request a hearing by November 9, 2000, but indicated 
that if the Commission makes a final no significant hazards 
consideration determination, any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendment, finding of exigent circumstances, and a final no significant 
hazards consideration determination are contained in a Safety 
Evaluation dated October 25, 2000.

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

    Date of application for amendments: November 24, 1999 (TS 99-16).
    Brief description of amendments: These amendments revised the 
Technical Specifications (TSs) to update the industry standard that is 
used to test the charcoal adsorber efficiency in safety-related 
ventilation systems.
    Date of issuance: November 2, 2000.
    Effective date: November 2, 2000.
    Amendment Nos.: 263 and 254.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1929). The September 21, 2000, supplement provided clarifying 
information that did not change the scope of the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 14, 2000, as 
supplemented on September 22, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to clarify the valve isolation signal information 
in the TS Table 4.7.2 and make an administrative change to the Table 
main steam isolation valves component identification.
    Date of Issuance: October 31, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 194.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 2000 (65 
FR 58111).
    The September 22, 2000, supplemental letter was within the scope of 
the original application and did not change the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 31, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of November 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-29250 Filed 11-14-00; 8:45 am]
BILLING CODE 7590-01-P