[Federal Register Volume 66, Number 192 (Wednesday, October 3, 2001)]
[Notices]
[Pages 50463-50480]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-24580]


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NUCLEAR REGULATORY COMMISSION


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

    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.

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 September 10, 2001 through September 21, 
2001. The last biweekly notice was published on September 19, 2001 (66 
FR 48283).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, 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 November 2, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for

[[Page 50464]]

Domestic Licensing Proceedings'' in 10 CFR part 2. Interested persons 
should consult a current copy of 10 CFR 2.714, which is available at 
the NRC's Public Document Room, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available records will be accessible electronically from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.html. If a request for a hearing or petition for leave 
to intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment 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 Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.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, 304-
415-4737 or by e-mail to [email protected].

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

    Date of amendment request: June 21, 2001 (U-603490)
    Description of amendment request: The proposed amendment would 
eliminate the Technical Specification leakage limit for any one main 
steam line as measured by main steam isolation valve leakage of less 
than or equal to 28 standard cubic feet per hour (scfh) and replace 
that requirement with an aggregate leakage limit of less than or equal 
to 112 scfh for all four main steam lines.
    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 MSIVs [main steam isolation valves] are not initiators of or 
precursors to any of the accident scenarios presented in the Updated 
Safety Analysis Report. Therefore, this change does not involve an 
increase in the probability of any accident previously evaluated.
    The proposed change to the TS [Technical Specifications] 
modifies the allowed main steam line leakage limit to an aggregate 
value (i.e., leakage for all four main steam lines combined) with no 
change to the currently allowed total leakage rate. This is the 
value currently used for calculation of dose consequences for the 
bounding accident for which MSIV closure is credited, the large-
break loss of coolant accident (LOCA). This proposed change does not 
impact or increase

[[Page 50465]]

the assumed radionuclide source term therefore; this change does not 
involve an increase in consequences of any accident previously 
evaluated.
    In summary, the proposed change does 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 change does not involve a change to the plant 
design or operation. No new equipment will be installed or utilized, 
and no new operating conditions will be initiated as a result of 
this change. The safety function of the MSIVs is to provide timely 
steam line isolation to mitigate the release of radioactive steam 
and limit reactor inventory loss under certain accident and 
transient conditions. Changing the leakage limits to include an 
aggregate value does not affect the isolation function performed by 
the MSIVs. Therefore, the proposed change does 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 total allowed leakage rate for all four main steam lines 
remains unchanged at 112 scfh. The proposed change does 
not challenge the integrity of the fuel cladding, reactor coolant 
pressure boundary, or the primary containment.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage. As stated above, the total allowed leakage rate for all 
four main steam lines remains unchanged. In addition, there will not 
be a change in the types or amounts of any effluents released 
offsite. The radiological analyses remain unchanged and within the 
guidelines of 10 CFR 100, ``Reactor Site Criteria,'' and 10 CFR 50, 
``Domestic Licensing of Production and Utilization Facilities,'' 
Appendix A, ``General Design Criteria for Nuclear Power Plants,'' 
General Design Criterion 19, ``Control Room.''
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900, 
Downers Grove, IL 60515.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 21, 2001 (U-603495).
    Description of amendment request: The proposed amendment would 
modify the Technical Specification requirement that the main steam line 
safety relief valves (SRVs) open when they are manually actuated by 
instead requiring that the SRV valve actuators stroke on a manual 
actuation.
    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 proposed changes modify TS [Technical Specification] SR 
[surveillance requirements] 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1. 
The proposed changes will eliminate the TS requirement that each 
valve opens during the manual actuation of the SRVs [safety relief 
valves]. Accidents are initiated by the malfunction of plant 
equipment, or the catastrophic failure of plant structures, systems 
or components. The performance of SRV testing is not a precursor to 
any accident previously evaluated and does not change the manner in 
which the SRVs are operated. The proposed testing requirements will 
not contribute to the failure of the SRVs nor any plant structure, 
system or component. Thus, the proposed changes to the performance 
of SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1 do not have any affect on 
the probability of an accident previously evaluated.
    The performance of SRV testing provides assurance that the SRVs 
are capable of depressurizing the reactor pressure vessel (RPV). 
This will protect the reactor vessel from overpressurization and 
allowing the combination of the Low Pressure Coolant Injection 
(LPCI) System and Low Pressure Core Spray (LPCS) System to inject 
into the RPV as designed. The LLS [low-low set] logic causes two LLS 
valves to be opened at a lower pressure than the relief or safety 
mode pressure setpoints and causes all the LLS valves to stay open 
longer, such that reopening of more than one SRV is prevented on 
subsequent actuations. Thus, the LLS function prevents excessive 
short duration SRV cycles with valve actuation at the relief 
setpoint. The proposed changes involve the manner in which the 
subject valves are tested, and have no affect on the types or 
amounts of radiation released or the predicted offsite doses in the 
event of an accident. The proposed testing requirements are 
sufficient to provide confidence that the SRVs, ADS valves and the 
LLS valves will perform their intended safety functions. 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 SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1 
do not affect the assumed accident performance of the SRVs, nor any 
plant structure, system or component previously evaluated. The 
proposed changes do not install any new equipment, and installed 
equipment is not being operated in a new or different manner. The 
valves continue to be bench-tested to verify the safety and relief 
modes of valve operation. The proposed changes will allow the 
testing of the manual actuation electrical circuitry, solenoid and 
air control valve, and the actuator without causing the SRV to open. 
No setpoints are being changed which would alter the dynamic 
response of plant equipment. Administrative controls, such as 
verifying that the actuator assembly has been recoupled following 
testing, minimize the potential for valve failures. Accordingly, no 
new failure modes are introduced. The changes credit the performance 
of bench testing, setpoint verification and in-situ actuator 
exercising with providing sufficient testing to ensure the valves 
will perform their required safety functions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1 
will allow the uncoupling of the SRV stem from the other components 
associated with the manual actuation of the SRVs. The proposed 
changes will allow the testing of the manual actuation electrical 
circuitry, solenoid and air control valve, and the actuator without 
causing the SRV to open. The SRVs will continue to be manually 
actuated by the bench-test valve control system of the setpoint 
testing program and prior to installation in the plant. The proposed 
changes do not effect the valve setpoint or the operational criteria 
that directs the SRVs to be manually opened during plant transients. 
There are no changes proposed which alter the setpoints at which 
protective actions are initiated, and there is no change to the 
operability requirements for equipment assumed to operate for 
accident mitigation.
    Thus, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Robert Helfrich, Mid-West Regional 
Operating Group, Exelon Generation Company, LLC, 1400 Opus Place, 
Suite 900, Downers Grove, IL 60515.
    NRC Section Chief: Anthony J. Mendiola.

[[Page 50466]]

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station, 
Unit 3, Oconee County, South Carolina

    Date of amendment request: March 5, 2001, supplemented September 
4, 2001
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications for Unit 3 to allow a one-time 
extension of the 10 CFR Part 50, Appendix J Containment Integrated 
Leak Rate Test interval. Presently the 10-year interval test is 
required to be performed prior to the operating cycle before the 
outage when the steam generators will be replaced. The proposed 
amendment would extend the test approximately 16 months to the 
outage when they will be replaced (i.e., no later than April 11, 
2005), thereby precluding the need to perform the test during two 
subsequent outages. The No Significant Hazards Consideration 
Determination contained in the March 5, 2001, submittal was 
superceded in the September 4, 2001, submittal and is presented 
below.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No
    The proposed revision to the Oconee Nuclear Station, Unit 3 (ONS-3) 
Technical Specifications (TS) adds a one-time extension to the current 
interval for Type A testing (containment Integrated Leak Rate Testing 
(ILRT)). The current test interval of 10 years, would be extended on a 
one time basis to 12 years 7 months from the last Type A test. The 
proposed extension to Type A testing cannot increase the probability of 
an accident previously evaluated since the containment Type A testing 
extension is not a modification to plant systems, or a change to plant 
operation that could initiate an accident. The proposed extension to 
Type A testing does not involve a significant increase in the 
consequences of an accident since research documented in NUREG-1493 
found that, generically, very few potential containment leakage paths 
fail to be identified by Type B and C tests. In fact, an analysis of 
144 ILRT results, including 23 failures, found that no failures were 
due to containment liner breach. The NUREG concluded that reducing the 
Type A testing frequency to one per twenty years would lead to an 
imperceptible increase in risk. The NUREG conclusions are supported by 
an ONS-3 specific evaluation of risk and consequences. ONS-3 provides a 
high degree of assurance through testing and inspection that the 
containment will not degrade in a manner detectable only by Type A 
testing. Inspections required by the Maintenance Rule and American 
Society of Mechanical Engineers (ASME) code are performed in order to 
identify indications of containment degradation that could affect leak 
tightness. Type B and C testing required by the ONS-3 TS will identify 
any containment opening, such as valves, that would otherwise be 
detected by the Type A tests. Type B and C testing is performed at the 
frequency specified by 10 CFR 50, Appendix J, Option A, Sec. D.2 and 
Sec. D.3, respectively. These factors show that a ONS-3 Type A test 
extension will not represent a significant increase in the consequences 
of an accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The proposed extension to Type A testing cannot create the 
possibility of a new or different type of accident since there are no 
physical changes [b]eing made to the plant. There are no changes to the 
operation of the plant that could introduce a new failure mode creating 
the possibility of a new or different kind of accident.
    3. Do the proposed changes involve a significant reduction in the 
margin of safety?
    Response: No
    The proposed extension to Type A testing will not significantly 
reduce the margin of safety. The NUREG-1493 generic study of the 
effects of extending containment leakage testing found that a 20 year 
extension in Type A leakage testing resulted in an imperceptible 
increase in risk to the public. NUREG-1493 found that, generically, the 
design containment leakage rate contributes a very small amount to the 
individual risk, and that the decrease in Type A testing frequency 
would have a minimal affect on this risk since most potential leakage 
paths are detected by Type C testing. The NUREG conclusions are 
supported by an ONS-3 specific evaluation of risk and consequences.

    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: Anne W. Cottington, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard L. Emch, Jr.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: April 24, 2001.
    Description of amendment request: The proposed amendment would 
revise the Indian Point 3 (IP3) Final Safety Analysis Report (FSAR) 
to reflect the original plant design. It will indicate that a 
portion of one loop of the Component Cooling Water (CCW) System is 
routed in the non-safety-related portion of the Fuel Storage 
Building (FSB).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the Indian Point 3 plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the FSAR revises it to reflect the as built 
configuration of the Component Cooling Water (CCW) system. One loop is 
routed to the Spent Fuel Pit Heat Exchanger (SFPHX) located in the Fuel 
Storage Building (FSB). The portion of the FSB where the CCW is located 
is seismic Class III rather than the seismic Class I required by design 
criteria in the FSAR. The proposed change demonstrates that the CCW 
loop and SFPHX will not be affected by a seismic event and that 
operator, action with credit for the Primary Water Storage Tank (PWST) 
providing redundancy (a source of water to maintain CCW), will assure 
that the CCW system function can be performed following a tornado. The 
proposed change does not affect the probability of an accident 
previously evaluated because there is no design change and the 
probability of natural phenomena does not change. The proposed change 
does not affect the consequences of an accident previously evaluated 
because the CCW system function is maintained by operator action 
following a tornado with missile damage to a small bore CCW pipe.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to the FSAR revises it to reflect the as built 
configuration of the CCW system. One loop is routed to the Spent Fuel 
Pit Heat Exchanger (SFPHX) located in the Fuel Storage Building (FSB). 
The portion of the FSB where the CCW is located is seismic Class III 
rather than the seismic Class I required by design criteria in the 
FSAR. The proposed change demonstrates that the CCW loop and SFPHX will 
not be affected by a seismic event and that operator action with credit 
for the PWST inventory (a source of water to maintain CCW) will assure

[[Page 50467]]

that the CCW system function can be performed following a tornado. The 
proposed change does not create the possibility of a new or different 
kind of accident because the CCW system will not be operated 
differently than designed and operator action and the use of components 
to perform redundant functions to cope with a tornado is currently 
approved in the FSAR. Also, the CCW is designed to be operated by 
separating the loops.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the FSAR revises it to reflect the as built 
configuration of the CCW system. One loop is routed to the Spent Fuel 
Pit Heat Exchanger (SFPHX) located in the Fuel Storage Building (FSB). 
The portion of the FSB where the CCW is located is seismic Class III 
rather than the seismic Class I required by design criteria in the 
FSAR. The proposed change demonstrates that the CCW loop and SFPHX will 
not be affected by a seismic event and that operator action with credit 
for the PWST inventory (a source of water to maintain CCW) will assure 
that the CCW system function can be performed following a tornado. The 
proposed change does not involve a significant reduction in a margin of 
safety because the existing plant design considers the use of operator 
action and redundant components to mitigate the effects of a tornado. 
Also, the proposed change is for damage caused by a low risk event. The 
risk of tornado damage to the CCW piping in the FSB is low. The IP3 
examination of external events found the probability of any tornado 
striking IP3 to be 1.59E-4/year. For tornados with wind speeds in 
excess of 180 mph, the frequency decreases to 8.62E-7/year. For the 
design basis tornado with a 300 mph wind speed, the frequency is 1.02E-
9/year. The risk of a tornado following a LOCA is lower. The frequency 
of a LOCA followed by any tornado within 30 days is 3.02E-8/year. The 
frequency of the event can be used as a conservative estimate of core 
damage frequency (CDF). When compared to the nominal CDF at IP3, the 
frequency is negligible.
    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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Generating Station, 600 Rocky Hill Road, Plymouth, MA 
02360.
    NRC Section Chief: Peter Tam, Acting.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois; Docket 
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power 
Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-373 and 
50-374, LaSalle County Station, Units 1 and 2, LaSalle County, 
Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power 
Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: March 23, 2001.
    Description of amendment request: The proposed amendment would 
incorporate changes to the Physical Security Plan and Guard Force 
Training and Qualification Plans for the identified facilities. The 
proposed changes would modify current escorting and control 
requirements for non-designated vehicles, lighting requirements for 
exterior areas within the protected area, and annual weapons 
qualifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    No physical plant changes are being made as a result of changing 
the vehicle, lighting, and weapons qualification requirements. The 
proposed changes involve revising requirements that provide little 
or no value in the protection of the facility with regards to the 
design basis threat as described in 10 CFR part 73, ``Physical 
Protection of Plants and Materials,'' paragraph 1(a). Because the 
defensive strategies at each station have been proven to be 
effective without reliance on these requirements, it is concluded 
that the proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no physical changes being made to the plant as a 
result of changing the vehicle, lighting, and weapons qualification 
requirements. The defensive strategies at each station remain 
unchanged under the proposed changes. A review of possible intrusion 
scenarios has confirmed that no event would result in a new sequence 
of events that could lead to a new accident scenario. Based on this 
review, it is concluded that no accident scenarios, failure 
mechanisms or limiting single failures are introduced as a result of 
the proposed changes. Therefore, the proposed Physical Security Plan 
and Guard Force Training and Qualification Plan changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    C. The proposed changes do not involve a significant reduction 
in the margin of safety.
    It has been shown during recent Operational Safeguards Readiness 
Evaluations (OSRE), that the proposed changes do not impact the 
security's ability to protect the facility from the threat of 
radiological sabotage. The risk of radiological sabotage would not 
be increased by changing the vehicle, lighting, and weapons 
qualification requirements. Additionally, proposed change in weapons 
qualifications provides a more realistic evaluation of a responder's 
ability to protect the station from the threat of radiological 
sabotage. Based on this review, the proposed amendment does not 
involve a significant reduction in the margin of safety.

    Therefore, based upon the above evaluation, Exelon Generation 
Company, LLC has concluded that these changes do not involve 
significant hazards considerations.

    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, Jr., Vice President 
and General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
KSB 3-W, Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver 
County, Pennsylvania

    Date of amendment requests: May 22, 2001.
    Description of amendment requests: The proposed amendments would 
revise the BVPS-1 and 2 technical specifications (TSs) to implement 
improvements endorsed in the Nuclear Regulatory Commission's (NRC's) 
Final Policy Statement on Technical Specification Improvements for 
Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). These 
license amendment requests propose the addition of an administrative 
control program for explosive gas and storage tank radioactivity 
monitoring to the administrative controls section of the BVPS-1 and 2 
TSs consistent with the corresponding standard TS program. The 
amendment requests propose to

[[Page 50468]]

relocate the TS requirements associated with the curie content limit 
for liquid and gaseous waste storage system and the explosive gas 
concentration limit for gaseous waste storage systems. The addition of 
the standard TS program provides an appropriate level of control for 
the affected requirements in the TSs that allows these details to be 
relocated. The TSs proposed for relocation will be placed in the BVPS-1 
and 2 Offsite Dose Calculation Manual or the BVPS-1 and 2 Licensing 
Requirements Manual. The net effect of the proposed changes is to 
provide adequate regulatory control in the TSs while making the content 
of the BVPS-1 and 2 TSs more consistent with the standard TSs for 
Westinghouse plants as presented in NUREG-1431 and simplifying the 
BVPS-1 and 2 TSs consistent with the goals of the NRC Final Policy 
Statement on TS improvements for nuclear power reactors.
    Additionally, revisions to BVPS-1 and 2 TS 6.9.3, ``Annual 
Radioactive Release Report,'' are proposed to include changes to the 
reporting requirements. The changes proposed also include various 
administrative revisions to support the relocations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because no 
changes are being made to any event initiator. Nor is any analyzed 
accident scenario being revised. The initiating conditions and 
assumptions for accidents described in the UFSAR [Updated Final 
Safety Analysis Report] remain as previously analyzed.
    The proposed amendment also does not involve a significant 
increase in the consequences of an accident previously evaluated. 
The amendment does not reduce the current operability requirements 
contained in the TS proposed for relocation. The proposed relocation 
of TS requirements only affects the level of regulatory control 
involved in future changes to the requirements. The proposed changes 
include additions to the TS in the form of programmatic controls 
that effectively replace the key TS requirements being relocated. As 
such, the TS proposed for relocation no longer meet the 10 CFR 50.36 
criteria for retention in the TS.
    The additional administrative changes are editorial in nature, 
and are made to support the relocation of TS. The additional 
administrative changes and the changes to Specification 6.9.3 have 
no adverse effect on the safety analyses for design basis accidents 
described in the UFSAR. The initiating conditions and assumptions 
for accidents described in the UFSAR remain as previously analyzed.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment does not involve any physical changes to 
the plant or the modes of plant operation defined in the TS. The 
proposed amendment does not involve the addition or modification of 
plant equipment nor does it alter the design or operation of any 
plant systems. No new accident scenarios, transient precursors, 
failure mechanisms, or limiting single failures are introduced as a 
result of these changes.
    There are no changes in this amendment that would cause the 
malfunction of safety-related equipment assumed to be operable in 
accident analyses. No new mode of failure has been created and no 
new equipment performance requirements are imposed. The proposed 
amendment has no effect on any previously evaluated accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters and systems within design requirements and 
safety analysis assumptions.
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this amendment 
revise any parameters or operating restrictions that are assumptions 
of a design basis accident. In addition, the proposed amendment does 
not affect the ability of safety systems to ensure that the facility 
can be placed and maintained in a shutdown condition for extended 
periods of time.
    The proposed change includes the addition of programmatic 
controls that allow the affected TS to be relocated. The relocation 
of TS does not reduce the effectiveness of the requirements being 
relocated. Rather, the relocation of the TS results in a change in 
the regulatory control required for future changes made to the 
requirements. Additionally, due to the new programmatic controls, 
the TS proposed for relocation no longer meet the 10 CFR 50.36 
criteria for retention in the TS.
    The requirements contained within the affected TS will continue 
to be implemented by the appropriate plant procedures (e.g., 
operating and maintenance procedures) in the same manner as before. 
However, future changes to the relocated requirements will be 
controlled in accordance with 10 CFR 50.59 instead of a license 
amendment pursuant to 10 CFR 50.90. The provisions of 10 CFR 50.59 
establish adequate controls over requirements removed from the TS 
and assure future changes to these requirements will be consistent 
with safe plant operation.
    The additional administrative changes are editorial in nature, 
and are made to support the relocation of TS. The additional 
administrative changes and the proposed changes to Specification 
6.9.3 do not alter any operating parameters or design requirements 
assumed in a safety analysis for systems or components important to 
the mitigation and control of design bases accident conditions 
within the facility. Nor do these changes alter safety limits or 
safety system settings required for safe operation of the plant or 
the assumptions of any safety analysis.
    Therefore, the proposed amendment 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Peter Tam (Acting).

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County, 
Pennsylvania

    Date of amendment request: July 25, 2001.
    Description of amendment request: The proposed license amendment 
for BVPS-2 would increase the limits for boron concentration in the 
refueling water storage tank (RWST) and in the reactor coolant system 
(RCS) accumulators. The RCS minimum boron concentration limit for Mode 
6 would also be revised to make it consistent with the RWST boron 
concentration limit. The increase in the boron concentration limits in 
the RWST and accumulators is needed to address higher reactor core 
reactivity levels associated with core operation at higher plant 
capacity factors. TS Bases changes are also proposed to reflect the 
changes discussed above.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 50469]]

    No. The proposed change to increase the boron concentration in 
the Beaver Valley Power Station (BVPS) Unit 2 Refueling Water 
Storage Tank (RWST), Accumulators and in the Reactor Coolant System 
(RCS) during Mode 6 will maintain the safety analyses results in 
Chapter 15 of the BVPS Unit 2 Updated Final Safety Analysis Report 
(UFSAR) as bounding values for all Loss Of Coolant Accident (LOCA) 
and non-LOCA design basis accidents. The proposed changes do not 
reduce the RWST or accumulators ability to meet their design bases, 
which will not result in a significant increase in the probability 
of an accident previously evaluated.
    Increased boron concentration limits for the RWST, Accumulators, 
and RCS in Mode 6 will not increase the consequences of an accident 
previously analyzed as described in the UFSAR. The increased boron 
concentration limits reduce the time to switchover from cold leg to 
hot leg recirculation, which will prevent boron precipitation in the 
reactor vessel following a LOCA. The post-LOCA long term core 
cooling minimum boron requirements have been determined to continue 
to be adequate to ensure adequate post-LOCA shutdown margin. The 
post-LOCA containment sump and containment spray pH remain within 
the limits specified in the UFSAR. All other transients either were 
not impacted or were made less severe as a result of the increased 
boron concentrations.
    Therefore, based upon the above, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed increase in boron concentration does not add 
new or different equipment to the facility. The proposed Technical 
Specification changes also do not alter the manner in which plant 
equipment is being operated. Although the increased boron 
concentration requires procedure changes to ensure that cold leg to 
hot leg recirculation after a LOCA occurs quicker, there are no 
changes to the methods utilized to respond to plant events. The 
proposed Technical Specification changes do not alter instrument or 
control setpoints that initiate protective or mitigative actions. 
These increased boron concentration limits are conservative and do 
not alter the RCS or Emergency Core Cooling Systems' ability to 
perform their design bases.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any previously evaluated 
accident since the RCS will continue to operate in accordance with 
their design bases.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The LOCA considerations, including Peak Cladding Temperature 
calculations, containment sump and spray pH requirements, boron 
solubility requirements, cold shutdown boration requirements, post-
LOCA long term core cooling minimum boron requirements, hot leg 
recirculation switchover requirements, post-LOCA hydrogen generation 
requirements, and radiological requirements have been evaluated and 
determined to be acceptable. The acceptance criteria of all non-LOCA 
design basis accidents continue to be met.
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not adversely affect the ability 
of systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling condition for extended periods of time.
    Based upon the above evaluations, [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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Peter Tam, Acting.

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

    Date of amendment request: August 22, 2001.
    Description of amendment request: The proposed amendments revise 
Section 6.0, ``Administrative Controls,'' of the Technical 
Specifications (TS) to change the title of the corporate executive 
responsible for overall nuclear safety from ``President-Nuclear 
Division'' to ``Chief Nuclear Officer.'' The proposed changes eliminate 
the reference to a specific organizational title and replace it with a 
generic organizational position title. This conforms the TS to a recent 
organizational change and precludes the need for future amendments in 
response to future corporate title changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature, changing 
the title of the corporate executive responsible for overall plant 
nuclear safety in St. Lucie Units 1 and 2 TS, and would not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated. These amendments do not affect 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant, nor do they affect TS that preserve 
safety analysis assumptions. Therefore, the proposed changes do not 
affect the probability or consequences of accidents previously 
analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to the TS are administrative in nature, 
changing the title of the corporate executive responsible for 
overall plant nuclear safety in St. Lucie Units 1 and 2 TS, and 
would not create the possibility of a new or different kind of 
accident from any previously evaluated. The proposed amendments will 
not change the physical plant or the modes of plant operation 
defined in the facility operating license. No new failure mode is 
introduced due to the administrative changes since the proposed 
changes do not involve the addition or modification of equipment, 
nor do they alter the design or operation of affected plant systems, 
structures, or components. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes are administrative in nature, changing the 
title of the corporate executive responsible for overall plant 
nuclear safety in the St. Lucie Units 1 and 2 TS, and would not 
reduce any of the margins of safety.The operating limits and 
functional capabilities of the affected systems, structures, and 
components remain unchanged by the proposed amendments. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: August 15, 2001.

[[Page 50470]]

    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to extend the channel 
calibration surveillance frequency for the automatic depressurization 
system (ADS) timers from 18 months to 24 months. Specifically, SR 
3.3.5.1.7 (18-month CHANNEL CALIBRATION) surveillance requirement 
listed in Table 3.3.5.1-1, functions 4.b. and 5.b. (ADS Timer), would 
be changed to SR 3.3.5.1.8 (24-month CHANNEL CALIBRATION.) This channel 
calibration surveillance would continue to be performed in the same 
manner but at a reduced frequency. No modifications to test 
methodologies or station equipment have been proposed in this request. 
This request is made to facilitate a change to the Duane Arnold Energy 
Center operating cycle from 18 months to 24 months. This request has 
been prepared following the guidance in Generic Letter 91-04, ``Changes 
in Technical Specification Surveillance Intervals to Accommodate a 24-
Month Fuel Cycle.''
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment extends the CHANNEL CALIBRATION 
surveillance frequency for the ADS timers from 18 months to 24 
months to facilitate a change in the DAEC operating cycle from 18 
months to 24 months. The proposed change does not physically impact 
the plant nor does it impact any design or functional requirements 
of the ADS. That is, the proposed change does not degrade the 
performance or increase the challenges of any safety systems assumed 
to function in the accident analysis. The proposed change alters the 
frequency but not the Surveillance Requirement itself nor the way in 
which the surveillance is performed. The proposed change does not 
affect the availability of equipment or systems required to mitigate 
the consequences of an accident because of the availability of 
redundant systems or equipment and because other tests performed 
more frequently will identify potential equipment problems. 
Furthermore, an evaluation of surveillance test results shows that 
the probability of exceeding the TS Allowable Value (AV) with the 
extended surveillance frequency is small and remains well within the 
setpoint methodology guideline. Therefore, the proposed amendment 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment extends the CHANNEL CALIBRATION 
surveillance frequency for the ADS timers from 18 months to 24 
months to facilitate a change in the DAEC operating cycle from 18 
months to 24 months. The proposed change does not introduce any 
failure mechanisms of a different type than those previously 
evaluated since there are no physical changes being made to the 
facility. In addition, only the frequency will change; the 
Surveillance Requirement itself and the way the surveillance is 
performed will remain unchanged. Furthermore, a review of the 
maintenance history of these timers indicated no evidence of any 
failures that would invalidate the above conclusions. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed amendment extends the CHANNEL CALIBRATION 
surveillance frequency for the ADS timers from 18 months to 24 
months to facilitate a change in the DAEC operating cycle from 18 
months to 24 months. Although the proposed change will result in an 
increase in the interval between surveillance tests, the impact on 
system availability is considered small based on other more frequent 
testing, the availability of redundant systems or equipment, and the 
fact that there is no evidence of any existing equipment failures 
that would impact the availability of the ADS. Furthermore, an 
evaluation of surveillance test results shows that the probability 
of exceeding the TS AV with the extended surveillance frequency is 
small and remains well within the setpoint methodology guideline. 
Therefore, the proposed amendment will 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: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: August 30, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) safety limit minimum critical 
power ratio (MCPR) for two recirculation pump operation for Cycle 21.
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed Safety Limit MCPR (SLMCPR), and its use to 
determine the Cycle 21 thermal limits, have been derived using NRC 
approved methods and uncertainties. These methods do not change 
operation of the plant, and have no effect on the probability of an 
accident initiating event or transient. The basis of the SLMCPR is 
to ensure no mechanistic fuel damage is calculated to occur if the 
limit is not violated. The new SLMCPR for Cycle 21 preserves the 
margin to transition boiling and the probability of fuel damage is 
not increased.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change results only from different inputs, for the 
Cycle 21 core reload. These methods and uncertainties have been 
reviewed and approved by the NRC, and do not involve any new or 
unapproved methods for operating the facility. No new initiating 
events or transients result from these changes.
    The SLMCPR remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity. A change in SLMCPR cannot create the possibility of any 
new type of accident. SLMCPR values for the new fuel cycle are 
calculated using previously transmitted methodology.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPR was derived using NRC approved methods and 
uncertainties which are in accordance with the current fuel design 
and licensing criteria. The SLMCPR remains high enough to ensure 
that greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated, thereby preserving 
the fuel cladding integrity.
    Fuel licensing acceptance criteria for SLMCPR calculations apply 
to Monticello Cycle 21 in the same manner as previously applied. 
SLMCPRs prepared using methodology previously transmitted to the NRC 
ensure that greater than 99.9% of all fuel rods in the core will 
avoid transition boiling if the limit is not violated, thereby 
preserving fuel cladding integrity. The

[[Page 50471]]

operating MCPR limit is set appropriately above the safety limit 
value to ensure adequate margin when the cycle specific transients 
are evaluated.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: June 8, 2001.
    Description of amendment request: The proposed amendment would 
delete the ``High Pressure Coolant Injection (HPCI) System Suppression 
Pool Water Level--High'' function from Technical Specification (TS) 
3.3.5.1, ``Emergency Core Cooling System (ECCS) Instrumentation.'' This 
change would eliminate automatic transfer of the HPCI pump suction 
source from the condensate storage tank (CST) to the suppression pool 
for a high suppression pool level. Elimination of this function is 
expected to increase the availability of the HPCI system during a 
postulated anticipated transient without scram (ATWS) with standby 
liquid control system (SLCS) failure and to reduce operator burden 
during a postulated station blackout (SBO) event.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Deletion of the automatic HPCI suction transfer from the CST to 
the suppression pool for a high suppression pool level condition was 
analyzed for impacts against all previously evaluated accidents and 
transients. Eliminating the automatic transfer increases the 
availability of the HPCI system during an ATWS event and operator 
burden is reduced during a postulated Station Blackout (SBO). There 
are no adverse effects, consequences, or changes in the probability 
of an accident occurring as a result of this change. HPCI operation 
is improved and all other plant systems remain unaffected in their 
ability to perform their design basis functions as a result of this 
change.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed?
    Implementation of this change increases the availability of the 
HPCI system during a postulated ATWS with Standby Liquid Control 
System (SLCS) failure. The change only affects the HPCI suction 
source and whether the source is automatically transferred from the 
preferred CST to the suppression pool for a high suppression pool 
level. Continued HPCI operation utilizing the CST as a suction 
source does not create a new or different type of accident from 
those previously analyzed. The primary effect of this change is to 
the suppression pool level which has been evaluated and found to be 
acceptable for all relevant accidents and transients. Therefore a 
new or different accident is not created and all other accident 
analyses are unaffected by the change.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    This change does not reduce any margin of safety. The increase 
in suppression pool water level does not cause containment 
hydrodynamic loads to exceed design limits under accident 
conditions. Overall, HPCI reliability is increased as it would 
remain operable during the ATWS with Loss of SLCS event. This 
increased availability of the HPCI system provides for additional 
defense in depth which reduces the probability of core damage.

    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: Peter Tam, Acting.

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

    Date of amendment request: July 17, 2001.
    Description of amendment request: The proposed amendment would 
revise the reactor pressure vessel (RPV) pressure-temperature (P-T) 
limits specified in the technical specifications (TSs). Editorial 
changes associated with the P-T limit revisions are also proposed. The 
proposed P-T limits rely on the methodology for determining allowable 
P-T limits specified in American Society of Mechanical Engineers (ASME) 
Code Case N-640. The revised P-T limits will allow required RPV 
hydrostatic and leak tests to be performed at a significantly lower 
temperature. This is expected to reduce challenges to plant operators 
associated with maintaining the reactor coolant system within a narrow 
temperature band during testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The changes to the calculational methodology for the pressure 
and temperature (P-T) limits based upon Code Case N-640 continue to 
provide adequate margin in the prevention of a brittle-type fracture 
of the reactor pressure vessel (RPV). The code case was developed 
based upon the knowledge gained through years of industry 
experience. P-T curves developed using the allowances of Code Case 
N-640 indeed yield more operating margin. However, the experience 
gained in the areas of fracture toughness of materials and pre-
existing undetected defects show that some of the existing 
assumptions used for the calculation of P-T limits are unnecessarily 
conservative and unrealistic. Therefore, providing the allowances of 
the subject Code Case in developing the P-T limit curves will 
continue to provide adequate protection against nonductile-type 
fractures of the RPV.
    The evaluation for the Unit 1 and Unit 2 P-T limit curves for 32 
EFPYs was performed using the approved methodologies of 10 CFR 50, 
Appendix G. The curves generated from these methods ensure the P-T 
limits will not be exceeded during any phase of reactor operation. 
Resolution of the current industry issues related to fluence 
calculation methodology requires PPL to limit applicability of the 
curves to May 1, 2005 for Unit 2 and May 1, 2006 for Unit 1. 
Therefore, the probability of occurrence and the consequences of a 
previously analyzed event are not significantly increased. Finally, 
the proposed changes will not affect any other system or piece of 
equipment designed for the prevention or mitigation of previously 
analyzed events. Thus, the probability of occurrence and the 
consequences of any previously analyzed event are not significantly 
increased as the result of the proposed changes.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously analyzed.
    The proposed changes provide more operating margin in the P-T 
limit curves for inservice leakage and hydrostatic pressure testing, 
non-nuclear heatup and cooldown, and criticality, with the benefits 
being primarily realizable during the pressure tests. Operation in 
the ``new'' regions of the newly developed P-T curves has been 
analyzed in accordance with the provisions of ASME Code, Section XI, 
Appendix G; 10 CFR 50 Appendix G, and ASME Code Case N-640, thus 
providing adequate protection against a

[[Page 50472]]

nonductile-type fracture of the RPV. These proposed changes do not 
create the possibility of any new or different type of accident. 
Further, they do not result in any new or unanalyzed operation of 
any system or piece of equipment important to safety.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    As mentioned previously, the revised P-T curves provide more 
operating margin and thus, more operational flexibility than the 
current P-T curves. However, the industry experience since the 
inception of the P-T limits in 1974 confirms that some of the 
existing methodologies used to develop P-T curves is unrealistic and 
unnecessarily conservative. Accordingly, ASME Code Case N-640 takes 
advantage of the acquired knowledge by establishing more realistic 
methodologies for the development of P-T curves.
    Use of Code Case N-640 to develop the revised P-T curves 
utilized the KIC fracture toughness curve in lieu of the 
KIA curve as the lower bound for fracture toughness. Use 
of the KIC curve to determine lower bound fracture 
toughness is more technically correct than using the KIA 
curve. P-T curves based on the KIC fracture toughness 
limits enhance overall plant safety by expanding the P-T window in 
the low-temperature operating region. The benefits which occur are a 
reduction in the duration of the pressure test and personnel safety 
while conducting inspections in primary containment with no decrease 
to the margin of safety.
    Therefore, operational flexibility is gained without a reduction 
in the margin of safety to RPV brittle fracture.
    The development of the P-T curves to 32 EFPYs was performed per 
the guidelines of 10 CFR [part] 50, and thus, the margin of safety 
is not reduced as the result of the proposed changes. Resolution of 
the current industry issues related to fluence calculation 
methodology requires PPL to limit applicability of the curves to May 
1, 2005 for Unit 2 and May 1, 2006 for Unit 1.

    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, Inc., 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Peter Tam, Acting.

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

    Date of amendment requests: August 24, 2001.
    Description of amendment requests: The licensee requests an 
amendment to the Technical Specifications for containment leakage rate 
testing as a result of recalculation of peak containment internal 
pressure following certain design-basis accidents. The purpose of the 
change is to make the Technical Specifications appropriately reflect 
up-to-date calculated peak containment pressure. The revised calculated 
peak containment pressure related to the design basis loss-of-coolant 
accident and the revised calculated peak containment pressure for the 
design basis Main Steam Line Break would be lower than the current 
Technical Specification values.
    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 amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change would revise the Operating Licenses for San 
Onofre Nuclear Generating Station Units 2 and 3 to amend Technical 
Specification (TS) 5.5.2.15, ``Containment Leakage Rate Testing 
Program,'' by changing the stated calculated values for peak 
containment internal pressure for the design basis Loss Of Coolant 
Accident (LOCA) and Main Steam Line Break (MSLB) accident. The 
current LOCA value of 55.1 psig would be changed to 45.9 psig and 
the current MSLB value of 56.6 psig would be changed to 56.5 psig.
    The proposed change does not affect the probability of 
occurrence of an accident previously evaluated because it relates 
solely to the consequences of hypothesized accidents given that the 
accident has already occurred.
    The proposed change does not increase the calculated peak 
containment internal pressure for the LOCA and MSLB accidents, and 
thus does not increase their consequences.
    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 relates to two accidents, MSLB and LOCA, 
already evaluated in the Updated Final Safety Analysis Report. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The recalculated peak containment internal pressures for the 
MSLB and LOCA accidents are less than the containment design 
pressure and less than the previously calculated pressures. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: August 16, 2001.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications (TS) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the

[[Page 50473]]

applicability of the following NSHC determination in its application 
dated August 16, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

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

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested changes do not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: August 2, 2001.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications (TSs) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated August 2, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in

[[Page 50474]]

aiding the plant staff in assessing the extent of core damage and 
subsequent offsite radiological dose projections. The system was not 
intended to and does not serve a function for preventing accidents 
and its elimination would not affect the probability of accidents 
previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

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

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested changes do not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of amendment request: January 18, 2001, as supplemented by 
letter dated June 26, 2001.
    Brief description of amendment request: The proposed amendment 
would revise the applicability of the current BVPS-1 heatup/cooldown 
curves from 15 effective full-power years (EFPY) to 14 EFPY. Proposed 
changes to Technical Specification (TS) 3.7.1.1, ``Main Steam Safety 
Valves (MSSVs),'' include revisions of the limiting condition for 
operation and to the title and content of Table 3.7-1 to provide 
consistency with the improved standard TSs, creation of new Actions to 
address inoperable MSSVs, reduction of the power range neutron flux-
high reactor trip setpoint to be consistent with TS Traveler Form--235, 
Revision 1, and changes to the maximum power levels permissible with 
inoperable MSSVs. TS Bases changes are also proposed for consistency.
    Date of publication of individual notice in Federal Register: July 
27, 2001 (66 FR 39212).
    Expiration date of individual notice: August 27, 2001.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental

[[Page 50475]]

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/NRC/ADAMS/index.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 e-mail to [email protected].

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: July 2, 2001.
    Brief description of amendments: The amendments deleted Technical 
Specifications Section 5.5.4, ``Post Accident Sampling,'' for Catawba 
Nuclear Station, Units 1 and 2, and thereby eliminated the requirements 
to have and maintain the post-accident sampling systems.
    Date of issuance: September 11, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: 193 and 185.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41615).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 11, 2001.
    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: July 2, 2001.
    Brief description of amendments: The amendments deleted Technical 
Specifications (TS) Section 5.5.4, ``Post Accident Sampling,'' for 
McGuire Nuclear Station, Units 1 and 2, and thereby eliminated the 
requirements to have and maintain the post-accident sampling systems 
(PASS). The amendments also delete PASS-related License Conditions 
2.C(11)c, ``Post Accident Sampling (II.B.3),'' for Unit 1 and 2.C(10)b, 
``Postaccident Sampling (II.B.3),'' for Unit 2.
    Date of issuance: September 17, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: 199 and 180.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41616).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 17, 2001.
    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: February 28, 2001, supplemented 
June 27, 2001.
    Brief description of amendments: The amendments add new Technical 
Specification 3.3.28 and Bases B 3.3.28 governing the addition of the 
low pressure service water standby pump automatic start circuitry.
    Date of Issuance: September 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
before the end of the Oconee Unit 3 End of Cycle 19 Refueling Outage.
    Amendment Nos.: 319, 319, and 319.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
1517).
    The supplement dated June 27, 2001, provided clarifying information 
that did not change the scope of the February 28, 2001, application nor 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 6, 2001.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 24, 2001, as supplemented by 
letter dated March 22, 2001.
    Brief description of amendment: The amendment changes the limit on 
the Low Power Setpoint, from 20 percent to 10 percent power, as 
specified in Technical Specification (TS) 3.1.3, ``Control Rod 
OPERABILITY,'' TS 3.1.6 ``Control Rod Pattern,'' and TS 3.3.2.1, 
``Control Rod Block Instrumentation.''
    Date of issuance: September 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 118.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15921).
    The supplemental letter dated March 22, 2001, provided additional 
information that did not expand the scope of the application or change 
the staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 2001.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 24, 2001, as supplemented by 
letters dated July 20 and August 7, 2001.
    Brief description of amendment: The amendment request proposes 
changes to the Technical Specifications (TSs) concerning certain 
operational conditions required when conducting core alterations or 
handling irradiated fuel in the primary containment. In addition, the 
licensee proposes to implement administrative controls in accordance 
with draft NUMARC 93-01, ``Industry Guidelines for Monitoring the 
Effectiveness of Maintenance at Nuclear Power Plants,'' Revision 3, 
Section 11.3.6.5, ``Containment--Primary (PWR [pressurized-water 
reactor])/ Secondary (BWR [boiling-water reactor]),'' Revision 3, 
Section 11.3.6, ``Assessment Methods for Shutdown Conditions,'' in lieu 
of License Condition 2.C.(17) and change terms to make them consistent 
with the terminology in other revised TSs.
    Date of issuance: September 14, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 119.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20001).
    The supplemental letters dated July 20 and August 7, 2001, provided

[[Page 50476]]

additional information that did not expand the scope of the application 
or change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 2001.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 24, 2001, as supplemented by 
letters dated July 2, and August 6 and 20, 2001.
    Brief description of amendment: The license amendment request 
consists of changes to the Technical Specifications (TSs) to revise the 
reactor vessel pressure/temperature (P/T or P-T) limits specified in TS 
3.4.11, ``RCS [Reactor Coolant System] Pressure and Temperature (P/T) 
Limits,'' for reactor heat-up. The current RCS P/T Limits in TS Figure 
3.4-11, ``Minimum Temperature Required Vs. RCS Pressure,'' would be 
replaced with recalculated RCS P/T limits based, in part, on an 
alternate methodology. The alternate methodology uses American Society 
of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code 
(Code) Case N-640, ``Alternative Requirement Fracture Toughness for 
Development of P-T Limit Curves for ASME B&PV Code Section XI, Division 
1,'' for alternate reference fracture toughness for reactor vessel 
materials in determining the P/T limits.
    Date of issuance: September 14, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 120.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15920).
    The supplemental letters dated July 2, and August 6 and 20, 2001, 
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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: April 27, 2001.
    Brief description of amendments: The amendments revise Technical 
Specifications (TS) Section 5.5.7, to provide an exception to the 
recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, 
Revision 1, which would allow either a qualified in-place ultrasonic 
volumetric examination over the volume from the inner bore of the 
flywheel to the circle of one-half the outer radius or a surface 
examination (magnetic particle testing and/or liquid penetrant testing) 
of exposed surfaces of the removed flywheel to be conducted at 
approximately 10-year intervals. The proposed change is in accordance 
with the Nuclear Regulatory Commission (NRC) approved Improved Standard 
TS Generic Change Traveler TSTF-237, Revision 1.
    Date of issuance: September 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 123, 123, 118, and 118.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated September 20, 2001.

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

    Date of application for amendments: June 26, 2001.
    Brief description of amendments: To extend the dates specified in 
Operating License Sections 2.C(8) and 3.P, ``Pressure--Temperature 
Limit Curves,'' for Dresden Nuclear Power Station, Units 2 and 3, 
respectively.
    Date of issuance: September 10, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 187 and 182.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Facility Operating License.
    Date of initial notice in Federal Register: The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated September 10, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 21, 2001, as supplemented 
June 28, 2001.
    Brief description of amendment: The amendment revised the Improved 
Technical Specifications (ITS) 5.6.2.10, ``OTSG [Once-Through Steam 
Generator] Tube Surveillance Program'' to implement a reroll process to 
repair degraded steam generator tubes and allow the reroll repairs to 
be used in both the upper and lower tubesheets.
    Date of issuance: September 10, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 198.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20006). The supplemental letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 10, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 3, 2000, as supplemented 
June 14, August 28, and September 7, 2001.
    Brief description of amendment: The amendment revised the Improved 
Technical Specifications (ITS) 3.7.12, ``Control Room Emergency 
Ventilation System (CREVS)''; ITS 5.6.2.12, ``Ventilation Filter 
Testing Program (VFTP)''; ITS 3.3.16, ``Control Room Isolation--High 
Radiation''; and ITS 3.7.18, ``Control Complex Cooling System.'' The 
proposed ITS changes are based on the results of revised public and 
control room dose calculations for CR-3 design basis radiological 
accidents using an alternative source term and the adoption of 
Technical Task Force Traveler (TSTF) 287. A new Section 5.6.2.21, 
``Control Complex Habitability Envelope Program,'' is added.

[[Page 50477]]

    Date of issuance: September 17, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 199.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69060). The supplemental letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 17, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 18, 2001, as supplemented 
August 24, 2001.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 6.9.1.11, ``Core Operating Limits Report (COLR),'' 
to include the ABB-Combustion Engineering Topical Report CENPD-387-P-A, 
Rev 000, in the list of analytical methods. This allows use of an 
improved heat flux correlation (designated ABB-NV) previously approved 
by the NRC. Additionally, the Bases for TS 2.1.1, ``Reactor Core,'' are 
modified to reflect use of the improved heat flux correlation.
    Date of Issuance: September 20, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 118.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29358). The August 24, 2001, supplement did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 20, 2001.
    No significant hazards consideration comments received: No.

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear 
Generating Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: July 25, 2000, as supplemented by letter 
dated June 21, 2001.
    Brief description of amendment request: The amendment changes Three 
Mile Island Nuclear Generating Station, Unit 2 (TMI-2), Technical 
Specification (TS) 6.7.2 to eliminate a change associated with periodic 
reviews of procedures. Currently, TS 6.7.2 states that required 
procedures shall be reviewed periodically as required by American 
National Standards Institute (ANSI) Standard N18.7-1976 (a biennial 
review). This amendment revises the wording for TS 6.7.2 to state that 
required procedures shall be reviewed periodically. This amendment is 
also consistent with the TMI-2 Post-Defueling Monitored Storage Quality 
Assurance Plan, which states that ``Procedural documentation shall be 
periodically reviewed for adequacy as set forth in administrative 
procedures.''
    Date of Issuance: September 7, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 56.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specification.
    Date of initial notice in Federal Register: December 13, 2000 (66 
FR 77920).
    The June 21, 2001, supplemental letter replaced in its entirety the 
original application dated July 25, 2000. The supplement did not expand 
the scope of the original request.
    The Commission's related evaluation is contained in a safety 
evaluation dated September 7, 2001.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: August 17, 2001.
    Brief description of amendment: The amendment allowed a one-time 
exception to Technical Specification Surveillance Requirement (SR) 
3.6.1.7.2 for suppression chamber-to-drywell vacuum breakers 2ISC*RV35A 
and 2ISC*RV35B. A note has been added to SR 3.6.1.7.2 stating that 
function testing of these vacuum breakers is not required to be met for 
the remainder of Cycle 8.
    Date of issuance: September 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 98.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Public Comments Requested as to Proposed No Significant Hazards 
Consideration: Yes (66 FR 44653) August 24, 2001. That notice provided 
an opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. Comments were received 
from one person, and were addressed in the safety evaluation associated 
with the amendment. The notice also provided for an opportunity to 
request a hearing by September 24, 2001, but indicated that if the 
Commission makes a final no significant hazards consideration 
determination, any such hearing would take place after the issuance of 
the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, final determination of no significant hazards 
consideration determination, and state consultation, are contained in a 
safety evaluation dated September 7, 2001.
    Attorney for the Licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Peter S. Tam, Acting.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: April 6, 2001.
    Brief description of amendment: The amendment changes Technical 
Specifications Section 5.5.10, ``Technical Specifications (TS) Bases 
Control Program,'' in accordance with Nuclear Energy Institute TS Task 
Force Standard TS Change Traveler, TSTF-364, ``Revision to TS Bases 
Control Program to Incorporate Changes to 10 CFR 50.59,'' Revision 0.
    Date of issuance: September 13, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41623).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2001.
    No significant hazards consideration comments received: No.

[[Page 50478]]

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

    Date of application for amendment: May 25, 2001, as supplemented 
August 17, 2001.
    Brief description of amendment: The amendment to the Kewaunee 
Nuclear Power Plant Technical Specifications (TSs) 4.2 revises TS 4.2 
to revise the surveillance requirements and bases for TS 4.2.b, ``Steam 
Generator Tubes,'' to account for changes associated with replacement 
of the original steam generators. Specifically, the changes delete 
inspection requirements associated with steam generator tube sleeving 
and repair limits and revise the phrasing of text within the TS to 
enhance clarity.
    Date of issuance: September 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 158.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31711).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated September 20, 2001.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 24, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specification definition of CORE ALTERATIONS.
    Date of issuance: September 11, 2001.
    Effective date: The amendments are effective as of the date of 
their issuance and shall be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: Unit 1--131; Unit 2--120.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36345)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 11, 2001.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendments request: May 24, 2001.
    Brief description of amendments: The amendments relocate Technical 
Specification 3/4.9.6, ``Refueling Machine'' and its associated Bases 
description to the Technical Requirements Manual.
    Date of issuance: September 13, 2001.
    Effective date: The amendments are effective as of the date of 
their issuance.
    Amendment Nos.: Unit 1--132; Unit 2--121.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36344).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 25, 2001.
    Brief description of amendment: The proposed amendment deletes 
Technical Specification (TS)-required Action 3.3.1.1.I.2, which limits 
plant operation to 120 days in the event of the inoperability of the 
Oscillation Power Range Monitor trip system. For this situation, the 
proposed change would allow plant operation to continue if the existing 
TS Required Action 3.3.1.1.I.1, to implement an alternate means to 
detect and suppress thermal hydraulic instability oscillations, was 
taken.
    Date of issuance: September 13, 2001.
    Effective date: Date of issuance and shall be implemented within 30 
days.
    Amendment No.: 231.
    Facility Operating License No. DPR-68: Amendment revises the TS.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41627).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 16, 2000, as supplemented 
September 27, 2000, and June 6, 2001.
    Brief Description of amendments: These amendments change the 
reactor protection system and engineered safety features actuation 
system analog instrumentation surveillance frequency from monthly to 
quarterly.
    Date of issuance: August 31, 2001.
    Effective date: August 31, 2001.
    Amendment Nos.: 228 and 228.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69067). The September 27, 2000, and June 6, 2001, supplements 
contained clarifying information only, and did not change the initial 
no significant hazards consideration determination or expand the scope 
of the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 31, 2001.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal

[[Page 50479]]

Register notice providing opportunity for public comment or has used 
local media to provide notice to the public in the area surrounding a 
licensee's facility of the licensee's application and of the 
Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Assess and Management Systems 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-415-4737 or by Email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By November 2, 2001, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the

[[Page 50480]]

amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: September 14, 2001.
    Brief description of amendment: The amendment authorizes a one-
time-only change to Technical Specifications Section 3.9.B.1 and 
associated Bases. Specifically, this change extends the Limiting 
Condition for Operation allowable out-of-service time for one incoming 
Reserve AC Power line (115KV line #3) and/or one reserve station 
transformer inoperable from 7 days to 14 days during the period 
commencing September 9, 2001 and extending through September 23, 2001.
    Date of issuance: September 15, 2001.
    Effective date: As of the date of issuance.
    Amendment No.: 272.
    Facility Operating License No. DPR-59: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration, are contained in a Safety Evaluation dated 
September 14, 2001.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Peter Tam (Acting)

    Note:
    The publication date for this notice will change from every 
other Wednesday to every other Tuesday, effective January 8, 2002. 
The notice will contain the same information and will continue to be 
published biweekly.


    Dated at Rockville, Maryland, this 25th day of September 2001.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-24580 Filed 10-2-01; 8:45 am]
BILLING CODE 7590-01-P