[Federal Register Volume 66, Number 220 (Wednesday, November 14, 2001)]
[Notices]
[Pages 57116-57131]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-28399]


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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

[[Page 57117]]

make immediately effective any amendment to an operating license upon a 
determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from October 22, 2001 through November 3, 2001. 
The last biweekly notice was published on October 31, 2001 (66 FR 
557007).

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 December 14, 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 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.

[[Page 57118]]

    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 email to [email protected].

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

    Date of amendment request: August 21, 2001.
    Description of amendment request: The proposed amendment would 
revise the actions taken for an inoperable battery charger, revise 
battery charger testing criteria, and relocate certain safety-related 
battery surveillance requirements from the Technical Specifications to 
a licensee-controlled program.
    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 restructure the TS [Technical 
Specifications] for the DC Electrical Power system. The proposed 
changes add actions to specifically address battery charger 
inoperability with increased completion times. This change will rely 
upon the capability of providing the battery charger function by an 
alternate means, (e.g., a spare battery changer that will function 
as a qualified backup) to take advantage of the proposed increased 
completion time. The CD power System or associated battery chargers 
are not initiators to any accident sequence analyzed in the Updated 
Safety Analysis Report (USAR). Operation in accordance with the 
proposed TS ensures that the DC Power System is capable of 
performing function as described in the USAR, therefore the 
mitigative functions supported by the DC Power System will continue 
to provide the protection assumed by the analysis.
    The relocation of preventive maintenance surveillance, and 
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.14, ``Battery Monitoring and Maintenance 
Program,'' will not challenge the ability of the DC Power System to 
perform its design function. The maintenance and monitoring required 
by current TS, which are based on industry standards, will continue 
to be performed. In addition, the DC Power System is within the 
scope of 10 CFR 50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants,'' which will 
ensure the control of maintenance activities associated with the DC 
Power System.
    In summary, 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 involve restructuring the TS for the DC 
Electrical Power system. This change will rely upon the capability 
of providing the battery charger function by an alternate means, 
(e.g., a spare battery charger that will function as a qualified 
backup) to take advantage of the proposed increased completion time. 
The DC Power System or associated battery chargers are not 
initiators to any accident sequence analyzed in the Updated Safety 
Analysis Report (USAR).
    Allowing the use of a spare battery charger will increase the 
reliability of the DC Electrical Power system. The mitigative 
functions supported by the DC Power System will continue to provide 
the protection assumed by the safety analysis described in the USAR. 
Therefore, there are no new types of failures that could be created 
by a failure of the spare battery charger. As such, no new or 
different kind of accident or transient is expected by these 
changes.
    Therefore, these 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 will not adversely affect operation of 
plant equipment. These changes will not result in a change to the 
setpoints at which protective actions are initiated. Sufficient DC 
capacity to support operation of mitigation equipment is ensured. 
The changes associated with the new Battery Maintenance and 
Monitoring Program will ensure that the station batteries are 
maintained in a highly reliable manner. The use of a spare battery 
charger will increased the reliability of the DC system during 
periods of normal battery charger inoperability. The equipment fed 
by the DC Electrical Sources will continue to provide adequate power 
to safety related loads in accordance with analysis assumptions.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: September 20, 2000, as supplemented 
August 2 and September 28, 2001.
    Description of amendment request: The proposed Technical 
Specification (TS) change would (1) delete the requirements for 
hydrogen monitoring instrumentation from TS sections 3.5.5.2, 3.6, and 
Tables 3.5-3 and 4.1-4 and correct a typographical error in item 8 of 
Table 4.1-4; (2) delete the requirements for hydrogen recombiners in TS 
section 4.4.4; and (3) delete the reference to the hydrogen purge 
system and hydrogen recombiners from the Bases of TS section 4.12.2.
    Basis for proposed no significanthazards consideration 
determination: As required by 10 CFR

[[Page 57119]]

50.91(a), the licensee has provided its analysis of the issue of no 
significant hazards consideration. The Nuclear Regulatory Commission 
(NRC) staff reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's analysis, which is based on the 
representation made by the licensee in the September 20, 2001, 
application as supplemented August 2 and September 28, 2001, is 
presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. This change has no effect on plant equipment provided for the 
reactor coolant system, reactor building heat removal, or the equipment 
provided for mixing of the reactor building atmosphere following an 
accident. This proposed change does not alter the design or 
configuration of the plant beyond that of the containment combustible 
gas control systems. The containment combustible gas control systems 
are currently classified as safety systems. The containment combustible 
gas control systems are composed of two hydrogen monitors and two 
hydrogen recombiners, backed up by a portion of the reactor building 
purge system that can be used to vent the reactor building. Hydrogen 
control components (hydrogen monitors, hydrogen recombiners, and 
hydrogen vents) do not affect any accident initiation sequence 
previously identified. Therefore, this change does not increase the 
probability of an accident previously evaluated.
    The containment combustible gas control systems are provided to 
ensure that reactor building hydrogen concentration is maintained below 
the lower flammability limit of 4.0 percent. The NRC staff has found 
hydrogen combustion to be a small contributor to containment failure 
for large, dry containment designs due to the robustness of these 
containment types and the likelihood of a spurious ignition source. The 
containment combustible gas control systems are not credited in the TMI 
Unit 1 probability risk assessment (PRA).
    Therefore, this change would not result in a significant increase 
the consequence of accidents previously evaluated.
    2. Will operation of the facility in accordance with the proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    No. This proposed change does not alter the design or configuration 
of the plant beyond that of the containment combustible gas control 
systems. Hydrogen generation following a design basis loss-of-coolant 
accident (LOCA) has been evaluated in accordance with regulatory 
requirements. Deletion of the containment combustible gas control 
system from the TSs does not alter the hydrogen generation processes 
post-LOCA. The NRC staff has found hydrogen combustion to be a small 
contributor to containment failure for large, dry containment designs 
due to the robustness of these containment types and the likelihood of 
a spurious ignition source. The containment combustible gas control 
systems are not credited in the TMI Unit 1 level 2 PRA.
    Therefore, since the accident evaluation does not credit these 
systems or assume that they operate during an accident, operation of 
the facility in accordance with this proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    No. This change has no effect on plant equipment provided for the 
reactor coolant system, reactor building heat removal, or the equipment 
provided for mixing of the reactor building atmosphere following an 
accident. This change only involves the deletion of requirements for 
containment combustible gas control equipment, (hydrogen monitors, 
hydrogen recombiners, and containment hydrogen vents). The NRC staff 
has found hydrogen combustion to be a small contributor to containment 
failure for large, dry containment designs due to the robustness of 
these containment types and the likelihood of a spurious ignition 
source. Use of the containment combustible gas control systems are not 
credited in the TMI Unit 1 PRA. TMI Unit 1 utilizes a large open 
containment design that precludes the buildup of hydrogen pockets that 
might be formed if the reactor building were of a compartmentalized 
design. The TMI-1 PRA concluded that the containment would remain 
intact for severe accidents which included hydrogen burns for which no 
credit was taken for the combustible gas control system as long as the 
containment heat removal systems (reactor building emergency cooling 
and reactor building sprays) remain functional.
    The proposed change will relax certain special treatment 
requirements associated with hydrogen monitors. As discussed in 
Regulatory Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear 
Power Plants to Assess Plant and Environs Conditions During and 
Following an Accident,'' Revision 3, dated May 1983, the NRC staff 
believes that the revised treatment is appropriate for instrumentation 
needed to assess the degree of core damage and confirm that spurious 
ignition of hydrogen has taken place.
    Therefore, operation of the facility in accordance with this 
proposed change will not involve a significant reduction in a margin of 
safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Vice President and 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, KSB 3-
W, Kennett Square, PA 19348.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: August 27, 2001.
    Description of amendment request: The proposed amendment changes 
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications (TSs) action and surveillance requirements associated 
with the containment air lock. The Bases of the affected TSs will be 
modified to address the proposed 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. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is 
based on the representation made by the licensee in the August 27, 
2001, application, is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will not revise the operability requirements 
for the containment air lock. As a result, the design-basis accidents 
will remain the same postulated events, and the consequences of the 
design-basis accidents will remain the same. Also, the containment air 
lock is not an accident initiator. Therefore, the proposed change will 
not involve any increase in the probability or consequences of an 
accident previously evaluated.

[[Page 57120]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the containment air lock is not an accident initiator, these 
proposed changes do not introduce any new failure modes. Therefore, the 
proposed changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the operability requirements for the containment air lock 
will not change, and the containment air lock will continue to function 
as assumed in the safety analysis, the proposed change will not result 
in a reduction in a margin of safety.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 26, 2001.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications (TSs) to relocate MP3 TSs related to the position 
indication system to the respective Technical Requirements Manual 
(TRM). The Bases of the affected TSs will be modified to address the 
proposed changes. Also, index pages will be revised to reflect the 
relocation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is 
based on the representation made by the licensee in the September 26, 
2001, application, is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed requirements remain the same except that the 
requirements will be relocated to the TRM. Since the proposed 
requirements are the same, this proposed change will not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the requirements remain the same, these proposed changes do 
not alter the way any system, structure, or component functions and do 
not alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the proposed 
changes will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the proposed changes are solely to relocate the existing 
requirements, it does not affect plant operation in any way. Therefore, 
the proposed change will not result in a reduction in a margin of 
safety.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 1, 2001.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical 
Specifications (TS) to change TS 3.4.6.2 ``Reactor Coolant System--
Operational Leakage''. The Bases for this TS will also be modified to 
reflect this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to Technical Specification 3.4.6.2 for 
[reactor coolant systems] RCS PIVs [pressure isolation valve] in the 
RHR [residual heat removal] flow path will not cause an accident to 
occur and will not result in any change in the operation of 
associated accident mitigation equipment. The ability of the RHR 
System to remove core decay heat will not be affected. The proposed 
changes will not affect the ability of the RCS or the RHR System to 
mitigate any design basis event. The design basis accidents will 
remain the same postulated events described in the Millstone Unit 
No. 3 Final Safety Analysis Report (FSAR), and the consequences of 
the design basis accidents will remain the same. Therefore, the 
proposed changes will not increase the probability or consequences 
of an accident previously evaluated.
    The proposed changes to delete SRs 4.4.6.2.1.a and 4.4.6.2.1.b 
and revise SR [Surveillance Requirement] 4.4.6.2.1.d will not cause 
an accident to occur and will not result in any change in the 
operation of associated accident mitigation equipment. The ability 
to measure RCS operational leakage will not be affected. The 
proposed changes will not affect the ability to mitigate any design 
basis event. The design basis accidents will remain the same 
postulated events described in the Millstone Unit No. 3 FSAR, and 
the consequences of the design basis accidents will remain the same. 
Therefore, the proposed changes will not increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to remove SR 4.4.6.2.2.c to perform post 
maintenance testing of the RCS PIVs will not cause an accident to 
occur and will not result in any change in the operation of the 
associated accident mitigation equipment. The proposed change will 
not revise the operability requirements (e.g., valve leakage limits) 
for the RCS PIVs. Proper operation of the RCS PIVs will still be 
verified, as appropriate, following maintenance activities. As a 
result, the design basis accidents will remain the same postulated 
events described in the Millstone Unit No. 3 FSAR, and the 
consequences of the design basis accidents will remain the same. 
Therefore, the proposed change will not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The proposed changes do not alter 
the way any structure, system, or component functions and do not 
alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
they have no impact on any accident analysis assumption. The 
proposed changes do not decrease the scope of equipment currently 
required to be operable or subject to surveillance testing, nor do 
the proposed changes affect any instrument setpoints or equipment 
safety functions. The effectiveness of Technical Specifications will 
be maintained since the changes will not alter the operation of any 
component or system, nor will the proposed

[[Page 57121]]

changes affect any safety limits or safety system settings. 
Therefore, there is no reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: August 22, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) Surveillance Requirement 3/
4.7.B.1.a.2 for the Standby Gas Treatment (SBGT) System by increasing 
the SBGT inlet heaters minimum output testing requirement from 14 kW to 
20 kW. The associated TS Bases 3/4.7.B.1 would also be revised as a 
result of the proposed TS change. The proposed change is based upon the 
licensee's revised design-basis calculations for the SBGT inlet heaters 
and by a modification that replaces the existing SBGT system inlet 
heaters with heaters of higher output capability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed change affects only the surveillance requirement for 
the SBGT inlet heaters output capability. The SBGT heaters are not the 
initiators of any accidents described in the safety analysis report 
(SAR). The proposed higher inlet heater output capability test is 
needed to ensure that the SBGT will continue to function as currently 
designed to decrease the relative humidity (RH) of the inlet air stream 
to 70% RH. The higher inlet heater output capability test does not 
change the consequences of an accident previously analyzed in the SAR. 
Therefore, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change to the SBGT inlet heaters capacity surveillance 
testing requirement is needed to continue to ensure that the SBGT will 
function to decrease the RH of the inlet air stream to 70% RH, as 
assumed in the current analysis. The SBGT heaters are not the 
initiators of any accidents described in the SAR. The proposed change 
in the surveillance testing requirement does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS change does not involve a significant reduction 
in the margin of safety.
    The proposed higher testing acceptance criteria for the inlet 
heater ensures that the SBGT will continue to function as currently 
designed to decrease the RH of the inlet air steam to 70% RH. The 
margin of safety is unaffected by this change. Therefore, the proposed 
change does not involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Clifford.

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: October 17, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) actions regarding inoperable 
redundant components when an Emergency Diesel Generator (EDG) becomes 
inoperable. TS 3.8.1.1 would be revised to require actions based on the 
TS for the inoperable redundant component(s). The proposed revision is 
consistent with NUREG-1432, Rev.2, ``Standard Technical Specifications, 
Combustion Engineering Plant.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would operation of the facility in accordance with the 
proposed amendments involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Neither the steam driven auxiliary feedwater pump nor the EDGs 
are accident initiators, but are accident mitigators. The proposed 
changes to the EDG TS do not affect the operation nor availability 
of the EDGs, the motor or steam driven auxiliary feedwater pumps, 
nor TS required redundant features. For those conditions that would 
require a unit shutdown, once the four hour completion time had 
expired, the shutdown would be performed in the manner and timeframe 
supported by the existing redundant feature TS. Therefore, the 
probability or consequences of any accident previously evaluated 
have not been significantly increased.
    2. Would operation of the facility in accordance with the 
proposed amendments create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new failure modes are introduced by the proposed TS changes 
and single failure considerations are adequately addressed by 
following the established conventions of NUREG-1432. The proposed 
four hour completion time from the discovery of inoperable redundant 
features and an EDG takes into account the operability of the 
redundant counterpart to the inoperable required feature, the 
capacity and capability of the remaining AC sources, a reasonable 
time for repairs, and the low probability of a DBA [design-basis 
accident] occurring during this period. The TS change required 
reformatting and moving the steam driven auxiliary feedwater pump 
operability requirements to the redundant feature(s) actions to be 
comparable with and meet the intent of the BASES requirements 
contained in NUREG-1432. Without creation of a new interaction of 
materials, operating configuration, or operating interfaces, there 
is no possibility that the proposed changes can introduce a new or 
different kind of accident.
    3. Would operation of the facility in accordance with the 
proposed amendments involve a significant reduction in a margin of 
safety?
    The margin of safety as defined in the basis for any Technical 
Specification or in any licensing document has not been reduced. The 
proposed changes remove the unconditional unit shutdown requirement 
should an EDG be inoperable while required features on the opposite 
train are inoperable. Instead, any TS required actions are 
appropriately based on the inoperability of the required feature. 
The proposed four hour completion time from the discovery of 
inoperable redundant features and an EDG takes into account the 
operability of the redundant counterpart to the inoperable required 
feature, the capacity and capability of the remaining AC sources, a 
reasonable time for repairs, and the low probability of

[[Page 57122]]

a DBA occurring during this period. For those conditions that would 
require a unit shutdown, once the four hour completion time had 
expired, the shutdown would be performed in the manner and timeframe 
supported by the existing redundant feature TS. Additionally, the TS 
requirements to assure that steam driven auxiliary feedwater pump 
operability is considered as part of the redundant features 
requirements remains and is comparable to the intent of the BASES of 
STS 3.8.1. Based on the preceding discussion, FPL concludes that the 
margin of safety will not be significantly reduced by operation of 
the facility in accordance with the proposed amendments.

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

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

    Date of amendment request: October 17, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) multiplier values for single-
loop operation (SLO) average planar linear heat generation rate 
(APLHGR).
    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 APLHGR multipliers, and their 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 purpose of the APLHGR limit is to assure that the 
fuel will not exceed a peak cladding temperature (PCT) of 2200  deg.F 
during a Loss of Coolant Accident [LOCA], as required by 10 CFR 50.46. 
Specifying appropriate APLHGR multipliers ensures that a LOCA in SLO 
will not produce a PCT any greater than the PCT produced by a LOCA in 
dual loop operation. These changes ensure that the appropriate SLO 
APLHGR multiplier, required for GE14 fuel, is incorporated into the 
Monticello TS. These changes do not alter the method of operating the 
plant.
    Therefore, the proposed TS changes do 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 changes result only from different inputs, including 
use of GE14 fuel, 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 single-loop operation APLHGR multiplier values are designed to 
ensure that the PCT resulting from a LOCA while operating in SLO are 
bounded by the PCT from a LOCA while operating in dual loop operation. 
This multiplier update results from application of GE Nuclear Energy's 
(GE's) current standard methodology for this analysis.
    Therefore, the proposed TS changes do 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 APLHGR limits are set appropriately below the value where 
significant fuel damage could occur in a Loss of Coolant Accident 
(LOCA). Application of new SLO APLHGR multiplier values ensure that SLO 
LOCA results are bounded by those for dual loop operation and thus 
maintain or improve the margin of safety for LOCA analyses.
    Therefore, the proposed TS changes do 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: William D. Reckley.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: September 19, 2001.
    Description of amendment request: The proposed amendment would 
revise the Unit 1 reactor pressure vessel (RPV) material surveillance 
program to defer the withdrawal of the second surveillance capsule for 
one operating cycle. Deferral is requested to support PPL Susquehanna, 
LLC's, participation in the Boiling Water Reactor Vessel and Internals 
Project Integrated Surveillance Program.
    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. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Pressure-temperature (P/T) limits are imposed on the reactor 
coolant system to ensure that adequate safety margins against non-
ductile or rapidly propagating failure exist during normal 
operation, anticipated operational occurrences, and system 
hydrostatic tests. The P/T limits are related to the nil-ductility 
reference temperature, RTndt. Changes in the fracture toughness 
properties of the Reactor Pressure Vessel (RPV) beltline materials, 
resulting from neutron irradiation and the thermal environment, are 
monitored by a surveillance program in compliance with the 
requirements of 10 CFR 50, Appendix H. The effect of neutron fluence 
on the shift in the nil-ductility reference temperature of pressure 
vessel steel is predicted by methods given in Regulatory Guide (RG) 
1.99, Revision 2 and Regulatory Guide 1.190, Revision 0. The 
Susquehanna SES [Steam Electric Station] Unit 1 current P/T limits 
were established based on adjusted reference temperatures developed 
in accordance with the procedures prescribed in RG 1.99, Revision 2. 
Calculation of adjusted reference temperature by these procedures 
includes a margin term to ensure upperbound values are used for the 
calculation of the P/T limits. Revision of the second capsule 
withdrawal schedule will not affect the P/T limits, because they 
will continue to be established in accordance with NRC approved 
methodology in accordance with RG 1.190 Revision 0 commitments. The 
existing P/T limits are based on 32 EFPY rather than for the planned 
withdrawal at 15 EFPY. This change is not related to any accidents 
previously evaluated. The proposed change will not affect reactor 
pressure vessel performance because no physical changes are involved 
and the RPV vessel P/T limits will remain in accordance with RG 
1.99, Revision 2 commitments. The proposed change will not cause the 
reactor pressure vessel or

[[Page 57123]]

interfacing safety systems to be operated outside of their design or 
testing limits. Also, the proposed change will not alter any 
assumptions previously made in evaluating the radiological 
consequences of accidents.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    B. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously analyzed?
    The proposed change defers the second RPV material surveillance 
capsule withdrawal for one fuel cycle. This proposed change does not 
involve a modification of the design of plant structures, systems, 
or components. The proposed change will not impact the manner in 
which the plant is operated as plant operating and testing 
procedures will not be affected by the change. The proposed change 
will not degrade the reliability of structures, systems, or 
components important-to-safety because equipment protection features 
will not be deleted or modified, equipment redundancy or 
independence will not be reduced, supporting system performance will 
not be downgraded, the frequency of operation of equipment 
important-to-safety will not be increased, and more severe testing 
of equipment important-to-safety will not be imposed. No new 
accident types or failure modes will be introduced as a result of 
the proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from previously 
analyzed.
    C. Does the proposed change involve a significant reduction in a 
margin of safety?
    Appendix G to 10 CFR 50 describes the conditions that require P/
T limits and provide the general bases for these limits. Until the 
results from the reactor vessel surveillance program become 
available, RG 1.99, Revision 2 is used to predict the amount of 
neutron irradiation damage. The use of operating limits based on 
these criteria, as defined by applicable regulations, codes, and 
standards, provide reasonable assurance that nonductile or rapidly 
propagating failure will not occur. The P/T limits are not derived 
from Design Basis Accident (DBA) analyses. They are prescribed 
during normal operation to avoid encountering pressure, temperature, 
and temperature rate of change conditions that might cause 
undetected flaws to propagate and cause nonductile failure of the 
reactor coolant pressure boundary (RCPB). Since the P/T limits are 
not derived from any DBA, there are no acceptance limits related to 
the P/T limits. Rather, the P/T limits are acceptance limits 
themselves since they preclude operation in an unanalyzed condition. 
The proposed change will not affect any safety limits, limiting 
safety system settings, or limiting conditions of operation. The 
proposed change does not represent a change in initial conditions, 
or in a system response time, or in any other parameter affecting 
the course of an accident analysis supporting the Bases of any 
Technical Specification. The proposed change does not involve 
revision of the P/T limits, but rather a revision of the withdrawal 
time for the second surveillance capsule. The current P/T limits 
were established based on adjusted reference temperatures for vessel 
beltline materials calculated in accordance with RG 1.99, Revision 
2. P/T limits will continue to be revised, as necessary, for changes 
in adjusted reference temperature due to changes in fluence when two 
or three credible surveillance data sets become available. When two 
or more credible surveillance data sets become available, P/T limits 
will be revised as prescribed in RG 1.190, Revision 0.
    Therefore, the proposed changes do not involve a significant 
reduction in any margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: L. Raghavan, 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 30, 2001, as supplemented August 7, 
and October 16, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.5.12, ``Primary Containment Leakage 
Rate Testing Program,'' to allow a one-time deferral of the Type A 
containment integrated leakage rate test (ILRT) at the Susquehanna 
Steam Electric Station (SSES), Units 1 and 2. The Unit 1 test would be 
deferred to no later than May 3, 2007, and the Unit 2 test would be 
deferred to no later than October 30, 2007, resulting in an extended 
interval of 15 years for performance of the next ILRT at each unit.
    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?
    The frequency of Type A testing does not change the probability 
of an event that results in core damage or vessel failure. Primary 
containment is the engineered feature that contains the energy and 
fission products from evaluated events. The SSES IPE [Individual 
Plant Examination] documents events that lead to containment 
failure. The frequency of events that lead to containment failure 
does not change because it is not a function of the Type A test 
interval. Containment failure is a function of loss of safety 
systems that shutdown the reactor, provide adequate core cooling, 
provide decay heat removal, and drywell sprays.
    The consequences of the evaluated accidents are the amount of 
radioactivity that is released to secondary containment and 
subsequently to the public. Normally, extending a test interval 
increases the probability that a Structure System or Component will 
be failed. However, NUREG-1493, Performance-Based Containment Leak-
Test Program, states that calculated risks in BWR's is very 
insensitive to the assumed leakage rates. The remaining testing and 
inspection programs provide the same coverage as the Type A test. 
These other programs will maintain containment leakage low. Any 
leakage path problems will be identified and repairs will be made. 
Additionally the containment is continuously monitored during power 
operation. Anomalies are investigated and resolved. Thus there is a 
high confidence that containment integrity will be maintained 
independent of the Type A test frequency.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence 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 previously analyzed?
    Primary containment is designed to contain energy and fission 
products during and after an event. The SSES IPE identifies events 
that lead to containment failure. Revision to the Type A test 
interval does not change this list of events. There are no physical 
changes being made to the plant and there are no changes to the 
operation of the plant that could introduce a new failure mode 
creating an accident or affecting mitigation of an accident.
    Therefore, this proposed amendment does not involve a 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed revision to Technical Specifications adds a one 
time extension to the current interval for Type A testing. The 
current level of 10 years, based on past performance, would be 
extended on a one time basis to 15 years from the last Type A test. 
The NUREG-1493 generic study of the effects of extending containment 
leakage testing found that a 20-year interval 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 about 0.1% to the individual risk and that 
increasing the Type A test interval would have minimal affect on 
this risk since 95% of the potential leakage paths are detected by 
Type B and Type C testing. Technical Specifications require that 
maximum allowable primary containment leakage rate is less than 1% 
primary containment air

[[Page 57124]]

weight per day. During unit startup following Type B and Type C 
testing, leakage rate acceptance criteria must be less than 0.6% 
primary containment air weight per day. (TS 5.5.12) Therefore, Type 
B and Type C testing combined with visual inspection programs will 
maintain containment leakage low.
    Therefore, these changes do not involve a significant reduction 
in margin of safety.

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

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

    Date of amendment request: April 16, 2001, as supplemented on July 
5, 2001.
    Description of amendment request: The proposed Technical 
Specifications (TSs) change would modify required actions and 
surveillance requirements (SR) associated with the 28 Volt Direct 
Current (VDC) Battery System. The proposed changes are consistent with 
TS and SR requirements for the 125 VDC Battery System, and NUREG-1431, 
``Standard Technical Specifications--Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to TS limiting conditions for operation (LCOs) 
and surveillance requirements (SRs) will not alter the plant's physical 
configuration or the operation of the 28 VDC Battery System. As a 
result, the parameters assumed in the Salem Updated Final Safety 
Analysis Report (UFSAR) Design Basis Accident or Transient Analyses 
remain unchanged. Therefore, the probability or consequences of an 
accident previously evaluated are not increased by the proposed change.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed changes to the 28 VDC Battery System TS LCOs and SRs 
do not modify the facility's design or physical configuration or change 
the method by which any safety-related system performs its function. 
Therefore, the proposed changes will not increase the possibility of a 
new or different kind of accident from any accident previously 
identified.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    The proposed changes do not alter the manner in which safety limits 
or limiting safety system setpoints are determined. As a result, 
margins of safety are not changed. Therefore, the proposed changes do 
not involve a significant reduction in any margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 24, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.9, ``Refueling Operations,'' by 
relocating requirements for boron concentration to the Core Operating 
Limits Report (COLR). The proposed amendment will revise Limiting 
Condition for Operation (LCO) 3.9.1 by stating that, while the plant is 
in Mode 6, boron concentration of the Reactor Coolant System (RCS), 
refueling canal, and the refueling cavity shall be maintained within 
the limits specified in the COLR. LCO 3.9.1 required actions will also 
be revised to reference the COLR, and associated surveillance 
requirements will be changed to state that boron concentration shall be 
verified to be within the limits provided in the COLR every 72 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Relocating the minimum required boron concentration values from the 
TSs to the COLR does not change boron concentration requirements. 
Specifying the required minimum boron concentration in the COLR will 
continue to ensure that the proper boron concentration will be 
maintained in accordance with all the assumptions of appropriate 
accident analyses. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed change relocates the minimum required boron 
concentration values from the TSs to the COLR. Moreover, the proposed 
change does not physically change the facility, plant operations, or 
the manner and frequency at which associated boron concentration 
testing is conducted. Therefore, the proposed change to relocate the 
required boron concentration to the COLR does not create the 
possibility of a new or different kind of accident from any previously 
analyzed.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    Minimum boron concentration limits are established to ensure that 
sufficient margins exist to prevent criticality in the RCS, refueling 
canal, and the refueling cavity during refueling operations. Since the 
COLR is prepared as part of each core reload safety evaluation to 
ensure that current safety analysis limits are met, relocating the 
minimum boron concentration from the TSs to the COLR will not reduce 
safety margins. Therefore, the new proposed change to relocate the 
required boron concentration to the COLR does not involve a significant 
reduction in the margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

[[Page 57125]]

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

    Date of amendment request: June 19, 2001.
    Description of amendment request: South Carolina Electric & Gas 
Company (SCE&G) proposes a change to the Virgil C. Summer Nuclear 
Station (VCSNS) Technical Specifications (TS) Surveillance Requirements 
to revise Table 3.7-1. This change will identify maximum allowable 
power range neutron flux high setpoints based on the plant safety 
analysis or conservatively derived values calculated in accordance with 
NRC Information Notice 94-60 and Westinghouse Nuclear Safety Advisory 
Letter NSAL-94-001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes to Technical Specification 3.7.1 and its 
associated bases do not contribute to the initiation of any accident 
previously evaluated. Supporting factors are as follows:
    All NSSS components are compatible with the revised core power 
limits and resulting operating conditions. Their structural 
integrity is maintained during all proposed plant conditions through 
compliance with the ASME code.
    Other systems important to safety are not adversely impacted and 
will continue to perform their design functions.
    The revised core power limits and resulting operating conditions 
remain within the design envelope of the plant.
    Therefore, since the reactor coolant pressure boundary integrity 
and system functions are not adversely impacted, the probability of 
occurrence of an accident previously evaluated will be no greater 
than the existing design basis of the plant. The revised method to 
derive allowable power levels with inoperable main steam safety 
valves results in lower High Flux Trip Setpoints. When implemented, 
the revised trip setpoints ensure that secondary system pressure 
will be limited to within 110% (1305 psig) of its design pressure of 
1185 psig during the most severe anticipated system operational 
transients. Since the ASME and regulatory limits on secondary side 
overpressurization will be met, the proposed changes will not create 
the potential for an increase in offsite releases or doses for any 
accident. Therefore, there is no increase in the consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes to Technical Specification 3.7.1 and its 
associated bases do not introduce any new accident initiator 
mechanisms. Structural integrity of the RCS and the secondary side 
is maintained during the allowed operating conditions, and ASME code 
limits continue to be met during all anticipated operating 
conditions. In addition, no new failure modes or limiting single 
failure or new design requirements for auxiliary systems are being 
introduced. Since the safety and design requirements continue to be 
met and the integrity of the primary and secondary pressure boundary 
is maintained, no new accident scenarios have been created. 
Therefore, the types of accidents previously defined continue to 
represent the credible spectrum of events to be analyzed. A new or 
different kind of accident is thus not created.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed changes to Technical Specification 3.7.1 and its 
associated bases preserve the results and conclusions of plants 
safety analyses presented in the FSAR. The proposed changes address 
an identified deficiency with the current Technical Specification 
and, when implemented, restores the margin of safety intended. 
Specifically, the proposed changes ensures overpressure ensure that 
the secondary system pressure will be limited to within 110% (1305 
psig) of its design pressure of 1185 psig during the most severe 
anticipated system operational transient. Therefore, there is no 
reduction in a margin of safety.

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: September 19, 2001.
    Description of amendment request: The proposed amendments would 
revise surveillance requirement 3.6.1.3.8 which currently requires 
verification of the actuation capability of each reactor 
instrumentation excess flow check valve (EFCV) every 18 months. The 
proposed amendments would state that a representative sample of the 
EFCVs will be tested every 18 months such that each EFCV will be tested 
at least once every 10 years. The proposed amendments are consistent 
with Technical Specification Task Force-334.
    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 a previously evaluated event?
    The Excess Flow Check Valves are designed to limit the flow from 
an instrument line break downstream of the check valve itself. Thus 
the previously analyzed event is the instrument line break, 
documented in the Unit 2 FSAR, section 15.4.13, for both units. This 
proposed revision does not alter the operation or maintenance of any 
instrument lines; the revision is made to reduce the surveillance 
requirements for the EFCVs. This revision does nothing which 
jeopardizes the integrity of the instrument lines and thus increase 
the probability of a line break.
    The line break analysis does not take credit for operation of 
the excess flow check valves, therefore, the radiological 
consequences of this event are not affected by this proposed TS 
revision.
    This amendment request does not affect any other previously 
evaluated line or pipe break analsis.
    For the above reasons, the probability of occurrence, or the 
consequences of a previously evaluated event are not increased by 
this proposed change.
    2. Do the proposed changes create the possibility of a new type 
event different from any previously evaluated?
    No changes are being made to the way in which the EFCVs are 
operated, or maintained; they will continue to be operated within 
the conditions for which they were designed. Since no new 
operational modes are proposed, no new failure modes are introduced.
    Furthermore, no changes to any systems designed for the 
prevention of transients or accidents are being made as a result of 
this proposed Technical Specification change.
    For the above reasons, this proposed change does not introduce 
the possibility of a different type event from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The reactor coolant pressure boundary line break analysis 
documented in Unit 2 FSAR section 15.4.13 does not assume credit for 
the EFCVs. Additionally, the failure rate of the Unit 1 and 2 EFCVs 
has been small, as verified by the failure rate analysis done for 
this proposed revision. Accordingly, reducing the frequency of the 
surveillance is justified and will not significantly reduce the 
margin of safety with respect to EFCV failure.
    Additionally, General Electric has performed a generic 
radiological evaluation of an instrument line break, with EFCV 
failure, which concluded that the dose

[[Page 57126]]

consequences would not exceed 10 CFR 100 guidelines. This analysis 
is documented in NEDO-32977-A, ``Excess Flow Check Valve 
relaxation'', a report commissioned by the Boiling Water Reactors 
Owners' Group (BWROG). Because the Hatch EFCV design is similar to 
the EFCV designs assumed in the NEDO, it is reasonable to conclude 
that the results of this generic analysis are bounding for Plant 
Hatch.

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: August 10, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.9.1, ``Refueling Equipment 
Interlocks,'' to provide alternative actions when the refueling 
equipment interlocks are inoperable.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The operation of refueling interlocks is explicitly assumed in 
the analyses of the control rod removal error and fuel loading error 
during refueling. Inadvertent criticality is prevented during the 
loading of fuel provided all control rods are fully inserted. The 
refueling interlocks accomplish this by preventing the loading of 
fuel into the core with any control rod withdrawn, or by preventing 
withdrawal of a rod from the core during fuel loading. Under 
existing TS when the refueling interlocks are inoperable, the 
current method of preventing fuel loading with control rods 
withdrawn is to prevent fuel movement. An alternate method to ensure 
that fuel is not loaded into a cell with a control rod withdrawn is 
to prevent control rods from being withdrawn and to verify that all 
control rods are fully inserted. The proposed TS Required Actions 
will require that a control rod block be placed in effect, thereby 
ensuring that control rods are not subsequently inappropriately 
withdrawn, and that all required control rods be verified to be 
fully inserted. This verification is in addition to the requirements 
to periodically verify control rod position by other TS 
requirements.
    The proposed actions will ensure that control rods are not 
withdrawn and cannot be inappropriately withdrawn, because a control 
rod withdrawal block is in place. Like the current TS requirements, 
the proposed actions will ensure that unacceptable operations are 
blocked. Hence, the proposed additional Required Actions provide an 
equivalent level of assurance that fuel will not be loaded into a 
core cell with a control rod withdrawn as does the current TS 
Required Action. Therefore, the proposed change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change in the TS requirements does not involve a change in 
plant design or to the analyzed condition of the reactor core during 
refueling. The proposed new Required Actions will ensure that 
control rods are not withdrawn and cannot be inappropriately 
withdrawn, because a block to control rod withdrawal is in place. 
Therefore, no new failure modes are introduced, and the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    As discussed in the Bases for the affected TS requirements, 
inadvertent criticality is prevented during the loading of fuel 
provided all control rods are fully inserted during the fuel 
insertion. The refueling interlocks function to support the 
refueling procedures by preventing control rod withdrawal during 
fuel movement and the inadvertent loading of fuel when a control rod 
is withdrawn. The proposed change will allow the refueling 
interlocks to be inoperable and fuel movement to continue only if a 
control rod withdrawal block is in effect and all control rods are 
verified to be fully inserted. These proposed Required Actions 
provide an equivalent level of protection as the refueling 
interlocks by preventing a configuration that could lead to an 
inadvertent criticality event. The refueling procedures will 
continue to be supported by the proposed Required Actions because 
control rods cannot be withdrawn and as a result, fuel cannot be 
inadvertently loaded when a control rod is withdrawn. Therefore, the 
proposed changes do not result in a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: August 17, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 3.3.1.1-1, ``Reactor 
Protection System [RPS] Instrumentation,'' to remove one RPS function 
and modify another.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Modifications to the Scram Discharge Instrument Volume (SDIV) 
System are being implemented to ensure that the SDIV high water 
level instrumentation will respond adequately to provide redundant, 
diverse trip functions for a Scram Discharge Volume (SDV) inleakage 
event. Since the scram function will be successfully performed, the 
removal of the low scram pilot air header pressure trip function 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The design criteria for the Scram Discharge System is contained 
in the Safety Evaluation Report on the BWR Scram Discharge System, 
which was transmitted by NRC letter dated December 9, 1980, to all 
BWR licensees. Modifications to the SDV System have been evaluated 
to demonstrate that the high water level instrumentation in the SDIV 
will respond adequately to provide the required trip function. No 
new system failure modes are created as a result of removing the low 
scram pilot air header trip, since the redundant and diverse SDIV 
high water level instruments will initiate a successful reactor 
scram. Therefore, the removal of the low scram pilot air header trip 
function does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The water level in the SDIV is monitored by both resistance-
temperature type detectors and float switches. Redundancy and 
diversity in the instrumentation that initiates the

[[Page 57127]]

scram signal is maintained even with the removal of the low scram 
pilot air header pressure trip function. Modifications to the SDIV 
System have been evaluated to demonstrate that the high water level 
instrumentation will respond adequately to provide the required trip 
function for an inleakage event. Therefore, the proposed amendment 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendments: October 9, 2001 (TS 01-10).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah (SQN) Unit 2 Operating License Technical 
Specifications (TSs), specifically TS 6.8.4.h, ``Containment Leakage 
Rate Testing Program,'' to allow a one-time 5-year extension to the 
current 10-year test interval for the containment performance-based 
leakage rate test program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed extension to Type A testing does not increase the 
probability of an accident previously evaluated since the change is 
not a modification to plant systems, nor a change to plant operation 
that could initiate an accident.
    TVA performed an evaluation of the risk significance for the 
proposed increase to the Sequoyah Unit 2 Type A test frequency. The 
results of the TVA evaluation indicate that the increase in Large 
Early Release Frequency (LERF) remains below the level of risk 
significance defined in NRC Regulatory Guide (RG) 1.174, ``An 
Approach for Using Probabilistic Risk Assessment In Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis.'' TVA's 
evaluation indicates that the increase in frequency for all releases 
(small, large, early and late) and the increase in radiation dose to 
the population is non-risk significant (3.5E-7/reactor year and 7.72 
person-rem, respectively).
    The proposed test interval extension does not involve a 
significant increase in the consequences of an accident because 
research documented in NUREG-1493 determined that generically, very 
few potential containment leakage paths fail to be identified by 
Type A tests. An analysis of 144 Type A test results, including 23 
failures, found that no failures were due to containment liner 
breach. The NUREG concluded that reducing the Type A test frequency 
to once per 20 years would lead to an imperceptible increase in 
risk. Furthermore, the NUREG concluded that Type B and C testing 
provides assurance that containment leakage from penetration leak 
paths (i.e., valves, flanges, containment air-locks) identify any 
leakage that would otherwise be detected by the Type A tests.
    In addition to the NUREG conclusions, TVA's American Society of 
Mechanical Engineers (ASME) IWE program performs containment 
inspections periodically in order to detect evidence of degradation 
that may affect either the containment structural integrity or leak 
tightness. Accordingly, TVA's proposed extension of the Type A test 
interval does not [significantly] increase the probability or 
consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to extend the Type A test interval does not 
create the possibility of a new or different type of accident since 
there are no physical changes made to the plant. There are no 
changes to the operation of the plant that would introduce a new 
failure mode creating the possibility of a new or different kind of 
accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to extend the Type A test interval will not 
significantly reduce the margin of safety. A generic study 
documented in NUREG-1493 indicates that extending the Type A leak 
test interval to 20 years would result in an imperceptible increase 
in risk to the public. The NUREG also found that, generically, the 
containment leakage rate contributes a very small amount to the 
individual risk and that the decrease in the Type A test frequency 
would have a minimal affect on risk because most potential leakage 
paths are detected by Type C testing.
    Previous Type A leakage tests conducted on Sequoyah Unit 2 
indicate that leakage from Unit 2 containment has been less than the 
10 CFR 50 Appendix J leakage limit of 1.0 La. A review of 
previous Unit 2 Type A test results indicate at least a 10 percent 
margin exists below the 1.0 La leakage limit. These test 
results provide assurance that the proposed extension to the Type A 
test interval would not significantly reduce the margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 57128]]

documents located in ADAMS, contact the NRC Public Document Room (PDR) 
Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

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

    Date of application for amendment: May 31, 2001, as supplemented 
September 14, 2001.
    Brief description of amendment: The amendment revised the TMI-1 
Technical Specifications (TSs) to incorporate Cycle 14 specific limits 
for the variable low reactor coolant system pressure-temperature core 
protection safety limits. These changes are reflected in revisions to 
Figures 2.1-1 and 2.1-3 of the TSs and the related Bases.
    Date of issuance: October 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 238.
    Facility Operating License No. DPR-50.: Amendment revised the TSs.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36337).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 30, 2000, as 
supplemented by letter dated September 13, 2001.
    Brief description of amendment: The amendment modified the Final 
Safety Analysis Report (FSAR) to reflect analysis of a HI-STORM 100 
spent fuel cask system, spent fuel pool description and crane 
operations.
    Date of issuance: October 26, 2001.
    Effective date: October 26, 2001, and shall be implemented in the 
next periodic update to the FSAR in accordance with 10 CFR 50.71(e).
    Amendment No.: 174.
    Facility Operating License No. NPF-21: The amendment revised the 
Final Safety Analysis Report.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15918). The September 13, 2001, supplemental letter provided additional 
clarifying information, did not expand the scope of the original 
Federal Register notice, 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 October 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 23, 2001, as supplemented 
by letter dated September 25, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications to eliminate the requirement to move control element 
assembly #43 for the remainder of Cycle 15.
    Date of issuance: October 22, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 235.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46478). The September 25, 2001, supplemental letter provided 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 18, 2001.
    Brief description of amendment: The amendment changes Technical 
Specifications (TS) Definitions 1.12 and 1.25, the effect of which will 
be to allow either an allocated or a measured response time to be 
utilized for the sensors in the Reactor Protective System and 
Engineered Safety Features Actuation System instrument loops.
    Date of issuance: October 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 175.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46479).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 29, 2001.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: October 24, 2000, as 
supplemented by letters dated June 18 and August 21, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification 3.8.3 regarding the lube oil inventories for the Grand 
Gulf Nuclear Station, Unit 1, Divisions I, II, and III emergency diesel 
generators (EDGs), and will result in additional margins for lube oil 
availability to provide for EDG operability for seven days following a 
postulated design basis accident.
    Date of issuance: October 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 149.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7680).
    The June 18 and August 21, 2001 supplemental letters did not change 
the scope of the original Federal Register notice or the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 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: October 1, 2001, as 
supplemented by letters dated October 9, 2001, and October 18, 2001. 
The supplemental letters provided clarifying information only and did 
not change the original proposed no significant hazards determination.
    Brief description of amendments: The amendments revise Byron and 
Braidwood technical specifications (TS) surveillance requirement (SR) 
3.7.2.1 and SR 3.7.2.2 to add a note stating that these surveillances 
are not required to be met until the first startup after September 27, 
2001. This change is

[[Page 57129]]

applicable to Byron Station Units 1 and 2, and Braidwood Unit 2 only. 
This change is not applicable to Braidwood Station, Unit 1, due to the 
recent restart of the unit after the refueling outage.
    Date of issuance: November 1, 2001.
    Effective date: November 1, 2001.
    Amendment Nos.: 124, 124, 119, and 119.
    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: October 23, 2001 (66 FR 
53643).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: December 2, 2000, as 
supplemented by letters dated September 4 and September 28, 2001.
    Brief description of amendment: This amendment increases the spent 
fuel pool (SFP) storage capability, as a result of the SFP re-racking 
project, from the current capacity of 735 fuel assemblies to a new 
capacity of 1624 fuel assemblies. The amendment also approves 
additional temporary storage of up to 90 fuel assemblies in the fuel 
transfer pit to support a complete re-racking of the SFP. The increase 
in SFP storage capacity will provide a full core offload capability 
during the plant's Cycle 13 operation and enable the Davis-Besse 
facility to meet its storage needs through April 22, 2017, which is the 
expiration date for the current operating license.
    Date of issuance: October 19, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 247.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2001 (66 
FR 46656).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 19, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346 Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: April 4, 2001.
    Brief description of amendments: This license amendment request: 
Deletes Technical Specification (TS) 1.7, Definitions-Reportable Event, 
and TS 6.6, Reportable Event--Action; Revise TS 6.5.3, Technical Review 
and Control--Activities, and TS Bases 4.0.3, Applicability.
    Date of issuance: November 2, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 248.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 2001 (66 
FR 31708).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 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: June 22, 2001, as supplemented 
August 24, 2001.
    Brief description of amendment: Revised Technical Specifications to 
allow the containment equipment door and airlock doors to be open 
during core alterations and fuel movement under administrative 
controls.
    Date of Issuance: October 22, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 120.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48287).
    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 October 22, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 24, 2000, as 
supplemented June 29, 2001.
    Brief description of amendments: The amendments would approve 
changes to the updated final safety analysis report to incorporate a 
supplemental methodology into the analysis of steam generator overfill 
following a steam generator tube rupture.
    Date of issuance: October 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 256 and 239.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7682). The supplemental letter 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 amendments 
is contained in a Safety Evaluation dated October 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 12, 2000, as supplemented 
by letters dated November 7, 2000, June 19, and August 17, 2001.
    Brief description of amendments: The amendments revised the 
technical specifications to change the standard by which you test 
charcoal used in engineered safeguard features systems to American 
Society for Testing and Materials D3808-1989. These revisions are made 
in accordance with Generic Letter 99-02, ``Laboratory Testing of 
Nuclear-grade Activated Charcoal.''
    Date of issuance: October 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 257 and 240.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 23, 2000 (65 FR 
51356). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination

[[Page 57130]]

and did not expand the scope of the original Federal Register notice. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated October 24, 2001.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: February 28, 2001, as supplemented by 
letters dated September 14, 18, and 27, 2001. The letters dated 
September 14, 18, and 27, 2001, provided clarifying information, and 
did not alter the NRC staff's conclusions regarding finding of no 
significant hazards consideration.
    Brief description of amendment: The amendment evaluates the 
licensee's revised calculation methodology for assessment of 
consequences of design basis accidents, and revises Technical 
Specifications.
    Date of issuance: October 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 187.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48288).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 2001.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: April 12, 2001.
    Brief description of amendment: The Amendment revises the Technical 
Specifications Bases Control Program to incorporate revisions to 10 CFR 
50.59.
    Date of issuance: October 25, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 188.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48289).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 25, 2001.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: April 12, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications surveillance test requirement SR 3.6.1.3.8, for excess 
flow check valves (EFCVs), to relax the 18-month EFCV surveillance 
frequency by limiting the number of tests to a ``representative 
sample'' every 18 months, such that each EFCV will be tested at least 
once every 10 years.
    Date of issuance: October 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 189.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48289).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 12, 2001.
    Description of amendment request: The amendment changes Technical 
Specification (TS) 4.4.10 to incorporate alternative reactor coolant 
pump flywheel inspections and makes administrative wording changes to 
TSs 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3.
    Date of issuance: October 22, 2001.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 79.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38764).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 18, 2001, as supplemented 
October 10, 2001.
    Brief description of amendment: The amendment (1) deletes a 
redundant requirement for valving out control rod drives, (2) revises 
control rod accumulator operability requirements, (3) adds the option 
to hydraulically isolate control rod drives, and (4) corrects an 
inconsistency describing when source range monitors are required to be 
operable during core monitoring.
    Date of issuance: October 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 123.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31711).
    The supplement provided clarifying information to the application 
that was within the scope of the original Federal Register notice and 
did not change the staff's initial proposed no significant hazards 
considerations determination. The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated October 26, 
2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 15, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to (1) reflect the replacement of Monticello's 
licensed operator initial and requalification training programs with an 
accredited systems-approach-to-training program and (2) relocate the 
existing TS requirements for procedures, records, and reviews to the 
Operational Quality Assurance Plan.
    Date of issuance: October 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 124.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48290).

[[Page 57131]]

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

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

    Date of application for amendments: August 31, 2001.
    Brief description of amendments: The amendments extend the 
implementation date for Amendment No. 184 for Unit 1 and Amendment No. 
158 for Unit 2 from November 1, 2001, to November 1, 2003. Amendment 
Nos. 184 and 158 approved technical specification changes to 
incorporate requirements related to oscillation power range monitoring 
(OPRM) instrumentation. The implementation date extension is needed to 
provide additional time to address software deficiencies with the OPRM 
system identified in a June 29, 2001, General Electric report filed 
pursuant to part 21 of Title 10 of the Code of Federal Regulations.
    Date of issuance: October 29, 2001.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 196 and 172.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the license.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48291).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 29, 2001.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit 1, San Diego County, California

    Date of application for amendment: October 30, 2000, as 
supplemented by letters dated May 7, June 13, 2001, and by internet 
memoranda dated June 28, July 3, July 23, and October 16, 2001.
    Brief description of amendments: Amendment Application No. 217 is a 
request to revise the San Onofre Nuclear Generating Station, Unit 1 
(SONGS 1) operating license and technical specifications to remove 
certain requirements that have been determined to be unnecessary and 
modify requirements to provide flexibility during the decommissioning 
of SONGS 1. This change removes the need to perform activities that are 
not providing a benefit to safely maintain the spent fuel in the spent 
fuel pool. This change also provides some flexibility in the operation 
of the spent fuel pool during the decommissioning of SONGS 1.
    Date of issuance: October 30, 2001.
    Effective date: October 30, 2001, to be implemented within 30 days 
of issuance.
    Amendment No.: Unit 1-160.
    Facility Operating License No. DPR-13: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77924).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 6, 2001 (TS 01-05).
    Brief description of amendments: The amendments revised the SQN 
Unit 1 and 2 Technical Specifications (TSs) by changing the 
surveillance requirements for verifying that containment isolation 
valves to be closed. More specifically, valves in high radiation areas 
may be verified by administrative means. In addition, valves which are 
locked sealed or otherwise secured do not need to be reverified closed 
and are eliminated from the scope of the surveillance.
    Date of issuance: October 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 271 and 260.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44177). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 23, 2001.
    Brief description of amendment: The amendment updates the license 
by deleting obsolete information, correcting errors, and making 
administrative changes to enhance the context and provide consistency.
    Date of Issuance: October 22, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-28: Amendment revised the 
License.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29363).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 22, 2001.
    No significant hazards consideration comments received: No.

    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 6th day of November 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-28399 Filed 11-13-01; 8:45 am]
BILLING CODE 7590-01-P