[Federal Register Volume 67, Number 161 (Tuesday, August 20, 2002)]
[Notices]
[Pages 53983-53998]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-20843]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from July 26, 2002 through August 8, 2002. The 
last biweekly notice was published on August 6, 2002 (67 FR 50947).

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; (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated; or (3) involve a significant reduction in a margin of 
safety. The basis for this proposed determination for each amendment 
request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By September 19, 2002, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714,\1\ 
which is available at the Commission's PDR, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
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    \1\ The most recent version of Title 10 of the CODE OF FEDERAL 
REGULATIONS, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''

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[[Page 53984]]

    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendments request: July 24, 2002.
    Description of amendments request: The proposed amendments would 
revise Technical Specifications (TS) Section 3.1.7, ``Standby Liquid 
Control (SLC) System,'' to reflect modifications being made to the 
system as a result of transition to the GE14 fuel design. To support 
this transition, the required in-vessel boron concentration, supplied 
by the SLC system, would be increased from 660 ppm natural boron to a 
concentration equivalent to 720 ppm natural boron. This would be 
accomplished by use of sodium pentaborate solution enriched with the 
Boron-10 isotope. As a result, (1) a new Surveillance Requirement (SR) 
3.1.7.8 would be added to verify sodium pentaborate enrichment, (2) the 
minimum sodium pentaborate concentration value would be lowered in TS 
Figure 3.1.7-1, ``Sodium Pentaborate Solution Volume Versus 
Concentration Requirements,'' and (3) the temperature versus 
concentration requirements of TS Figure 3.1.7-2, ``Sodium Pentaborate 
Solution Temperature Versus Concentration Requirements,'' would be 
revised. In a related change, SR 3.1.7.3 would also be revised. 
Currently, the SR verifies temperature of the SLC pump suction piping. 
The SR would be revised to verify temperature of the suction and 
discharge piping up to the SLC injection valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 53985]]

consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendments do not alter the design or operation of 
the Standby Liquid Control (SLC) system, but rather revise Technical 
Specification (TS) Section 3.1.7 requirements to ensure acceptable 
SLC boron solution volume and concentration values to produce a 
minimum in-vessel boron concentration which is sufficient to bring 
the reactor to a subcritical condition without taking credit for 
control rod movement. The existing design of the SLC system is 
sufficient to handle enriched sodium pentaborate solution, which is 
chemically and physically similar to the current solution. The SLC 
system is not considered to be an initiator of any analyzed event. 
Therefore, the proposed amendments do not increase the probability 
of a previously evaluated accident.
    The current TS Section 3.1.7 requirements ensure acceptable SLC 
boron solution volume and concentration values to produce a minimum 
in-vessel natural boron concentration of 660 ppm. The proposed 
change revises the boron solution requirements of TS Figures 3.1.7-1 
and 3.1.7-2, to ensure a minimum in-vessel concentration equivalent 
to 720 ppm natural boron. A minimum concentration equivalent to 720 
ppm natural boron in the reactor is sufficient to bring the reactor, 
at any time in a fuel cycle, from full power and minimum control rod 
inventory to a subcritical condition with the reactor in the most 
reactive, xenon free state without taking credit for control rod 
movement. This concentration was determined by General Electric 
using the approved methods described in Revision 14 of General 
Electric Standard Application for Reactor Fuel (GESTAR II), NEDE 
24011-P-A. The analysis assumes Brunswick Steam Electric Plant 
(BSEP) operation with an equilibrium core of GE14 fuel, operating at 
2923 megawatts thermal (MWt) with 24 month operating cycles.
    As stated above, the in-vessel boron concentration is being 
raised from 660 ppm natural to 720 ppm equivalent. This will be 
accomplished by use of sodium pentaborate solution enriched with the 
Boron-10 isotope. As a result, a new Surveillance Requirement (SR) 
3.1.7.8 is added. This SR verifies sodium pentaborate enrichment is 
greater than or equal to 47 atom percent Boron-10 prior to addition 
to the SLC tank, thereby ensuring a minimum concentration equivalent 
to 720 ppm natural boron in the reactor will be achieved.
    Use of sodium pentaborate enriched to 47 atom percent Boron-10 
allows the volume versus concentration requirements of TS Figure 
3.1.7-1 to be lowered. This, in turn, lowers the solution's 
saturation temperature. Accordingly, the temperature versus 
concentration requirements of TS Figure 3.1.7-2 are revised. The 
existing 5 deg. F margin to the saturation temperature specified in 
the bases is maintained in the revised TS Figure 3.1.7-2.
    The concentration requirements of the SLC system boron solution 
will ensure that the SLC system continues to comply with the 
requirements of 10 CFR 50.62(c)(4).
    The SLC system is also used to maintain suppression pool pH 
level above 7 following a loss-of-coolant-accident (LOCA) involving 
significant fission product releases. This ensures that iodine will 
be retained in the suppression pool water post-LOCA. The revised 
sodium pentaborate solution requirements were evaluated using the 
methodology provided in NUREG-1465, ``Accident Source Terms for 
Light-Water Nuclear Power Plants, Final Report,'' dated February 1, 
1995 and NUREG/CR-5950, ``Iodine Evolution and pH Control,'' dated 
December 1992. This evaluation demonstrated that the SLC system 
continues to meet its post-LOCA suppression pool pH control design 
function.
    The change to SR 3.1.7.3 is conservative in nature and is 
consistent with both the current Bases for SR 3.1.7.3 and plant 
operating practice. Required verification of the discharge as well 
as suction piping temperature provides additional assurance of 
system operability.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendments do not alter the design or operation of 
the SLC system, but rather revise TS Section 3.1.7 requirements to 
ensure acceptable SLC boron solution volume and concentration values 
to produce a minimum in-vessel boron concentration which is 
sufficient to bring the reactor to a subcritical condition without 
taking credit for control rod movement. The existing design of the 
SLC system is sufficient to handle enriched sodium pentaborate 
solution, which is chemically and physically similar to the current 
solution. Using the enriched solution does not change any of the key 
SLC system process parameters (i.e., flow rates, discharge pressure, 
required net positive suction head, etc.). Correct enrichment is 
ensured by the addition of a new SR to verify sodium pentaborate 
enrichment prior to addition to the SLC tank. The existing 5 deg. F 
margin to the saturation temperature specified in the bases is 
maintained. The change to SR 3.1.7.3 is conservative in nature and 
is consistent with both the current Bases for SR 3.1.7.3 and plant 
operating practice. Required verification of the discharge as well 
as suction piping temperature provides additional assurance of 
system operability. Therefore, the proposed amendments cannot create 
a new or different kind of accident from any accident previously 
evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the boron solution requirements of 
TS Figures 3.1.7-1 and 3.1.7-2, to ensure a minimum in-vessel 
concentration equivalent to 720 ppm natural boron. A minimum 
concentration equivalent to 720 ppm natural boron in the reactor is 
sufficient to bring the reactor, at any time in a fuel cycle, from 
full power and minimum control rod inventory to a subcritical 
condition with the reactor in the most reactive, xenon free state 
without taking credit for control rod movement. This concentration 
was determined by General Electric using the approved methods 
described in GESTAR II. The existing design of the SLC system is 
sufficient to handle enriched sodium pentaborate solution, which is 
chemically and physically similar to the current solution. Correct 
enrichment is ensured by the addition of a new SR 3.1.7.8 to verify 
sodium pentaborate enrichment prior to addition to the SLC tank. The 
existing 5 deg. F margin to the saturation temperature specified in 
the bases is maintained. The existing SLC system design requires 
that SLC inject a quantity of boron that includes an additional 25% 
above that needed for an in-vessel boron concentration of 660 ppm. 
This additional 25% is injected to compensate for imperfect mixing, 
leakage, and volume in other small piping connected to the reactor. 
This margin will be maintained such that an additional 25% above 
that needed for an in-vessel boron concentration equivalent to 720 
ppm natural boron will also be injected. The minimum sodium 
pentaborate concentration of 8.5 weight percent, proposed by this 
amendment request, ensures that the SLC system continues to meet its 
post-LOCA suppression pool pH control design function. The change to 
SR 3.1.7.3 is conservative in nature and is consistent with both the 
current Bases for SR 3.1.7.3 and plant operating practice. Required 
verification of the discharge as well as suction piping temperature 
provides additional assurance of system operability.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: July 16, 2002.
    Description of amendment request: Energy Northwest is requesting 
changes

[[Page 53986]]

to the technical specifications (TS) to change the specified minimum 
emergency diesel generator (DG) steady state output voltage from 3740 
volts to 3910 volts.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed license amendment is administrative and does not 
involve any design changes or physical changes to plant equipment. 
The ability of the DGs to perform their safety functions to mitigate 
consequences is not affected and will continue to be demonstrated in 
the same manner. Therefore the proposed amendment will not affect 
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 license amendment does not involve changes to plant 
equipment and the DGs will continue to perform to their required 
parameters in the same manner. Because the performance of the DGs 
will remain unchanged, this proposed amendment does not present 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 amendment is solely a request to revise the 
Technical Specifications regarding minimum steady state DG output 
voltage requirements. This change would not affect any operating 
parameter or equipment performance. Because this proposed amendment 
would not affect operation, the margin of safety maintained by 
Columbia would remain unchanged.

    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, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: July 8, 2002.
    Description of amendment request: The proposed amendments would 
change Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change would 
add two footnotes to TS Table 3.3.8.1-1, ``Loss of Power 
Instrumentation,'' Functions 1.e and 2.e, ``Degraded Voltage--Time 
Delay, LOCA,'' and makes an editorial change to the heading of TS Table 
3.3.8.1-1. The Degraded Voltage--Time Delay, LOCA, function is 
currently required to be OPERABLE during plant configurations when the 
emergency core cooling system (ECCS) instrumentation that generates the 
loss-of-coolant accident (LOCA) signal is not required to be OPERABLE. 
The proposed changes correct this inconsistency by adding two new 
footnotes to TS Table 3.3.8.1-1 that modify the required OPERABILITY of 
the Degraded Voltage--Time Delay, LOCA, function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The TS Table 3.3.8.1-1 Function column heading change to add the 
reference to the Opposite Unit Division 2 is an editorial change. It 
was always the intent and practice of LaSalle County Station to 
apply TS requirements from this column to the Opposite Unit Division 
2 4.16 kV emergency bus.
    The operation of the Degraded Voltage--Time Delay, LOCA, 
function is not a precursor to any accident previously evaluated. 
Thus, the proposed changes to modify the OPERABILITY of the Degraded 
Voltage--Time Delay, LOCA, function to be consistent with the 
OPERABILITY of the ECCS instrumentation that generate the timer 
initiating LOCA signal do not have any affect on the probability of 
an accident previously evaluated.
    Successful operation of the required safety functions of the 
ECCS is dependent upon the availability of adequate power sources 
for energizing the various components such as pump motors, motor 
operated valves, and the associated control components. Offsite 
power is the preferred source of power for the 4.16 kV emergency 
buses. The Degraded Voltage--Time Delay, LOCA, function does provide 
assurance that the ECCS will perform as designed by initiating the 
disconnect of the 4.16 kV emergency buses from the offsite power 
sources and connected to the onsite DG power sources, if it is 
determined that insufficient offsite voltage is available. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to modify the OPERABILITY of the Degraded 
Voltage--Time Delay, LOCA, function to be consistent with the 
OPERABILITY of the ECCS instrumentation that generate the timer 
initiating LOCA signal, will not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed changes do not introduce any new 
equipment, modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The Degraded Voltage Time Delay circuitry is composed of two 
time delay components. Upon detection of a degraded voltage 
condition, the Degraded Voltage--Time Delay, No LOCA, function timer 
is initiated with a TS Allowable Value of [ge] 270.1 seconds and 
[le] 329.9 seconds. If a coincident LOCA signal is detected, the 
Degraded Voltage--Time Delay, No LOCA, function timer is bypassed 
and the Degraded Voltage--Time Delay, LOCA, function timer is 
initiated. The Degraded Voltage--Time Delay, LOCA, function timer 
has a TS Allowable Value of [ge] 9.4 seconds and [le] 10.9 seconds. 
The Time Delay Allowable Values are long enough to provide time for 
the offsite power supply to recover to normal voltages, but short 
enough to ensure that sufficient power is available to the required 
equipment. The shorter time delay associated with a coincident LOCA 
signal is required to ensure that the ECCS injection assumptions of 
the LOCA analyses are met. The proposed changes do not affect the 
Time Delay Allowable Values.
    The Drywell Pressure--High instrumentation is required to be 
OPERABLE in MODES 1, 2 and 3. In MODES 4 and 5, the Drywell 
Pressure--High instrumentation is not required to be OPERABLE since 
there is insufficient energy in the reactor to pressurize the 
drywell to the Drywell Pressure--High setpoint.
    The Reactor Vessel Water Level--Low Low Low, Level 1 and Reactor 
Vessel Water Level--Low Low, Level 2 ECCS instrumentation is 
required to be OPERABLE in MODES 1, 2, 3 and 4. In MODE 5, the ECCS 
instrumentation is required to be OPERABLE except with the spent 
fuel storage pool gates removed and the water level [ge] 22 feet 
over the top of the reactor pressure vessel flange. In this 
situation, the water level provides sufficient coolant inventory to 
allow operator action to terminate the inventory loss prior to fuel 
uncovery in case of an inadvertent draindown.
    The Drywell Pressure--High, Reactor Vessel Water Level--Low Low 
Low, Level 1 and the Reactor Vessel Water Level--Low Low, Level 2 
ECCS instrumentation are not

[[Page 53987]]

required to be OPERABLE when not in MODES 1, 2, 3, 4, and 5 (i.e., 
no fuel in the vessel).
    The proposed changes will modify the OPERABILITY of the Degraded 
Voltage--Time Delay, LOCA, function to be consistent with the 
OPERABILITY of the above described ECCS instrumentation that 
generate the timer initiating LOCA signal. Thus, the proposed 
changes are consistent with the ECCS injection assumptions of the 
LOCA analyses.
    The Degraded Voltage--Time Delay, No LOCA, function provides 
adequate protection to ensure that other required systems powered 
from the diesel generators (DGs) function as designed in any non-
LOCA accident in which a loss of power is assumed when the Degraded 
Voltage--Time Delay, LOCA, function is not required to be OPERABLE.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: July 18, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) regarding Engineered Safety 
Feature Actuation System instrumentation. Specifically, they would 
limit the period of time that inoperable recirculation actuation signal 
(RAS), containment spray actuation signal (CSAS), and auxiliary 
feedwater actuation signal (AFAS) input channels could be in the 
bypassed and/or tripped condition. Generally, the proposed TS employ a 
48-hour completion time to restore an inoperable channel, which, in 
most cases, is more restrictive than the existing TS, is comparable to 
the value used in the Standard TS for Combustion Engineering plants, 
and is a reasonable expected repair time based on plant maintenance 
history.
    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. Would operation of the facility in accordance with the 
proposed amendments involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No, facility operation under the new Technical Specification 
(TS) restrictions would not increase the probability of occurrence 
of any accident previously evaluated. The proposed changes only 
affect the emergency safety features actuation system (ESFAS) 
functions of RAS, CSAS, and AFAS; generally limiting the time that 
any instrument channel may be inoperable in a bypassed or tripped 
condition. No physical plant changes are proposed in conjunction 
with these revisions. The proposed changes to RAS and AFAS channel 
operability greatly reduce the time that actuation systems are 
vulnerable to spurious, inadvertent actuation. The proposed changes 
do allow a new unlimited time for trip of one CSAS channel on Unit 
1. Unit 2 already contains provision for the indefinite single 
channel trip of CSAS, and this change will also make the two units 
similar. Additionally, it is important to note that inadvertent 
actuation of any of these functions (RAS, CSAS, or AFAS) during 
plant operation is not an accident initiating event. Therefore, with 
no physical effects on the plant and no increase in probability that 
the subject ESFAS functions will initiate an accident, there is no 
increased probability that any previously evaluated accident will 
occur. The changes provided in this safety evaluation do not affect 
the assumptions or results of any accident evaluated in the UFSAR 
[Updated Final Safety Analysis Report].
    Likewise, the consequences of any accident previously evaluated 
have not been increased. The proposed changes, by limiting the time 
that ESFAS functions are inoperable, will increase the reliability 
of the associated ESFAS functions to respond to accidents. In 
particular, the revision to the RAS TS will limit the time that the 
RAS will be vulnerable to single failure and will therefore improve 
the system reliability during an accident. As these proposed changes 
constitute no physical change to the facility and only serve to 
increase ESFAS function reliability, FPL concludes that the 
consequences of previously evaluated accidents are not increased. 
The ability of the ESFAS to respond to accident conditions as 
assumed in any accident analysis has not been affected.
    2. Would operation of the facility in accordance with the 
proposed amendments create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No, the proposed activity does not create the possibility of an 
accident of a different type than any previously evaluated. The 
proposed changes only affect the ESFAS functions of RAS, CSAS, and 
AFAS; generally limiting the time that any instrument channel may be 
inoperable in a bypassed or tripped condition. No physical plant 
changes are proposed in conjunction with these revisions. Thereby, 
the proposed changes do not create any new equipment interfaces, 
equipment response characteristics, or operating configurations. 
Without creation of a new interaction of materials, operating 
configuration, or operating interface, there is no possibility that 
the proposed changes can introduce a new or different kind of 
accident.
    3. Would operation of the facility in accordance with the 
proposed amendments involve a significant reduction in a margin of 
safety?
    The margin of safety as defined in the basis for any Technical 
Specification or in any licensing document has not been reduced. 
Except for the change in end state specified for the AFAS automatic 
actuation logic LCO [Limiting Condition for Operation], the TS Bases 
for the associated ESFAS LCO do not explicitly discuss a related 
margin of safety. Changing the AFAS automatic actuation logic LCO 
end state from Mode 5 to Mode 4 is not a reduction in a margin of 
safety. That proposed change is consistent with the TS applicability 
for the AFAS and auxiliary feedwater systems as well as the bases 
for TS LCOs. Additionally, by virtue of the increased ESFAS 
reliability provided by the proposed amendments, it is evident that 
the margin of safety will not be reduced in any manner.

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

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: June 24, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less* * *'' to ``* * *up to 24 hours or up to 
the limit of the specified Frequency, whichever is greater* * *.'' In 
addition, the following requirement would be added to SR 3.0.3: ``A 
risk evaluation shall be performed for any Surveillance delayed greater 
than 24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal

[[Page 53988]]

Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated June 24, 2002. The NSHC determination is restated below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in [a] margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on [a] margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: July 17, 2002.
    Description of amendment request: The proposed amendment would 
change the Unit 2 Technical Specifications (TSs) by including the Unit 
2 Cycle 12 (U2C12) Minimum Critical Power Ratio (MCPR) Safety Limits in 
Section 2.1.1.2, changing the references listed in Section 5.6.5.b, and 
changing the design features in Section 4.2.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No: The proposed change to the MCPR Safety Limits does not 
directly or indirectly affect any plant system, equipment, 
component, or change the processes used to operate the plant. 
Further, the U2C12 MCPR Safety Limits are generated using NRC 
approved methodology and meet the applicable acceptance criteria. 
Thus, this proposed amendment does not involve a significant 
increase in the probability of occurrence of an accident previously 
evaluated.
    Prior to the startup of U2C12, licensing analyses are performed 
(using NRC approved methodology referenced in Technical 
Specification Section 5.6.5.b) to determine changes in the critical 
power ratio as a result of anticipated operational occurrences. 
These results are added to the MCPR Safety Limit values proposed 
herein to generate the MCPR operating limits in the U2C12 [Core 
Operating Limits Report] COLR. These limits could be different from 
those specified for the U2C11 COLR. The COLR operating limits thus 
assure that the MCPR Safety Limit will not be exceeded during normal 
operation or anticipated operational occurrences, thus providing the 
required level of protection for the fuel rod cladding. Postulated 
accidents are also analyzed prior to the startup of U2C12 and the 
results shown to be within the NRC approved criteria. The proposed 
change to the MCPR Safety Limit will have a negligible impact on the 
results of these accident analyses.
    The U2C12 reload fuel bundles will utilize a small amount of 
depleted uranium (``tails'') in certain fuel rods, in addition to 
natural and slightly enriched uranium. There is no change to the 
composition of the fuel pellets containing tails material, (i.e., 
UO2) except a slight decrease in the amount of [uranium-235] 
U235. Therefore, the use of depleted uranium (``tails'') 
in the fuel rods does not affect the mechanical performance of the 
fuel rods. The impact of the use of tails on core performance is 
included in the reload licensing analysis.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C12 core operating limits. The use of this approved methodology 
does not increase the probability of occurrence or consequences of 
an accident previously evaluated.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously analyzed?
    No: The change to the MCPR Safety Limits and the U2C12 core 
loading which it supports does not directly or indirectly affect any 
plant system, equipment, or component (other than the core itself) 
and therefore does not affect the failure modes of any of these. 
Thus, the proposed [change does] not create the possibility of a 
previously unevaluated operator error or a new single failure.
    The use of depleted uranium (``tails'') in the fuel rods does 
not affect the mechanical performance of the fuel rods.

[[Page 53989]]

    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C12 core operating limits. The use of this approved methodology 
does not create the possibility of a new or different kind of 
accident.
    Therefore, this proposed amendment does not involve a 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No: Since the proposed [change does] not alter any plant system, 
equipment, component, or the processes used to operate the plant, 
the proposed change will not jeopardize or degrade the function or 
operation of any plant system or component governed by Technical 
Specifications. The proposed MCPR Safety Limits do not involve a 
significant reduction in the margin of safety as currently defined 
in the Bases of the applicable Technical Specification sections, 
because the MCPR Safety Limits calculated for U2C12 preserve the 
required margin of safety.
    The use of depleted uranium (``tails'') in the fuel rods does 
not affect the mechanical performance of the fuel rods.
    The changes to the references in Section 5.6.5.b were made to 
properly reflect the NRC approved methodology used to generate the 
U2C12 core operating limits. This approved methodology is used to 
demonstrate that all applicable criteria are met, thus, 
demonstrating that there is no reduction in the margin of safety.
    Therefore, these changes do not involve a significant reduction 
in [the] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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: July 18, 2002.
    Description of amendment request: The licensee proposes to change 
Salem Technical Specifications (TSs) requirements associated with its 
containment spray nozzles. The frequency of TS Surveillance Requirement 
(SR) 4.6.2.1.d for verifying that the containment spray nozzles are 
unobstructed would be changed from a fixed 10-year frequency to after 
activities that could result in nozzle blockage. PSEG proposes to 
either evaluate the work performed to determine the impact to the 
containment spray system, or perform an air or smoke flow test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the testing requirements for the 
containment spray nozzles to only require verification that each 
spray nozzle is unobstructed following activities that could result 
in nozzle blockage. The proposed change does not have a detrimental 
impact on the integrity of any plant structure, system, or component 
that initiates an analyzed event. No active or passive failure 
mechanisms that could lead to an accident are affected. The proposed 
change will not alter the operation of, or otherwise increase the 
failure probability of any plant equipment that initiates an 
analyzed accident. The containment spray system is not an accident 
initiator but is used for mitigation of design basis accidents. As a 
result, the probability of any accident previously evaluated, is not 
significantly increased.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change revises the current 
Surveillance Frequency from 10 years to following activities that 
could result in spray nozzle blockage. Since activities that could 
introduce foreign material into the system (such as inadvertent 
actuation of the containment spray system or loss of foreign 
material control) are the most likely cause for obstruction, testing 
or inspection following such activities would verify that the 
nozzle(s) are unobstructed, and the system is capable of performing 
its safety function. No other evolutions require the system boundary 
to be breached, so introduction of debris during times when 
maintenance activities are not in progress are precluded. 
Introduction of foreign materials into the system from the exterior 
is highly unlikely due to the location of the spray headers, the 
passive nature of the nozzles, and the fact that the stainless steel 
containment spray headers are maintained dry which does not lend 
itself to active degradation mechanisms such as corrosion. The 
proposed testing requirements are considered sufficient to provide a 
high degree of confidence that containment spray will function when 
required.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the test frequency for the containment 
spray system nozzles does not involve the use or installation of new 
equipment. Installed equipment is not operated in a new or different 
manner. No new or different system interactions are created, and no 
new processes are introduced. The current foreign material exclusion 
practices have been reviewed and judged sufficient to provide high 
confidence that debris will not be introduced during times when the 
system boundary is breached.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The revision to the containment spray nozzle testing frequency 
does not introduce any new setpoints at which protective or 
mitigative actions are initiated. No current setpoints are altered 
by this change. The design and functioning of the containment spray 
system is unchanged. Since the system is not susceptible to 
corrosion induced obstruction nor is the introduction of foreign 
material from the exterior likely, the proposed testing frequency is 
sufficient to provide high confidence that the containment spray 
system will be available to provide the flow necessary to mitigate 
the consequences of a design basis accident. Therefore, the 
capability of the system will remain unchanged. As a result, 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 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: Jacob I. Zimmerman, Acting.

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: July 29, 2002.
    Description of amendment request: The proposed change to the 
Technical Specifications (TSs) would revise the requirements for 
containment closure associated with the equipment hatch and personnel 
airlocks during Core Alterations and movement of irradiated fuel within 
the containment. This proposed change would allow the equipment hatch 
and the personnel airlocks to remain open during fuel movement in the 
containment provided

[[Page 53990]]

administrative controls are developed and implemented, ensuring the 
closure of the equipment hatch and personnel airlock following a fuel 
handling accident within the containment building. In addition, the 
associated TS Bases are revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    An alternate source term calculation has been performed for 
Salem Nuclear Station that demonstrates that offsite and control 
room dose consequences of a postulated fuel handling accident remain 
within the limits provided sufficient decay has occurred prior to 
the movement of irradiated fuel without taking credit for certain 
mitigation features such as ventilation filter systems and 
containment closure. Fuel movement is allowed provided that 
irradiated fuel has undergone the required decay time.
    The proposed amendment would allow movement of sufficiently 
decayed irradiated fuel within the containment building with the 
equipment hatch and personnel air locks open provided that 
administrative controls are implemented to promptly (within 1 hour) 
close the containment penetrations.
    Either the Containment Purge system or the Auxiliary Building 
Ventilation System with suction from the containment atmosphere, 
with associated radiation monitoring will be available whenever 
movement of irradiated fuel is in progress in the containment 
building and the equipment hatch is open. If for any reason, this 
ventilation requirement can not be met, movement of fuel assemblies 
within the containment building shall be discontinued until the flow 
path(s) can be reestablished or close the equipment hatch and 
personnel airlocks. The amendment also would allow movement of 
irradiated fuel assemblies within the Fuel Handling Building with 
the Fuel Handling Area Ventilation System (FHAVS) in operation but 
no credit taken for filtration.
    This amendment does not alter the methodology of the FHA [Fuel-
Handling Accident] or equipment used directly in fuel handling 
operations. Neither ventilation filter systems, the CPES 
[Containment Purge Exhaust System] nor the FHAVS, is used to 
actually handle fuel. Therefore neither of these systems is an 
``accident initiator''. Similarly, neither the equipment hatch, the 
personnel air locks, nor any other containment penetration, nor any 
component thereof is an accident initiator.
    In the postulated Fuel Handling Accident, the revised dose 
calculations, performed using 10 CFR 50.67 and Regulatory Guide 
1.183, Alternative Source Term, do not take credit for automatic 
containment purge isolation thus allowing for continuous monitoring 
of containment activity until containment closure is achieved. If 
required, containment purge isolation can be initiated manually from 
the control room.
    Actual fuel handling operations are not affected by the proposed 
changes. Therefore, the probability of a Fuel Handling Accident is 
not affected with the proposed amendment. No other accident 
initiator is affected by the proposed changes.
    The FHA in the Fuel Handling Building has been analyzed without 
credit for filtration by the FHAVS. The analyses of these design 
basis events were conducted with the Alternative Source Term 
Methodology in accordance with 10 CFR 50.67 and Regulatory Guide 
1.183. These analyses show that the resultant radiation doses are 
within the limits specified in these documents.
    The TEDE [Total Effective Dose Equivalent] radiation doses from 
the analyses supporting this LCR [License Change Request] have been 
compared to equivalent TEDE radiation doses estimated with the 
guidelines of R.G. [Regulatory Guide] 1.183. The new values are 
shown to be within the regulatory guidelines.
    The revision to the definition of Core Alterations simply 
reflects the definition in the Standard Technical Specifications, 
NUREG 1431 for Westinghouse Plants and is supported by the bounding 
effects of the Fuel Handling Accident analysis.
    The deletion of Core Alterations from the APPLICABILITY section 
of the affected LCO's [Limiting Conditions for Operation] is based 
on the fact that, during Core Alterations only, the FHA results in 
cladding damage and potential radiological release. Consequently, 
the deletion of Core Alterations is consistent with industry 
approved practice and guidance documents (ex: TSTF [Technical 
Specification Task Force]-51, revision 2).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve addition or modification to 
any plant system, structure, or component. The proposed amendment 
would permit the equipment hatch and personnel air locks to be open 
during movement of irradiated fuel. The proposed amendment does not 
involve any change in the operation of these containment 
penetrations. Having these penetrations open does not create the 
possibility of a new accident.
    The proposed amendment also would remove the requirements for 
operability of the FHAVS Filtration System during movement of 
sufficiently decayed irradiated fuel. It does not alter the 
operation of these systems. Therefore, the system is not an accident 
initiator. Modification of the requirements of operability for the 
system from the plant Technical Specifications does not create the 
possibility of a new accident.
    The revision to the definition of Core Alterations simply 
reflects the industry position supported by the definition in the 
Standard Technical Specifications, NUREG 1431 for Westinghouse 
Plants and is supported by the bounding effects of the Fuel Handling 
Accident analysis.
    The deletion of Core Alterations from the APPLICABILITY section 
of the affected LCO's is based on the fact that, during Core 
Alterations only, the FHA results in cladding damage and potential 
radiological release. Consequently, the deletion of Core Alterations 
is consistent with industry approved practice and guidance documents 
(ex: TSTF-51, revision 2).
    The proposed amendment does not create the possibility of a new 
or different kind of accident than any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The assumptions and input used in the analysis are conservative 
as noted below. The design basis Fuel Handling Accidents have been 
defined to identify conservative conditions. The source term and 
radioactivity releases have been calculated pursuant to Regulatory 
Guide 1.183 and with conservative assumptions concerning prior 
reactor operation. The control room atmospheric dispersion factors 
have been calculated with conservative assumptions associated with 
the release. The conservative assumptions and input noted above 
ensure that the radiation doses cited in this License Change Request 
are the upper bound to radiological consequences of a Fuel Handling 
Accident either in Containment or the Fuel Handling Building. The 
analyses show that there is a significant margin between the TEDE 
radiation doses calculated for the postulated Fuel Handling Accident 
using the Alternative Source Term and the acceptance limits of 10 
CFR 50.67 and Regulatory Guide 1.183.
    The revision to the definition of Core Alterations simply 
reflects the industry position supported by the definition in the 
Standard Technical Specifications, NUREG 1431 for Westinghouse 
Plants and is supported by the bounding effects of the Fuel Handling 
Accident analysis.
    The deletion of Core Alterations from the APPLICABILITY section 
of the affected LCO's is based on the fact that, during Core 
Alterations only the FHA results in cladding damage and potential 
radiological release. Consequently, the deletion of Core Alterations 
is consistent with industry approved practice and guidance documents 
(ex: TSTF-51, revision 2).
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 53991]]

    NRC Section Chief: Jacob Zimmerman, Acting.

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

    Date of amendment request: March 20, 2002.
    Description of amendment request: This proposed change revises the 
iodine dose conversion factors used in the determination of the dose 
equivalent I-131 reactor coolant specific activity and in the 
calculation of the offsite radiological consequences for those Final 
Safety Analysis Report (FSAR) Chapter 15 accidents that include iodine 
spiking effects.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    This proposed change revises the iodine dose conversion factors 
used in the determination of the dose equivalent I-131 reactor 
coolant specific activity and in the calculation of the offsite 
radiological consequences for those FSAR Chapter 15 accidents that 
include iodine spiking effects. The iodine dose conversion factors 
are changed from the values in TID [Technical Information Document] 
14844 to the values in ICRP [International Commission on 
Radiological Protection] 30, consistent with NUREG-1431. The 
accidents affected by this change are the steam generator tube 
rupture, main steam line break and CVCS [Chemical and Volume Control 
System] line rupture. The proposed change also revises certain input 
assumptions (letdown demineralize iodine removal efficiency, primary 
coolant leakage and uncertainty in letdown flow) used in determining 
the accident initiated (concurrent) iodine spiking source terms 
input to the offsite radiological consequences calculations. The 
change in dose conversion factors and the input assumptions does not 
affect any normal operation or accident scenarios. There are no 
changes to any plant procedures or equipment that would relate to 
the probability of an accident. The change in the iodine spiking 
input assumptions identified in NSAL [Nuclear Safety Advisory 
Letter]-00-04 results in an increase in the calculated offsite dose 
consequences for the steam generator tube rupture, main steam line 
break and CVCS line rupture. Use of the ICRP 30 iodine dose 
conversion factors offsets this increase such that the resulting 
calculated offsite dose consequences are less severe than those 
previously presented in the FSAR for the steam generator tube 
rupture, main steam line break and CVCS line rupture. * * *
    Thyroid doses for the other accidents described in the FSAR will 
continue to be reported using the conservative TID 14844 iodine dose 
conversion factors until a future update is required.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes to Technical Specification [TS] 1.10, 
Definitions, Dose Equivalent I-131 and the use of the iodine spiking 
input assumptions listed in NSAL-00-04 [do] not introduce any new 
accident initiator mechanisms. The dose conversion factors are used 
in determining the reactor coolant dose equivalent I-131 specific 
activity and in the calculation of offsite dose consequences for 
certain design basis accidents which include the effects of iodine 
spiking as revised based on the iodine spiking input changes 
(letdown demineralizer iodine removal efficiency, primary coolant 
leakage and uncertainty in letdown flow) provided in NSAL-00-04. No 
existing accident scenarios are affected and no new scenarios are 
created. The proposed change does not introduce alterations to 
system operations, changes to equipment operability or technical 
specification operability requirements, nor to Engineered Safety 
Features Actuation System instrumentation or setpoints. The proposed 
change does not revise any of the actual equipment or 
instrumentation in the plant nor does it change the actual alarm 
setpoints or information available to the operators to monitor 
Technical Specification commitments. It does not introduce any new 
or different failure mechanisms or limiting single failures. A new 
or different kind of accident is thus not created.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change to Technical Specification 1.10, 
Definitions, Dose Equivalent I-131 preserves the conclusions of 
plant safety analyses presented in the FSAR. This proposed change 
revises the iodine dose conversion factors used in the calculation 
of the potential offsite radiological consequences following those 
Chapter 15 accidents that include iodine spiking effects as revised 
based on the iodine spiking input changes provided in NSAL-00-04. 
The dose conversion factors are changed from the values in TID-14844 
to the values in ICRP 30, consistent with the criteria in NUREG-
1431. This activity relates to TS section B3/4.4.8 and TS section 
1.10 Dose Equivalent I-131. TS section B3/4.4.8 states that the 
limitation on the specific activity of the primary coolant ensures 
that the resulting 2 hour doses at the site boundary will not exceed 
an appropriately small fraction of Part 100 limits following a steam 
generator tube rupture accident. TS Section 1.10 defines the 
acceptable values for the iodine dose conversion factors. The change 
in the accident initiated iodine spiking calculation input 
parameters identified in NSAL-00-04 results in an increase in the 
calculated offsite dose consequences for the steam generator tube 
rupture, main steam line break and CVCS line rupture. Use of the 
ICRP 30 iodine dose conversion factors offsets this increase such 
that the resulting calculated offsite dose consequences are less 
severe than those previously presented in the FSAR. Therefore, the 
margin of safety is not reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: March 4, 2002, as supplemented by letter 
dated July 11, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical specifications (TS) 5.5.9.3.a, ``Steam Generator Tube 
Surveillance Program, Inspection Frequencies.'' Specifically, the 
proposed changes would revise the Farley Nuclear Plant, Unit 1 TS to 
allow a 40 month inspection interval after its first (post-replacement) 
inservice inspection, rather than after two consecutive inspections 
resulting in C-1 classification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed one-time change revises the steam generator (SG) 
inspection interval requirements in Technical Specification (TS) 
5.5.9.3, ``Steam Generator (SG) Tube Surveillance Program, 
Inspection Frequencies,'' for the Farley Nuclear Plant, Unit 1, 
Spring 2003 refueling outage, to allow a 40 month inspection 
frequency after one inspection, rather than after two consecutive 
inspections with results that are within the C-1 category. C-1 
category is defined as ``less than 5% of the total tubes inspected 
are degraded tubes and none of the inspected tubes are defective.''
    The proposed one-time extension of the Unit 1 SG tube inservice 
inspection interval does not involve changing any structure, system, 
or component, or affect reactor operations. It is not an initiator 
of an accident and does not change any existing safety analysis 
previously analyzed in the Farley Nuclear Plants' Final Safety 
Analysis Report

[[Page 53992]]

(FSAR). As such, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    Since the proposed change does not alter the plant design, there 
is no direct increase in SG leakage. Industry experience indicates 
that the probability of increased SG tube degradation would not go 
undetected.
    Additionally, steps described below will further minimize the 
risk associated with this extension. For example, the scope of 
inspections performed during the last Farley Nuclear Plant, Unit 1, 
refueling outage (i.e., the first refueling outage following SG 
replacement) exceeded the TS requirements for the first two 
refueling outages after SG replacement. That is, more tubes were 
inspected than were required by TS. Currently, Farley Nuclear Plant, 
Unit 1, does not have an SG damage mechanism, and will meet the 
current industry examination guidelines without performing SG 
inspections during the next refueling outage. Additionally, as part 
of our SG Program, both a Condition Monitoring Assessment and an 
Operational Assessment are performed after each inspection and 
compared to the Nuclear Energy Institute (NEI) 97-06, ``Steam 
Generator Program Guidelines,'' performance criteria. The results of 
the Condition Monitoring Assessment demonstrated that all 
performance criteria were met during the Farley Nuclear Plant, Unit 
1, Fall 2001 refueling outage, and the results of the Operational 
Assessment show that all performance criteria will be met over the 
proposed operating period. Considering these actions, along with the 
improved SG design and reliability of Westinghouse replacement SGs, 
extending the SG tube inspection frequency does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change revises the SG inspection frequency 
requirements in TS 5.5.9.3.a, ``Steam Generator (SG) Tube 
Surveillance Program, Inspection Frequencies,'' for the Farley 
Nuclear Plant, Unit 1, Spring 2003 refueling outage, to allow a 40 
month inspection interval after one inspection, rather than after 
two consecutive inspections, with inspection results within the C-1 
category.
    The proposed change will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections performed during the last Farley Nuclear 
Plant, Unit 1, refueling outage (i.e., the first refueling outage 
following SG replacement) significantly exceeded the TS requirements 
for the scope of the first two refueling outages after SG 
replacement.
    Primary-to-secondary leakage that may be experienced during all 
plant conditions is expected to remain within current accident 
analysis assumptions. The proposed change does not affect the design 
of the SGs, the method of SG operation, or reactor coolant chemistry 
controls. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. The 
proposed change involves a one-time extension to the SG tube 
inservice inspection frequency, and, therefore, will not give rise 
to new failure modes. In addition, the proposed change does not 
impact any other plant systems or components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The SG tubes are an integral part of the Reactor Coolant System 
(RCS) pressure boundary that are relied upon to maintain the RCS 
pressure and inventory. The SG tubes isolate the radioactive fission 
products in the reactor coolant from the secondary system. The 
safety function of the SGs is maintained by ensuring the integrity 
of the SG tubes. In addition, the SG tubes comprise the heat 
transfer surface between the primary and secondary systems such that 
residual heat can be removed from the primary system.
    SG tube integrity is a function of the design, environment, and 
current physical condition. Extending the SG tube inservice 
inspection frequency by one operating cycle will not alter the 
function or design of the SGs. SG inspections conducted during the 
first refueling outage following SG replacement demonstrated that 
the SGs do not have an active damage mechanism, and the scope of 
those inspections significantly exceeded those required by the TS. 
These inspection results were comparable to similar inspection 
results for second generation alloy 690 models of replacement SGs 
installed at other plants, and subsequent inspections at those 
plants yielded results that support this extension request. The 
improved design of the replacement SGs also provides reasonable 
assurance that significant tube degradation is not likely to occur 
over the proposed operating period.
    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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: July 25, 2002.
    Description of amendment request: The amendment would revise 
Surveillance Requirements 3.3.1.2 and 3.3.1.3 of Technical 
Specification 3.3.1, ``Reactor Trip System (RTS) Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The RTS instrumentation will be unaffected. 
Protection systems will continue to function in a manner consistent 
with the plant design basis. All design, material, and construction 
standards that were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the USAR [Updated Safety Analysis Report] are not 
adversely affected because the change to the NIS [Nuclear 
Instrumentation System] power range channel daily surveillance 
assures the conservative response of the channel even at part-power 
levels.
    The proposed changes modify the NIS power range channel daily 
surveillance requirement to assure the NIS power range functions are 
tested in a manner consistent with the safety analysis and licensing 
basis.
    The proposed changes will not affect the probability of any 
event initiators. There will be no degradation in the performance 
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation. 
There will be no change to the normal plant operating parameters or 
accident mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the USAR.
    Therefore, the proposed changes do 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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements or response time limits will be affected; however, the 
proposed TS Bases changes impose explicit NIS power range high trip 
setpoint adjustment requirements prior to adjusting indicated power 
in a decreasing power direction. These requirements are consistent

[[Page 53993]]

with assumptions made in the safety analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes require a revision to the criteria for 
implementation of NIS power range channel adjustments based on 
secondary power calorimetric calculations; however, the changes do 
not eliminate any RTS surveillances or alter the frequency of 
surveillances required by the Technical Specifications. The revision 
to the criteria for implementation of the daily surveillance will 
have a conservative effect on the performance of the NIS power range 
channels, particularly at part-power conditions. The nominal trip 
setpoints specified in the Technical Specification Bases and the 
safety analysis limits assumed in the transient and accident 
analyses are unchanged. None of the acceptance criteria for any 
accident analysis is changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the output power 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (F[Delta]H), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    The imposition of appropriate surveillance testing requirement 
will not reduce any margin of safety since the changes will assure 
that safety analysis assumptions on equipment operability are 
verified on a periodic frequency.
    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: July 25, 2002.
    Description of amendment request: The amendment would revise 
Chapter 5.0, ``Administrative Controls,'' of the technical 
specifications (TSs) to allow the use of generic personnel titles in 
the TSs in place of plant-specific personnel titles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect accident initiators or 
assumptions. The radiological consequences of an accident previously 
evaluated remain unchanged. These changes involve administrative 
changes concerning the use of personnel titles and do not affect 
responsibilities or qualifications of plant personnel.
    Therefore, the proposed changes do 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 changes are administrative in nature. As such, 
there are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating parameters. No new accident 
scenarios, transient precursors, failure mechanisms, or limiting 
single failures are introduced as a result of this amendment. There 
will be no adverse effects or challenges imposed on any safety-
related system as a result of this amendment.
    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 a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. The use of generic personnel titles will 
not reduce any margin of safety. (These changes involve 
administrative changes concerning the use of personnel titles and do 
not affect responsibilities or qualifications of plant personnel.)
    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

South Carolina Electric & Gas, Docket No. 50-395, Virgil C. Summer 
Nuclear Station, Fairfield County, South Carolina

    Date of amendment request: July 24, 2001, as supplemented April 4, 
2002, May 7, 2002, June 17, 2002, July 2, 2002, July 15, 2002, and July 
25, 2002.
    Brief description of amendment request: This amendment would 
increase the spent fuel pool storage capacity by replacing all 11 
existing rack modules with 12 new high density storage racks. The 
rerack will increase the storage capacity from 1,276 storage cells to 
1,712 storage cells. The degrading Boraflex neutron-absorbing material 
in the existing racks will be replaced by Boral material that will be 
used in the new racks.
    Date of publication of individual notice in Federal Register: June 
25, 2002 (67 FR 42810).
    Expiration date of individual notice: July 25, 2002.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following

[[Page 53994]]

amendments. The Commission has determined for each of these amendments 
that the application complies with the standards and requirements of 
the Atomic Energy Act of 1954, as amended (the Act), and the 
Commission's rules and regulations. The Commission has made appropriate 
findings as required by the Act and the Commission's rules and 
regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

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

    Date of application for amendments: December 13, 2001, as 
supplemented by letter dated May 1, 2002.
    Brief description of amendments: The amendments add the following 
to the Technical Specifications: (1) The phrase, ``or if open, capable 
of being closed,'' to the Limiting Condition for Operation 3.9.3 for 
the equipment hatch, during core alterations or movement of irradiated 
fuel assemblies inside containment, and (2) the requirement to verify 
the capability to close the equipment hatch in a new Surveillance 
Requirement 3.9.3.3. The amendments allow the equipment hatch to be 
open in refueling outages during the conditions stated above.
    Date of issuance: July 25, 2002.
    Effective date: July 25, 2002, and shall be implemented within 90 
days of the date of issuance, including the incorporation of the 
changes to the Technical Specification Bases as described in the 
licensee's letters dated December 13, 2001, and May 1, 2002.
    Amendment Nos.: Unit 1-143, Unit 2-143, Unit 3-143.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2919). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 31, 2002.
    Brief description of amendments: These amendments correct 
administrative errors in Section 5.6.5, ``Core Operating Limits Report 
(COLR),'' of the Technical Specifications and Section 2.0, 
``Environmental Protection Issues,'' of the Environmental Protection 
Program.
    Date of issuance: August 6, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 254/231.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36927). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 6, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: August 28, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3/4.5.1, ``Safety Injection Tanks (SITs)'' to delete 
surveillance requirement (SR) 4.5.1.f. This SR provided verification of 
the automatic opening features of the SIT outlet isolation valves.
    Date of issuance: August 7, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 268.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55010). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 7, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: September 26, 2001.
    Brief description of amendment: The amendment modifies the 
Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications (TSs) to relocate MP3 TSs related to the control rod 
position indication system requirements for shutdown to the licensee-
controlled Technical Requirements Manual (TRM). The Index and Bases 
pages of the affected TSs will also be modified to address the proposed 
changes.
    Date of issuance: July 30, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days from the date of issuance.
    Amendment No.: 207.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57120). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.

[[Page 53995]]

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

    Date of application for amendments: April 16, 2002.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period before entering a 
Limiting Condition for Operation following a missed surveillance. The 
delay period is extended from the current limit of `` * * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to `` * * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: August 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 201 & 194.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34482). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 1, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 16, 2002.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period before entering a 
Limiting Condition for Operation following a missed surveillance. The 
delay period is extended from the current limit of `` * * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to `` * * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: July 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 205/186.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34483). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.
    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina
    Date of application of amendments: April 16, 2002.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of `` * * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to `` * * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of Issuance: July 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 327, 327, 328.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34483). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: June 13, 2002.
    Brief description of amendment: The amendment revised Technical 
Specifications Section 4.13.A, ``Inspection Requirements,'' to allow 
the use of the optimum eddy current probe size when performing steam 
generator tube inspections. The amendment also corrects grammatical and 
typographical errors.
    Date of issuance: July 29, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 230.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42806). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 29, 2002.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: January 8, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.1.B, ``Heatup and Cooldown,'' to delete the 
requirements governing the reactor vessel surveillance program, 
including the reactor vessel specimen withdrawal schedule. In addition, 
the changes corrected errors in TS 4.2, ``Inservice Inspection and 
Testing;'' TS 5.2.C, ``Design Features--Containment;'' and TS 6.4, 
``Administrative Controls--Training.'' TS Sections 6.1, 
``Responsibility'' and 6.2, ``Organization'' were changed to reflect 
the organizational changes resulting from the transfer of the operating 
license to Entergy Nuclear Operations, Inc. on September 6, 2001.
    Date of issuance: July 30, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 231.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10011). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
County, Pennsylvania

    Date of application for amendments: March 19, 2002.
    Brief description of amendments: These amendments allow plant 
operation to continue if the temperature of the normal heat sink (NHS) 
exceeds the Technical Specification (TS) limit of 90  deg.F provided 
that the NHS temperature averaged over the previous

[[Page 53996]]

24-hour period is verified at least once per hour to be less than or 
equal to 90  deg.F, and the NHS temperature does not exceed a maximum 
value of 92  deg.F. The format for this change had been previously 
approved by the Nuclear Regulatory Commission for the Standard TSs as 
per TS Task Force (TSTF) change TSTF-330, Revision 3, ``Allowed Outage 
Time--Ultimate Heat Sink'', on October 13, 2000. In addition, an 
administrative change removes references to a temporary TS change which 
expired on May 31, 2000.
    Date of issuance: July 29, 2002.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendments Nos.: 244/248.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34486). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 29, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254, Quad Cities Nuclear 
Power Station, Unit 1, Rock Island County, Illinois

    Date of application for amendment: April 8, 2002, as supplemented 
June 18 and July 3, 2002.
    Brief description of amendment: The amendment revises the safety 
limit minimum critical power ratio for two-loop and single-loop 
operation.
    Date of issuance: July 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-29: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34487). The supplements dated June 18 and July 3, 2002, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 29, 2002.
    No significant hazards consideration comments received: No.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: August 16, 2001, as supplemented 
by letter dated November 19, 2001.
    Brief description of amendment: The amendment revises the license 
to incorporate a new License Condition 2.B.(9). The license condition 
terminates license jurisdiction for a portion of the Maine Yankee 
Atomic Power Station site (referred to as the Non-Impacted Backlands 
(west of Bailey Cove and west of Young's Brook and north of Old Ferry 
Road)), thereby releasing these lands from Facility Operating License 
No. DPR-36.
    Date of issuance: July 30, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. DPR-36: The amendment revised the 
license.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12604). The November 19, 2001, supplemental letter provided additional 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: November 26, 2001, as 
supplemented on May 20, 2002.
    Brief description of amendment: The amendment deleted Section 3/
4.2.6, ``Inservice Inspection and Testing,'' revised Section 4.2.7, 
``Reactor Coolant System Isolation Valves,'' added a new Section 6.17, 
``Inservice Testing Program,'' and deleted several reporting 
requirements in Section 6.9.3, ``Special Reports.''
    Date of issuance: August 5, 2002.
    Effective date: August 5, 2002.
    Amendment No.: 173.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66468). The May 20, 2002, supplemental letter provided clarifying 
information that was within the scope of the amendment request 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 August 5, 2002.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendments: June 8, 2001, as supplemented 
by letters dated February 4, April 8, May 7, June 6, and June 28, 2002.
    Brief description of amendments: These amendments revised TS 
3.3.5.1, ``Emergency Core Cooling System Instrumentation,'' by deleting 
Function 3e, thus preventing the automatic swap of the suction source 
for the high pressure coolant injection pump from the condensate 
storage tank to the suppression pool on high suppression pool level.
    Date of issuance: August 5, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of their associated plant modifications, and no later 
than December 31, 2002.
    Amendment Nos.: 204/178.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50471). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 5, 2002.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket No. 50-498, South Texas Project, 
Unit 1, Matagorda County, Texas

    Date of amendment request: January 28, 2002, as supplemented by 
letters dated June 20 and July 3, and 30, 2002. The supplemental 
information provided clarification that did not change the scope or the 
initial no significant hazards consideration determination.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.4.5.3a, ``Steam Generator Surveillance 
Requirements.'' Specifically, the changes would revise the South Texas 
Project Unit 1 TS to a 40-month inspection interval after its first 
(post-replacement) inservice inspection rather than two consecutive 
inspection resulting in C-1 classification.
    Date of issuance: July 31, 2002.
    Effective date: July 31, 2002.

[[Page 53997]]

    Amendment No.: 140.
    Facility Operating License No. NPF-76: The amendment revises the 
Technical Specification.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12607). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 31, 2002.
    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: October 5, 2001, as supplemented 
on November 7 and 8, 2001, and January 23 and April 30, 2002.
    Brief description of amendment: The amendment revises the license 
and technical specifications to reflect changes related to the transfer 
of the license for the Vermont Yankee Nuclear Power Station, previously 
held by Vermont Yankee Nuclear Power Corporation, to Entergy Nuclear 
Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.
    Date of Issuance: July 31, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 208.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 7, 2001 (66 FR 
63566). The letters dated January 23 and April 30, 2002, provided 
clarifying information and did not expand the application beyond the 
scope of the notice or affect the applicability of the Commission's 
generic no significant hazards consideration determination pursuant to 
10 CFR 2.1315.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 21, 2002.
    Brief description of amendment: The amendment revises several of 
the Required Actions in the technical specifications that require 
suspension of operations involving positive reactivity additions or 
suspension of operations involving reactor coolant system (RCS) boron 
concentration reductions. In addition, the proposed amendment revises 
several Limiting Condition for Operation (LCO) Notes that preclude 
reductions in RCS boron concentration. This amendment revises these 
Required Actions and LCO Notes to allow small, controlled, safe 
insertions of positive reactivity, but limits the introduction of 
positive reactivity such that compliance with the required shutdown 
margin or refueling boron concentration limits will still be satisfied. 
This amendment is based on an NRC-approved traveler, Technical 
Specification Task Force (TSTF)-286, Revision 2.
    Date of issuance: July 29, 2002.
    Effective date: July 29, 2002, and shall be implemented within 60 
days of the date of issuance.
    Amendment No.: 145.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2002 (67 FR 
18650). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated: July 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 7, 2001, as supplemented by 
letter dated February 20, 2002.
    Brief description of amendment: The amendment revises Limiting 
Condition for Operation 3.9.4 to allow the equipment hatch to be open 
during core alterations or movement of irradiated fuel assemblies 
inside containment, and adds the requirement to verify the capability 
to install the equipment hatch in a new Surveillance Requirement 
3.9.4.2. The existing SR 3.9.4.2 would be renumbered SR 3.9.4.3, but 
would otherwise not be changed.
    Date of issuance: July 30, 2002.
    Effective date: July 30, 2002, and shall be implemented within 6 
months of the date of issuance, including the incorporation of changes 
to the Technical Specification Bases as described in licensee's 
application dated August 7, 2001, and supplemental letter dated 
February 20, 2002.
    Amendment No.: 146.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46482). The supplemental letter dated February 20, 2002, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2002.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendment to Facility Operating License and Final 
No Significant Hazards Consideration Determination

    During the period since publication of the last biweekly notice, 
individual notices of issuance of amendments have been issued for the 
facilities as listed below. These notices were previously published as 
separate individual notices. They are repeated here because this 
biweekly notice lists all amendments that have been issued for which 
the Commission has made a final determination that an amendment 
involves no significant hazards consideration.
    In this case, a prior Notice of Consideration of Issuance of 
Amendment, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing was issued, a hearing was requested, and 
the amendment was issued before any hearing because the Commission made 
a final determination that the amendment involves no significant 
hazards consideration.
    Details are contained in the individual notice as cited.

Entergy Nuclear Indian Point 2, LLC, et al., Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: July 13, 2001, as supplemented 
November 30, 2001, March 13, April 3, May 30, and June 13, 2002.
    Brief description of amendment: The amendment made a one-time only 
change to the Technical Specification Surveillance Requirement 4.4.A.3 
to revise the frequency for the containment integrated leak rate test 
(ILRT, Type A test) from at least once per 10 years to once per 15 
years. This change applies only to the interval following the last Type 
A test that was performed satisfactorily in June 1991 at IP2.
    Date of issuance: August 5, 2002.
    Amendment No.: 232.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Facility Operating License No. DPR-26: Amendment revise the 
technical specifications.

[[Page 53998]]

    Date of individual notice in Federal Register: August 22, 2001 (66 
FR 44165). The November 30, 2001, March 13, April 3, May 30, and June 
13, 2002, letters provided clarifying information that did not expand 
the application beyond the scope of the initial notice or change the 
initial proposed no significant hazards consideration determination.

    Dated at Rockville, Maryland, this 9th day of August 2002.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-20843 Filed 8-19-02; 8:45 am]
BILLING CODE 7590-01-P