[Federal Register Volume 66, Number 143 (Wednesday, July 25, 2001)]
[Notices]
[Pages 38756-38771]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-18324]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 2, 2001 through July 13, 2001. The last 
biweekly notice was published on July 11, 2001 (66 FR 36335).

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

[[Page 38757]]

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 20852. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By August 24, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to [email protected].

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

    Date of amendment request: June 7, 2001.
    Description of amendment request: The proposed amendment request 
would revise the requirement for the Senior Manager-Operations to hold 
a Senior Reactor Operator license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 38758]]


    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    The proposed change to TS [Technical Specification] 6.2.2.2.j 
revises the requirement concerning the Operations management 
position that must hold an SRO [Senior Reactor Operator] license. At 
least one of the Operations Managers or the Senior Manager-
Operations will continue to meet NRC requirements for maintaining an 
SRO license. The training, qualification, and experience 
requirements for Operations management personnel will continue to 
satisfy ANSI/ANS [American National Standards Institute/American 
Nuclear Society] 3.1-1978 as required by TS 6.3.1. This change does 
not involve any physical modifications to plant structures, systems, 
or components (SSC), or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected. As the proposed change 
is administrative in nature, operation of the facility in accordance 
with the proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change to TS 6.2.2.2.j revises the requirement 
concerning the Operations management position that must hold an SRO 
license. At least one of the Operations Managers or the Senior 
Manager-Operations will continue to meet NRC requirements for 
maintaining an SRO license. The training, qualification and 
experience requirements for Operations management personnel will 
continue to satisfy ANSI/ANS 3.1-1978 as required by TS 6.3.1. This 
change does not involve any physical modifications to SSCs, or the 
manner in which SSCs are operated, maintained, modified, tested, or 
inspected. As the proposed change is administrative in nature, 
operation of the facility in accordance with the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed change to TS 6.2.2.2.j revises the requirement 
concerning the Operations management position that must hold an SRO 
license. At least one of the Operations Managers or the Senior 
Manager-Operations will continue to meet NRC requirements for 
maintaining an SRO license. If the Senior Manager-Operations does 
not hold an SRO license, then an Operations Manager must hold an SRO 
license. This individual will be qualified to fill the Senior 
Manager-Operations position and have the same management authority 
over licensed operators as the Senior Manager-Operations. In 
addition, administrative procedures will ensure that there is always 
an individual holding a current SRO license within Operations 
management. The training, qualification and experience requirements 
for Operations management personnel will continue to satisfy ANSI/
ANS 3.1-1978 as required by TS 6.3.1. This change does not involve 
any physical modifications to SSCs, or the manner in which SSCs are 
operated, maintained, modified, tested, or inspected. As the 
proposed change is administrative in nature, operation of the 
facility in accordance with the proposed amendment does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: March 29, 2001, as supplemented by 
letter dated June 27, 2001.
    Description of amendment request: The proposed amendment revises 
the pressure-temperature (P-T) limits of Technical Specification (TS) 
3.1.2 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1). The 
proposed amendment will revise the heatup, cooldown, and inservice 
hydrostatic test limitations, and the respective heatup and cooldown 
rates for the reactor coolant system (RCS). The service period for the 
new P-T limits will be for a maximum of 29 effective full power years. 
The related Bases are also revised. Sections 3.1.2.4 and 3.1.2.5 of the 
TSs are revised to remove reference to Sections V.B and V.C of Appendix 
G, of 10 CFR Part 50, as these sections no longer exist in the 
regulations. The proposed amendment also revises TS Figures 3.1-1 and 
3.1-2 to permit TMI-1 to be operated during low temperature conditions 
with two reactor coolant pumps in operation in a single loop.
    The proposed amendment also revises TS Section 3.1.12, ``Low 
Temperature Overpressure Protection (LTOP)'' setpoints. The RCS Power-
Operated Relief Valve (PORV) low setpoint is being revised to 552 psig 
as a result of the P-T limit changes, which is the error-adjusted 
maximum setpoint. The enable temperature for the PORV setpoint (TS 
3.1.12.2) and the LTOP setpoint (TS 3.1.12.1) is revised to 329 degrees 
Farenheit to be consistent with the new P-T bases. Section 3.1.12 is 
also revised to reorganize and clarify the LTOP system protection 
parameters and applicable conditions. Reference to nominal setpoint 
pressure values which do not affect the specified maximum and minimum 
setpoint values has been deleted. The related TS 3.1.12 Bases is 
revised to reflect the above changes to limits and setpoints. The Table 
of Contents page ii is revised to reflect the changes to Section 3.1.12 
and to correct a previously issued typographical error in the listed 
titles of Sections 3.4.1 and 3.4.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    These proposed Technical Specification changes were developed 
utilizing the procedures of ASME [American Society for Mechanical 
Engineers] [Boiler and Pressure Vessel Code (Code), Section] XI, 
Appendix G, in conjunction with Code Cases N-588 and N-640. Usage of 
these procedures provides compliance with the underlying intent of 
10 CFR 50 Appendix G and provides safety limits and margins of 
safety which ensure that failure of a reactor vessel will not occur.
    The proposed changes do not impact the capability of the reactor 
coolant pressure boundary (i.e., no change in operating pressure, 
materials, seismic loading, etc.) and therefore, do not increase the 
potential for the occurrence of a loss of coolant accident (LOCA). 
The changes do not modify the reactor coolant system pressure 
boundary, nor make any physical changes to the facility design, 
material, or construction standards.
    The probability of any design basis accident (DBA) is not 
affected by this change, nor are the consequences of any DBA 
affected by this change. The proposed Pressure-Temperature (P-T) 
limits, Low Temperature Overpressure (LTOP) limits and setpoints, 
and allowable operating reactor coolant pump combinations are not 
considered to be an initiator or contributor to any accident 
analysis addressed in the TMI Unit 1 UFSAR [updated final safety 
analysis report].
    The proposed changes do not adversely affect the integrity of 
the RCS such that its function in the control of radiological 
consequences is affected. Radiological off-site exposures from 
normal operation and operational transients, and faults of moderate 
frequency do not exceed the guidelines of 10 CFR [Part] 100. In 
addition, the proposed changes do not affect any fission product 
barrier. The revised PORV LTOP setpoint is established to protect 
[the] reactor coolant pressure boundary. The changes do not degrade 
or prevent the response of the PORV or safety-related systems to 
previously evaluated accidents. In addition, the changes do not 
alter any assumption previously made in the mitigation of the 
radiological

[[Page 38759]]

consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed license amendment revises the TMI Unit 1 reactor 
vessel P-T limits, LTOP limits and setpoints, and allowable 
operating reactor coolant pump combinations. Compliance with 10 CFR 
50 Appendix G, includes utilization of ASME XI, Appendix G, as 
modified by Code Cases N-588 and N-640 to meet the underlying intent 
of the regulations. The criteria of 10 CFR 50.61 remains satisfied, 
thus ensuring an adequate margin of safety for potential thermal 
shock events. The proposed limits are developed utilizing NRC-
approved methodology and conservatively account for material 
property changes as required by regulation. The design basis event 
related to this change is nonductile failure of the reactor coolant 
pressure boundary. The proposed amendment provides assurance of 
protection against nonductile failure of the reactor coolant 
pressure boundary for operation of 29 Effective Full Power Years 
(EFPY) and is unrelated to the possibility of creating a new or 
different kind of accident. The proposed amendment does not 
introduce any new systems or components, or create any new component 
failure modes. Sufficient pressure margin is maintained to 
accommodate the proposed change to the allowable operating reactor 
coolant pump combinations.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed Technical Specification (TS) changes were developed 
utilizing the procedures of ASME XI, Appendix G, in conjunction with 
Code Cases N-588 and N-640. Usage of these procedures provides 
compliance with the underlying intent of 10 CFR 50 Appendix G and 
provides safety limits and margins of safety which ensure that 
failure of a reactor vessel will not occur.
    No plant safety limits, set points, or design parameters are 
adversely affected. The fuel, fuel cladding, and Reactor Coolant 
System are not impacted.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendments request: June 26, 2001.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications to support a modification that 
would install a digital Power Range Neutron Monitoring (PRNM) system. 
The modification would supersede plant modifications previously 
installed in support of Carolina Power & Light Company's implementation 
of Enhanced Option I-A, and will allow full implementation of the 
Boiling Water Reactor Owners Group (BWROG) Option III Reactor Stability 
Long-Term Solution.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change will replace the currently installed and NRC 
approved Enhanced Option I-A long-term stability solution, which 
prohibits operation in areas with the potential for instability, 
with an NRC approved Option III long-term stability solution. The 
PRNM hardware meets the General Design Criteria (GDC) 10 and 12 
requirements by automatically detecting and suppressing design basis 
thermal-hydraulic oscillations prior to exceeding the fuel Minimum 
Critical Power Ratio (MCPR) Safety Limit. The accident probability 
will not change since the instability is suppressed prior to 
exceeding the MCPR Safety Limit, the solution has defense-in-depth 
features, and is of robust design. In addition, the PRNM system does 
not interact with equipment whose failure could cause an accident, 
and compliance is retained for regulatory criteria established for 
PRNM system and associated plant equipment. Scram setpoints in the 
PRNM system will be established so that analytical limits are met. 
The reliability of the new system will meet or exceed that of the 
existing system and, as a result, the scram reliability will be 
equal to or better than the existing system. No new challenges to 
safety-related equipment will result from the PRNM system.
    Proper operation of the PRNM system does not affect any fission 
product barrier or Engineered Safety Feature. Thus, the proposed 
change cannot change the consequences of any accident previously 
evaluated. As stated above, the PRNM system meets the requirements 
of GDC 10 and 12 by automatically detecting and suppressing design 
basis thermal-hydraulic oscillations prior to exceeding the fuel 
MCPR Safety Limit.
    Based on the above, the operation of the new PRNM system and 
replacement of the currently installed Enhanced Option I-A stability 
solution with the Option III Oscillation Power Range Monitor (OPRM) 
function will not increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The components of the PRNM system will be supplied to equivalent 
or better design and qualification criteria than is currently 
required for the plant. Equipment that could be affected by the PRNM 
system has been evaluated. No new operating mode, safety-related 
equipment lineup, accident scenario, system interaction, or 
equipment failure mode was identified. Therefore, the PRNM system 
will not adversely affect plant equipment.
    The current plant design using the Enhanced Option I-A long-term 
stability solution depends on prohibited operating regions with an 
automatic scram if the exclusion region of the power/flow map is 
entered and an automatic rod block if the restricted region of the 
power/flow map is entered. The current design also relies on 
operator action to manually scram the plant if automatic monitoring 
of neutron flux through the period based detection system (PBDS) 
provides an instability alarm when in a region that has a potential 
for instability. The modification implementing PRNM replaces these 
automatic and manual requirements with a fully automatic detect and 
suppress capability to assure that instability events that occur 
will be terminated before the MCPR Safety Limit is exceeded. The 
``scram and rod block enforced'' restrictions on the operating 
region are relaxed. Potential failures in the OPRM Upscale function 
could result in either failure to take the required mitigating 
action or an unintended reactor scram, which are the same potential 
effects of failure of the currently installed Enhanced Option I-A 
functions.
    The PRNM modification and associated changes to the Technical 
Specifications involve equipment that is designed to detect the 
symptoms of certain events or accidents and initiate mitigating 
actions. The worst [case] failure of the equipment involved in the 
modification is a failure to initiate mitigating action (i.e., scram 
or rod block), but no failure can cause an accident of a new or 
different kind than any previously evaluated.
    Based on the above, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 38760]]

    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The current safety analyses assume that the existing Enhanced 
Option I-A related Technical Specification requirements are adequate 
to prevent an instability event. PBDS is provided as part of the 
design to detect and suppress an instability event as a defense-in-
depth feature. As a result, there is currently no impact on the MCPR 
Safety Limit identified for an instability event.
    The Option III OPRM trip function is being implemented to fully 
automate the detection, via direct measurement of neutron flux, and 
subsequent suppression, via scram, of an instability event prior to 
exceeding the MCPR Safety Limit. Other OPRM trip features (i.e., 
Growth and Amplitude Algorithms) are provided as part of a robust 
design and defense-in-depth feature for unanticipated oscillations. 
Currently, the MCPR Safety Limit is not challenged by an instability 
event since the event is prevented by automatic means or mitigated 
by automatic and manual means via the Enhanced Option I-A functions. 
In both methods the margin of safety associated with the MCPR Safety 
Limit is maintained.
    Other changes such as setpoint revisions, removing the Average 
Power Range Monitor Downscale function from the Reactor Protection 
System trip logic, removing the number of operable Local Power Range 
Monitors from the automatic trip logic, and lengthening the 
Surveillance Requirement frequencies are shown to be acceptable, as 
documented in licensing topical report (LTR) NEDC-32410P-A, 
``Nuclear Measurement Analysis and Control Power Range Neutron 
Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip 
Function,'' October 1995, and LTR NEDC-32410P-A Supplement 1, 
``Nuclear Measurement Analysis and Control Power Range Neutron 
Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip 
Function,'' November 1997. Both of these LTRs have been reviewed and 
approved by the NRC.
    Based on the above, the proposed change will not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: May 7, 2001, as supplemented on June 29, 
2001.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.4.3 and the associated Surveillance 
Requirements (SR) to eliminate the pressurizer water volume value in 
the specification and change ``volume'' to ``level'' in TS 3.4.3 and SR 
4.4.3.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not affect operations of the Reactor 
Coolant System (RCS) components. The proposed change is 
administrative in nature in that it deletes a value from the TS that 
is not used as a control limit since it cannot be monitored 
directly. Instead, pressurizer level is used as the control 
parameter and level can be monitored. The volume specified in the 
current TS is redundant information to the level limit in the 
specification. The specification is made consistent with the 
Improved Technical Specifications [ITS] with this change. The ITS 
only identify a limit for percent pressurizer level. No change to 
the HNP TS for the pressurizer level value is being proposed.
    Therefore, the proposed amendment does not involve a significant 
increase in 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 proposed changes do not involve new plant components or 
procedures, but only removes a value for volume in the pressurizer 
which is essentially redundant to the percent level indication and 
not a directly monitorable parameter for plant operation. These 
changes are administrative in nature and do not place SSCs 
[structures, systems, and components] in conditions outside of their 
design basis. There is no revision to operating setpoints or 
conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes to the pressurizer level TS and associated 
bases only remove unnecessary information from the specification. 
The information is not needed for plant operation and control. The 
deletion of this information represents an administrative change 
only since no change to the maximum level setpoint or operational 
limit is being made. The effect of this change is to make the plant 
TS consistent with the current ITS with no change to the margin of 
safety as described in the TS.

    Therefore, the proposed change does not involve a reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power and Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Patrick M. Madden, Acting.

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

    Date of amendment request: April 11, 2001, as supplemented June 14, 
2001.
    Description of amendment request: The proposed amendment would 
update the list of documents describing the analytical methods used to 
determine the core operating limits specified in Technical 
Specification (TS) 6.9.1.8b. Specifically, these changes would update 
the documents describing the analytical methods used in the current 
Small Break Loss of Coolant Accident analysis (SBLOCA), setpoint 
methodology, and non-LOCA methodology. In addition, the revision number 
and the date of documents listed in TS 6.9.1.8b would be deleted, in a 
manner consistent with that approved by the NRC in Standard Technical 
Specification Change Traveler TSTF-363.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration in 
their April 11, 2001, application. However, the NRC staff found that 
the licensee's no significant hazards consideration was not fully 
supported. In response to the staff's request, the licensee submitted a 
revised no significant hazards consideration on June 14, 2001, which is 
presented below:


[[Page 38761]]


    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change in document 6 and the deletion of document 7 
of Technical Specification 6.9.1.8b are made to identify the most 
recent, Nuclear Regulatory Commission (NRC) approved, model used in 
Small Break Loss of Coolant Accident (SBLOCA) applications. This 
methodology meets the requirements of 10 CFR 50.46 and 10 CFR 50 
Appendix K. This change has no impact on plant equipment operation. 
Since the change only affects the SBLOCA analysis, it cannot affect 
the likelihood or consequences of accidents. Therefore, this change 
will not increase the probability or consequences of an accident 
previously evaluated.
    The proposed change in document 15 (renumbered 14) of Technical 
Specification 6.9.1.8b is made to identify the most recent, NRC 
approved, setpoint methodology for Combustion Engineering type 
reactors. This change has no impact on plant equipment operation. 
The proposed change does not revise any setpoints assumed in the 
accident analyses. Therefore, it cannot affect the likelihood or 
consequences of accidents. Therefore, this change will not increase 
the probability or consequences of an accident previously evaluated.
    The proposed change to add a new document as 6.9.1.8b.15 is 
required to identify the most recent Non-LOCA methodology to be used 
in the Millstone Unit No. 2 Non-LOCA analysis. The use of this 
methodology will demonstrate that the acceptance criteria for Non-
LOCA events are met. This change has no impact on plant equipment 
operation. The change does not affect the acceptance criteria for 
Non-LOCA accident[s]. Therefore, it cannot affect the likelihood or 
consequences of accidents. Therefore, this change will not increase 
the probability or consequences of an accident previously evaluated.
    Deleting the revision number and the date from the documents 
contained in sections 6.9.1.8b.1 through 6.9.1.8b.15 has no impact 
on the actual analytical methods used to determine the core 
operating limits, nor does it have impact on the calculations 
performed for current or future reloads. This change is 
administrative in nature. This change has no impact on plant 
equipment operation nor does it affect the likelihood or 
consequences of accidents. Therefore, this 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 will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. These 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.
    The proposed changes have no impact on plant equipment 
operation. The proposed changes do not revise any setpoints assumed 
in the analyses and do not affect the acceptance criteria for Non-
LOCA accidents. Therefore, the proposed changes will not result in 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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, 
Connecticut 06385.
    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: May 31, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.5.6.b, ``Technical Specification (TS) 
Bases Control Program,'' to provide consistency with the changes to 10 
CFR 50.59 which were published in the Federal Register (64 FR 53582) on 
October 4, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The current Bases Control Program allows the licensee to make 
changes to the Technical Specification Bases that do not modify the 
Technical Specification requirements and which are allowed without 
prior NRC approval via 10 CFR 50.59. The proposed change does not 
modify these requirements and is administrative in nature. The 
revised change modifies the wording of the Bases Control Program to 
be consistent with the revised 10 CFR 50.59 program. The evaluation 
requirements of 10 CFR 50.59 will ensure that changes to the 
Technical Specification Bases will not result in more than a minimal 
increase in the probability or consequences of an accident without 
NRC prior review and approval. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    The current Bases Control Program allows the licensee to make 
changes to the Technical Specification Bases that do not modify the 
Technical Specification requirements and which are allowed without 
prior NRC approval via 10 CFR 50.59. The proposed change does not 
modify these requirements and is administrative in nature. The 
revised change modifies the wording of the Bases Control Program to 
be consistent with the revised 10 CFR 50.59 program. The evaluation 
requirements of 10 CFR 50.59 will ensure that changes to the 
Technical Specification Bases will not result in a new or different 
kind of accident than any previously evaluated in the final safety 
analysis report without NRC prior review and approval. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The current Bases Control Program allows the licensee to make 
changes to the Technical Specification Bases that do not modify the 
Technical Specification requirements and which are allowed without 
prior NRC approval via 10 CFR 50.59. The proposed change does not 
modify these requirements and is administrative in nature. The 
revised change modifies the wording of the Bases Control Program to 
be consistent with the revised 10 CFR 50.59 program. The evaluation 
requirements of 10 CFR 50.59 will ensure that changes to the 
Technical Specification Bases will not result in significant 
reduction in the margin of safety without NRC prior review and 
approval. This change is administrative in nature based on the 
amending of 10 CFR 50.59. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

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

    Date of amendment request: May 30, 2001.
    Description of amendment request: The proposed amendments would 
change the Technical Specification (TS) Surveillance Requirement (SR) 
3.6.1.1.3 and add two new SRs, SR 3.6.1.1.4 and SR 3.6.1.1.5, covering 
the testing of Suppression Chamber-Drywell Vacuum Breakers. The 
proposed changes will decrease the frequency of the Drywell-to-
Suppression Chamber bypass leakage

[[Page 38762]]

test while maintaining the current leakage test frequency for the 
Suppression Chamber-Drywell Vacuum Breakers, and establish new leakage 
acceptance criteria for the Suppression Chamber-Drywell Vacuum Breakers 
when the valves are tested individually. The proposed TS changes are 
similar to TS changes approved for Susquehanna Steam Electric Station 
on September 6, 1996.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes modify Technical Specifications (TS) 
Surveillance Requirement (SR) 3.6.1.1.3 and add two new SRs, SR 
3.6.1.1.4 and SR 3.6.1.1.5. The proposed changes will decrease the 
frequency for the Drywell-to-Suppression Chamber bypass leakage test 
while maintaining the current leakage testing frequency for the 
Suppression Chamber-Drywell Vacuum Breakers, and establish new 
leakage acceptance criteria for the Suppression Chamber-Drywell 
Vacuum Breakers when the valves are tested individually.
    The performance of a Drywell-to-Suppression Chamber bypass 
leakage test or Suppression Chamber-Drywell Vacuum Breaker leakage 
test is not a precursor to any accident previously evaluated. Thus, 
the proposed changes to the performance of the leakage tests do not 
have any effect on the probability of an accident previously 
evaluated.
    The performance of a Drywell-to-Suppression Chamber bypass 
leakage test or a Suppression Chamber-Drywell Vacuum Breaker test 
does provide assurance that the containment will perform as 
designed. Thus, the radiological consequences of any accident 
previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of an accident from any accident previously evaluated?
    The proposed changes to SR 3.6.1.1.3, SR 3.6.1.1.4, and SR 
3.6.1.1.5 do not affect the assumed accident performance of any 
LaSalle County Station structure, system or component previously 
evaluated. The proposed changes do not introduce any new modes of 
system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The current frequency associated with a Drywell-to-Suppression 
Chamber bypass leakage test in SR 3.6.1.1.3 is 24 months or 12 
months if two consecutive tests fail and continues at this frequency 
until two consecutive tests pass. The proposed SR change will modify 
the leakage test frequency to be consistent with the Primary 
Containment Leakage Rate Testing Program for Type A Tests, or 48 
months following one test failure or 24 months if two consecutive 
tests fail and continues at this frequency until two consecutive 
tests pass. The proposed change in SR 3.6.1.1.3 frequency is 
acceptable as the results from previous tests show that the measured 
Drywell-to-Suppression Chamber bypass leakage at the current TS 
frequency has been a small percentage of the allowable leakage. 
Acceptability is further demonstrated by the design requirements 
applied to the primary containment components and other periodically 
performed primary containment inspections.
    The proposed SR 3.6.1.1.4 will establish a leakage test 
frequency of 24 months for each Suppression Chamber-Drywell Vacuum 
Breaker except when the leakage test of SR 3.6.1.1.3 has been 
performed within 24 months. SR 3.6.1.1.4 specifies a leakage limit 
for each Suppression Chamber-Drywell Vacuum Breaker of less than or 
equal to 12% of the bypass leakage limit of TS 3.6.1.1.3. The 
proposed SR 3.6.1.1.5 will establish a total leakage limit of less 
than or equal to 30% of the bypass leakage limit of SR 3.6.1.1.3 
when the Suppression Chamber-Drywell Vacuum Breakers are tested in 
accordance with SR 3.6.1.1.4. The proposed changes to establish 
leakage limits for the Suppression Chamber-Drywell Vacuum Breakers 
are acceptable as demonstrated by the results from previous 
Suppression Chamber-Drywell Vacuum Breaker leakage tests that show 
the measured leakage has been a small percentage of the allowable 
leakage.
    Thus, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: April 23, 2001.
    Description of amendment request: The proposed change would delete 
the loose parts monitoring system (LPMS) and the associated Technical 
Specifications (TSs) and Bases currently contained in the Limerick 
Generating Station, Units 1 and 2 Technical Specifications. The 
licensee bases its proposal to delete the LPMS on the conclusions of 
the Boiling Water Reactor Owners' Group Topical Report NEDC-32975P, 
``Regulatory Relaxation for BWR Loose Parts Monitoring Systems''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This Technical Specification (TS) Change Request will delete the 
Loose Parts Monitoring System and the associated Technical 
Specifications and Bases currently contained in the Limerick 
Generating Station (LGS), Units 1 and 2, Technical Specifications. 
The Loose Parts Monitoring System (LPMS) is not an accident 
initiating system. The LPMS was designed in conformance with 
Regulatory Guide 1.133 (``Loose-Parts Detection Program for the 
Primary System of Light-Water-Cooled Reactors,'' Revision 1, May 
1981), to detect and alarm for loose parts in the reactor coolant 
system. A secondary function of the system is to assist the 
operators in locating the detected loose parts. The LPMS is used for 
information purposes only and is not a safety-related system. The 
operators do not rely solely on this system or information provided 
by this system for the performance of any safety-related action. 
Review of the Updated Final Safety Analysis (UFSAR) indicates that 
this system is not relied upon by other systems for input or data. 
This is a monitoring system that does not perform any automatic or 
control functions, and is not relied upon for any accident or 
transient evaluation. The removal of the LPMS from operation will 
not increase the need for operator intervention or increase operator 
burden to support any system used to mitigate an accident under 
normal or off normal conditions. Therefore, the proposed changes 
will not significantly increase the probability of an accident 
previously evaluated.
    The removal of the LPMS will not change or degrade the physical 
barriers or systems designed to contain radiation, and will have no 
affect on the on-site or off-site radiological conditions. 
Therefore, the proposed TS changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This TS Change Request will delete the Loose Parts Monitoring 
System and the associated Technical Specifications and

[[Page 38763]]

Bases currently contained in the LGS, Units 1 and 2, Technical 
Specifications. Removal of this system will not create a new mode of 
operation of the plant. The LPMS is a nonsafety-related monitoring 
system. The proposed changes do not create a system-level failure 
mode different than those that already exist. In addition, there are 
no operation or failure modes of the LPMS that are accident 
initiators. Therefore, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    This TS Change Request does not affect any safety limits or 
analytical limits. Also there are no changes to accident or 
transient core thermal hydraulic conditions, or fuel or reactor 
coolant boundary design limits, as a result of these proposed 
changes. 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: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of amendment request: March 28, 2001.
    Description of amendment request: The proposed amendment request 
for BVPS-1 would increase the limits for boron concentration in the 
Refueling Water Storage Tank (RWST) and in the Reactor Coolant System 
(RCS) accumulators. This proposed license amendment would also revise 
the limits and associated surveillance requirements on boron 
concentration in the Boron Injection Tank (BIT) to be consistent with 
the limits specified for the RWST. This proposed amendment would revise 
the RCS minimum boron concentration limit for Mode 6 to make it 
consistent with the RWST boron concentration limit.
    The increase in the boron concentration limits in the RWST and 
Accumulators is needed to address higher reactor core reactivity levels 
associated with core operation with higher plant capacity factors. The 
RCS boron concentration limit in Mode 6 during refueling needs to be 
revised whenever the RWST/Accumulator minimum boron concentration limit 
is adjusted, for consistency. Boron concentration above the upper limit 
of the RWST are not needed in the BIT in order to satisfy applicable 
safety analyses. Therefore, revising the boron concentration limits 
removes the need to maintain associated temperature controls and their 
associated surveillance requirements on the BIT. The Note at the bottom 
of Technical Specification (TS) 3/4.5.4.1.1 is being deleted since N-1 
loop operation during Modes 1, 2 and 3 (when this Specification is 
applicable) is not permitted by TS 3/4.4.1.4.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to the BVPS Unit 1 RWST, Accumulators, BIT 
and in the RCS during Mode 6 will maintain the safety analyses 
results in Chapter 14 of the BVPS Unit 1 UFSAR [Updated Final Safety 
Analysis Report] as bounding values for all Loss of Coolant Accident 
(LOCA) and non-LOCA design basis accidents. The proposed changes do 
not invalidate the RWST, accumulators or BIT's ability to meet its 
design bases.
    Increased boron concentration limits for the RWST, Accumulators, 
BIT and RCS in Mode 6 will not increase the consequences of an 
accident previously analyzed. The increased boron concentration 
limits reduce the time to switchover from cold leg to hot leg 
recirculation, which will prevent boron precipitation in the reactor 
vessel following a LOCA. The post-LOCA long term core cooling 
minimum boron requirements have been determined to continue to be 
adequate to ensure adequate post-LOCA shutdown margin. The post-LOCA 
containment sump and containment spray pH remain within the limits 
specified in the UFSAR. All other transients either were not 
impacted or were made less severe as a result of the increased boron 
concentrations.
    The deletion of the Note in Technical Specification 3/4.5.4.1.1 
does not alter the safety analyses as evaluated in the UFSAR since 
N-1 operation is currently prohibited by Technical Specification 3/
4.4.1.4.1. With the reduced upper limit on boron concentration in 
the BIT, the controls on temperature for the BIT are eliminated 
since boron precipitation is precluded above freezing.
    Therefore, this change will not increase the probability of 
occurrence of a postulated accident or the consequences of an 
accident previously evaluated since the change would continue to 
comply with the current BVPS Unit 1 licensing basis as it relates to 
the peak cladding temperature criteria of 10 CFR Part 50, Appendix K 
and the dose limits of GDC 19 and 10 CFR Part 100.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed increase in boron concentration does not add new or 
different equipment to the facility. The proposed Technical 
Specification changes also do not alter the manner in which plant 
equipment is being operated. Although the increased boron 
concentration requires procedure changes to ensure that cold leg to 
hot leg recirculation after a LOCA occurs quicker, there are no 
changes to the methods utilized to respond to plant events. The 
proposed Technical Specification changes do not alter instrument or 
control setpoints that initiate protective or mitigative actions. 
These increased boron concentration limits are conservative and do 
not alter the RCS or Emergency Core Cooling System's ability to 
perform their design bases.
    The deletion of the Note in Technical Specification 3/4.5.4.1.1 
does not alter the safety analyses as evaluated in the UFSAR since 
N-1 operation is currently prohibited by Technical Specification 3/
4.4.1.4.1. With the reduced upper limit on boron concentration in 
the BIT, the controls on temperature for the BIT are eliminated 
since boron precipitation is precluded above freezing.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any previously evaluated 
accident [* * *]
    3. Does the change involve a significant reduction in a margin 
of safety?
    The LOCA considerations, including Peak Cladding Temperature 
calculations, containment sump and spray pH requirements, boron 
solubility requirements, cold shutdown boration requirements, post-
LOCA long term core cooling minimum boron requirements, hot leg 
recirculation switchover requirements, post-LOCA hydrogen generation 
requirements, and radiological requirements have been evaluated and 
determined to be acceptable. The acceptance criteria of all non-LOCA 
design basis accidents continue to be met.
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not adversely affect the ability 
of systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling conditions for extended periods of time.
    Based upon the above evaluations, the [change does not involve a 
significant reduction in the margin of safety [* * *]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 38764]]

    Attorney for Licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard Correia, Acting.

North Atlantic Energy Service Corporation, 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 requested amendment would 
revise the Technical Specifications (TSs) by changing Surveillance 
4.4.10 to incorporate alternative reactor coolant pump (RCP) flywheel 
inspection requirements and would make various administrative wording 
changes to TSs 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3.
    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:

    In accordance with 10 CFR 50.92, North Atlantic has concluded 
that the proposed changes do not involve a significant hazards 
consideration (SHC). The basis for the conclusion that the proposed 
changes do not involve a SHC is as follows:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    This proposed revision to TS Surveillance 4.4.10, incorporates 
alternative reactor coolant pump flywheel inspection requirements 
into TS Surveillance 4.4.10 based on Topical Report WCAP-14535A. 
WCAP-14535A provided a technical basis for the elimination of 
inspection requirements for reactor coolant pump flywheels based on 
industry data. The industry data indicated that no indications that 
would affect the integrity of flywheels were revealed during 729 
examinations of 217 flywheels at 57 plants (including Seabrook 
Station). The NRC, during their review and approval of the WCAP 
required continued inspections on a ten-year interval to protect 
against events and degradation that were not anticipated and had not 
been considered in the WCAP analysis. The proposed alternate 
inspection requirements are consistent with the conclusions of an 
NRC review and generic approval of Topical Report WCAP-14535A. Thus, 
it is concluded that the proposed revision to TS Surveillance 4.4.10 
does not significantly increase the probability of an accident. 
Additionally, the performance of reactor coolant pump flywheel 
surveillances does not increase the consequence of an accident 
previously evaluated.
    The proposed changes to TS 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 do 
not adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, and configuration of the facility or 
the manner in which the plant is operated and maintained. In 
addition, these proposed changes do not affect the manner in which 
the plant responds in normal operation, transient or accident 
conditions, nor do they change procedures related to operation of 
the plant. The proposed changes to TS 6.4.1.7.b, 6.4.2.2.d, and 
6.4.2.3 do not alter or prevent the ability of structures, systems 
and components (SSCs) to perform their intended function to mitigate 
the consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report (UFSAR). These 
proposed changes are administrative in nature and only update the 
Operation [sic] License.
    The proposed changes to TS 4.4.10, 6.4.1.7.b, 6.4.2.2.d, and 
6.4.2.3 are administrative in nature and only update the Seabrook 
Station Operating License. These proposed changes do not affect the 
source term, containment isolation or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated in the Seabrook Station UFSAR. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, it is concluded that these proposed revisions to TS 
4.4.10, 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 do not involve a 
significant increase in the probability or consequence of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    This proposed revision to TS Surveillance 4.4.10 does not change 
the operation or the design basis of any plant system or component 
during normal or accident conditions. The proposed change 
incorporates alternate inspection requirements for the reactor 
coolant pump flywheels, which were generically approved by the NRC 
for use by licensees. This change does not include any physical 
changes to the plant. The proposed changes do not change the 
function or operation of plant equipment or introduce any new 
failure mechanisms. The plant equipment will continue to respond per 
the design and analyses and there will not be a malfunction of a new 
or different type introduced by the proposed changes.
    The proposed changes to TS 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 are 
administrative in nature and only update the Seabrook Station 
Operating License. These proposed changes do not modify the 
facility, nor do they modify the manner in which the plant will be 
operated, nor do they affect the plant's response to normal, 
transient or accident conditions. The proposed changes to TS 
6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 do not introduce a new mode of 
plant operation. The plant's design and design basis are not revised 
and the current safety analyses will remain in effect and the plant 
will continue to be operated in accordance with the existing 
Technical Specifications.
    Thus, these proposed revisions to TS 4.4.10, 6.4.1.7.b, 
6.4.2.2.d, and 6.4.2.3 do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    This proposed revision to TS Surveillance 4.4.10 incorporates 
alternative reactor coolant pump flywheel inspection requirements 
into TS Surveillance 4.4.10 that are consistent with the conclusions 
of an NRC review and generic approval of Topical Report WCAP-14535A. 
The current inspection requirements of TS Surveillance 4.4.10 and 
the NRC review of WCAP-14535A were both based on the recommendations 
of Regulatory Guide 1.14. The proposed changes do not change the 
function or operation of plant equipment or affect the response of 
that equipment if it is called upon to operate. The performance 
capability of the reactor coolant pumps will not be affected. 
Reactor coolant pump reliability and availability will be unaffected 
by implementation of the proposed changes.
    The proposed changes to TS 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 are 
administrative in nature and only update the Seabrook Station 
Operating License. The safety margins established through Limiting 
Conditions for Operation, Limiting Safety System Settings and Safety 
Limits as specified in the TSs are not revised. Neither the plant 
design, nor its method of operation, are revised by these proposed 
changes. Finally, the proposed changes to TS 6.4.1.7.b, 6.4.2.2.d, 
and 6.4.2.3 do not change the physical design or the operation of 
the plant.
    Thus, it is concluded that these proposed revisions to TS 
4.4.10, 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 do not involve a 
significant reduction in a margin of safety.
    Based on the above evaluation, North Atlantic concludes that the 
proposed changes to TS 4.4.10, 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3 do 
not constitute a significant hazard.

    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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: June 18, 2001.
    Description of amendment request: The proposed amendment would 
revise the reference point for reactor vessel level instrumentation 
specifications to use instrument ``zero'' instead of ``top of active 
fuel'; simplify the Safety Limits and Limiting Safety System Settings 
to eliminate specifications that are

[[Page 38765]]

unnecessary, outdated, or redundant to other Technical Specifications 
(TSs); change the reactor coolant system pressure Safety Limit from 
1335 psig to 1332 psig to correct a minor calculation error; and make 
corresponding TS Bases changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The requested changes are administrative in nature in that they 
change instrumentation reference points, reformat sections to 
conform to current NRC guidance, or correct minor errors.
    One change involves a small conservative reduction in the 
reactor coolant system pressure limit. This change corrects a long 
standing minor discrepancy in this numerical limit.
    Another change eliminates the extra 12 inches above the top of 
active fuel currently specified in the reactor water level Safety 
Limit. It is sufficient to require that all active fuel is covered 
by water to satisfy the objective of the Safety Limit and assure the 
integrity of the fuel cladding.
    None of these changes affect the configuration or method of 
operation of any plant equipment that is used to mitigate the 
consequences of an accident, nor do they affect any assumptions or 
conditions in any of the accident analyses. Since the accident 
analyses remain bounding, their radiological consequences are not 
adversely affected.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not involve a change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor do they affect 
any assumptions or conditions in any of the accident analyses. 
Accordingly, no new failure modes have been created for any plant 
system or component important to safety nor has any new limiting 
single failure been identified as a result of the proposed changes.
    Therefore the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    One change involves a small conservative reduction in the 
reactor coolant system pressure limit. This change corrects a long 
standing minor discrepancy in the derivation of the numerical value 
of this limit of less than 0.3%. The correction is conservative.
    Another change eliminates the extra 12 inches above the top of 
active fuel currently specified in the reactor water level Safety 
Limit. The additional 12 inches of water does not significantly 
contribute to fuel cooling under plant conditions for which the 
Safety Limit would be applicable. While the change in reactor water 
level represents a less restrictive limit, the proposed numerical 
value still ensures an adequate margin for core cooling and provides 
an adequate margin for effective action. The benefits gained from 
achievement of uniformity with the reactor water level Safety Limit 
established by the NRC for plants similar to Monticello outweigh any 
negative aspects of this change.
    The remainder of the requested changes are administrative in 
nature or correct minor errors.
    Therefore, a significant reduction in the margin of safety is 
not involved in the proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: June 18, 2001.
    Description of amendment request: The proposed amendment will 
delete Technical Specification (TS) Sections 5.14.3 and 5.14.4, ``Post-
Accident Radiological Sampling and Monitoring,'' requirements to 
maintain a Post Accident Sampling System (PASS). Licensees were 
generally required to implement PASS upgrades as a result of NUREG-
0737, ``Clarification of TMI [Three Mile Island] Action Plan 
Requirements,'' and Regulatory Guide 1.97, Revision 3, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the NRC's lessons 
learned from the accident that occurred at TMI Unit 2. Requirements 
related to PASS were imposed by Order for many facilities and were 
added to or included in the TS for nuclear power reactors currently 
licensed to operate. Lessons learned and improvements implemented over 
the last 20 years have shown that the information obtained from PASS 
can be readily obtained through other means or is of little use in the 
assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in a license amendment application in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated June 18, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency

[[Page 38766]]

response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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: May 24, 2001.
    Description of amendment request: The Virgil C. Summer Nuclear 
Station (VCSNS) Technical Specifications (TS) Sections 4.2.2.2.e and g 
and 4.2.2.4.e and g would be changed to adopt a revised methodology 
that relocates the Heat Flux Hot Channel Factor FQ(z) 
penalty for increasing FQ(z) versus burnup to a table in the 
Core Operating Limits Report (COLR). Also proposed is an increase in 
the FQ(z) surveillance region to be consistent with the 
current core design and to provide assurance that the peak 
FQ(z) is monitored and evaluated near end of core life.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No.
    The proposed changes to the measurement and evaluation of the 
maximum FQ(z) will provide conservative limits for 
assuring the plant is operated in a safe and consistent manner. No 
changes are being made that could initiate an accident. The 
consequences of accidents previously evaluated are unaffected by 
these proposed changes as no change to equipment response or 
accident mitigation capabilities (including assessment capabilities) 
has occurred. The proposed changes have no impact on the principal 
safety barriers of the plant.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No.
    The proposed changes decrease the size of the core region that 
is excluded from the evaluation of peak FQ(z) and 
relocate penalties from the TS to the COLR per an approved 
methodology. No new accident scenarios, failure mechanisms or 
limiting single failures are introduced as the result of this 
proposed change. This change does not challenge the integrity or 
performance of any safety-related system.
    Therefore, the possibility of a new or different kind of 
accident is not created.
    3. Does this change involve a significant reduction in margin of 
safety?
    No.
    The proposed change relocates the penalties associated with 
measuring FQ(z) and decreases the size of the core 
regions excluded from the TS required surveillance for peak 
FQ(z). There is no effect on the availability, 
operability, or performance of the safety-related systems, 
structures, or components. The margin of safety associated with the 
acceptance criteria for any accident is unchanged. All surveillances 
will be performed at their required frequencies and with the same 
acceptance criteria, which assures the plant conditions prior to 
transients, events, and accidents [remain] within the conditions 
assumed in the safety analyses.
    The Bases of the TS are founded in part on the ability of the 
regulatory criteria being satisfied assuming limiting conditions for 
operation for various systems. Conformance to the regulatory 
criteria for operation with FMQ(z) penalty 
factor relocation and the FMQ(z) exclusion 
region changes is demonstrated, and the regulatory limits are not 
exceeded. Therefore, there is no significant reduction in the margin 
of safety resulting from the proposed changes.

    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.

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: This proposed change supports the 
inclusion of the newer versions of the process rack circuit boards into 
the response time testing elimination population. These versions of the 
cards were not included in the original Failure Modes and Effect 
Analysis performed for WCAP-14036-P-A, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 38767]]

    This change to the Technical Specifications (TS) does not result 
in a condition where the design, material, and construction 
standards that were applicable prior to the change are altered. The 
same [Reactor Trip System] RTS and [Engineered Safety Features 
Actuation System] ESFAS instrumentation is being used; the time 
response allocations/modeling assumptions in the Final Safety 
Analysis Report (FSAR) Chapter 15 analyses are still the same; only 
the method of verifying the time response is changed. The proposed 
change will not modify any system interface and could not increase 
the likelihood of an accident since these events are independent of 
this change. The proposed change will not change, degrade or prevent 
actions or alter any assumptions previously made in evaluating the 
radiological consequences of an accident described in the FSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This change does not alter the performance of process protection 
racks, Nuclear Instrumentation, and/or logic systems used in the 
plant protection systems. These systems will still have response 
time verified by test before being placed in operational service. 
Changing the method of periodically verifying instrument[s] for 
these systems (assuring equipment operability) from response time 
testing to calibration and channel checks will not create any new 
accident initiators or scenarios. Periodic surveillance of these 
systems will continue and may be used to detect degradation that 
could cause the response time to exceed the total allowance. The 
total time response allowance for each function bounds all 
degradation that cannot be detected by periodic surveillance. 
Implementation of the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for the process protection racks, Nuclear 
Instrumentation, and logic systems is modified to allow the use of 
actual test data or engineering data. The method of verification 
still provides assurance that the total system response is within 
that defined in the safety analysis, since calibration tests will 
continue to be performed and may be used to detect any degradation 
which might cause the system response time to exceed the total 
allowance. The total response time allowance for each function 
bounds all degradation that cannot be detected by periodic 
surveillance. Based on the above, it is concluded that the proposed 
change does not result in a significant reduction in margin with 
respect to plant 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., et al., Docket Nos. 50-
424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendment request: June 27, 2001.
    Description of amendment request: The proposed amendment would 
revise the Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 
Technical Specifications Surveillance Requirement (SR) 3.8.1.13 
frequency from once every 18 months (with a maximum of 22.5 months 
including the 25% grace period of SR 3.0.2) to once every 24 months 
(for a maximum of 30 months including the 25% grace period of SR 
3.0.2). The proposed change would allow SR 3.8.1.13 to be performed 
following the Diesel Generator inspection/maintenance, which is 
performed at 24-month intervals in accordance with manufacturer 
recommendations. Similarly, the frequency of SR 3.8.1.14 would be 
revised from once every 18 months to once every 24 months. The proposed 
change would allow SR 3.8.1.14 to be performed following SR 3.8.1.13.
    Bases for proposed no significant hazards consideration 
determination: As required by 10 CFR Part 50 the licensee has provided 
its analysis of the requested revision of the surveillance as an issue 
of no significant hazards consideration and is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The surveillance intervals associated with SRs 3.8.1.13 and 
3.8.1.14 have no bearing on the likelihood of any of the initiating 
events assumed for any of the accidents previously evaluated. 
Therefore, increasing the intervals for SRs 3.8.1.13 and 3.8.1.14 do 
not involve a significant increase in the probability of any 
accident previously evaluated. The operability of the emergency 
diesel generators (DGs) will continue to be demonstrated by all of 
the other surveillance requirements associated with TS Limiting 
Condition for Operation (LCO) 3.8.1 which are not affected by the 
proposed change. Endurance and margin will continue to be 
demonstrated by SR 3.8.1.13, and hot restart functional capability 
will continue to be demonstrated by SR 3.8.1.14. The only difference 
will be the increased surveillance intervals, which have been shown 
to have a minimal impact on safety in accordance with Generic Letter 
91-04. Therefore, the DGs will remain capable of performing their 
safety function as assumed in the accident analyses, and the 
proposed changes do not involve a significant increase in the 
consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not introduce any new equipment or 
create new failure modes for existing equipment. No new limiting 
single failure is created, and plant operation will not be altered. 
The DGs will remain capable of performing their safety function as 
assumed in the safety analyses. No other safety-related or 
important-to-safety equipment is affected by the proposed changes. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The operability of the emergency diesel generators (DGs) 
will continue to be demonstrated by all of the other surveillance 
requirements associated with TS Limiting Condition to Operation 
(LCO) 3.8.1 which are not affected by the proposed changes. 
Endurance and margin and hot restart functional capability will 
continue to be demonstrated by SRs 3.8.1.13 and 3.8.1.14, 
respectively. The only difference will be the increased intervals, 
which have been shown to have a minimal impact on safety in 
accordance with Generic Letter 91-04. The proposed changes are 
consistent with current regulatory guidance and licensing actions 
for increasing TS surveillance intervals to accommodate operating 
cycles that have been extended to 24 months. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1 (WBN), Rhea County, Tennessee

    Date of amendment request: May 14, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 3.3.5,

[[Page 38768]]

``Loss of Power (LOP) Diesel Generator Start Instrumentation,'' to 
increase the time delay setting of the 6.9 kV Shutdown Board degraded 
voltage relays from a nominal 6 seconds to 10 seconds. This change will 
provide the plant with operating margin by allowing additional time for 
the Class 1E Auxiliary Power System to react to projected voltage 
transients on the offsite grid. This will aid in preventing unnecessary 
challenges to the WBN Class 1E power supply due to spurious relay 
actuations which result in automatic transfer from the WBN preferred 
offsite power supply to the emergency standby diesel generators.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The degraded voltage protection relays and associated time delay 
relays provided for each of the four 6.9 kV Shutdown Boards act to 
mitigate the consequences of previously analyzed accidents by 
detecting a sustained undervoltage condition, isolating the safety 
buses from offsite power, and starting the associated diesel 
generators. This safety function and logic of the degraded voltage 
relay circuits remains unchanged. The revised time delay setpoint 
will allow automatic load tap changers on CSSTs [Common Station 
Service Transformers] C and D additional time to react to voltage 
transients on the offsite grid. This will aid in preventing 
unnecessary relay actuation and isolation from offsite power 
sources, which in turn will reduce the probability of a loss of 
offsite power to the unit due to voltages transients on the offsite 
grid. The additional four second time delay does not introduce any 
new constraints that would prevent safety equipment from performing 
its designed function. The only impact to equipment previously 
evaluated is an increase in the exposure to a degraded voltage 
condition (for the loads fed from the 6.9 kV Shutdown Boards) for a 
duration of an additional four seconds. However, the required 
safety-related equipment would continue to operate throughout the 10 
second delay. The proposed change will not contribute to any 
radiological dose during an accident. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The 6.9 kV Shutdown Power System will continue to function as 
specified in the design basis. The Class 1E loads supplied by the 
6.9 kV Shutdown Boards will continue to be available to perform 
their intended safety function during the degraded voltage 
condition. The affected 6.9 kV Shutdown Boards will satisfactorily 
recover the voltage either by: (1) Stabilization of the offsite 
power grid if the degraded voltage condition is resolved within 10 
seconds, or (2) transfer to emergency power if condition is present 
at the end of 10 seconds. There are no changes in the credible 
failure modes of the 6.9 kV Shutdown Boards (including the degraded 
voltage relays and timers) from those identified and evaluated 
previously in the FSAR [Final Safety Analysis Report]. Therefore, 
the proposed change 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 ability of Class 1E loads fed from the 6.9 kV Shutdown 
Boards to perform their safety function is not compromised by this 
change. The lower boundary dropout and the upper reset setpoint of 
the degraded voltage relays remains unchanged. Increasing the delay 
time from 6 to 10 seconds will not change the voltage recovery 
profile. Analyses has shown that all motors will have adequate 
voltage to accelerate to their rated speed within their required 
times and therefore, there is no impact on operating equipment. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: May 14, 2001.
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant Unit 1 Technical Specifications (TS) 
and TS Bases to eliminate the requirements associated with core 
alterations from those limiting condition of operations (LCOs) that 
provide safety functions to mitigate the consequences of a fuel 
handling accident. The affected specifications are LCOs 3.3.6, 3.3.7, 
3.7.10, 3.7.11, 3.9.4, 3.9.7, and 3.9.8.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision eliminates requirements associated with 
core alterations for specifications that are intended to mitigate 
the consequences of a fuel handling accident (FHA). These functions 
will not impact accident generation because their function is to 
support mitigation of accidents and they are not considered to be 
the source of a postulated accident. The removal of these actions 
affects functions that are not necessary during core alterations 
because postulated events during these activities do not have the 
potential to result in major fuel cladding damage like that assumed 
for an FHA. Therefore, there is no adverse impact to nuclear safety 
by eliminating core alteration requirements for specifications that 
provide for the mitigation of an FHA.
    The proposed revision does not adversely alter any plant 
equipment or operating practices; therefore, the probability of an 
accident is not significantly increased. In addition, the 
consequences of an accident are not significantly increased by 
eliminating core alteration requirements for specifications that 
only support the mitigation of FHAs. This is based on sufficient 
safety function capabilities being available for the mitigation of 
an FHA or other potential events that could occur during core 
alteration activities.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed allowance to eliminate core alteration requirements 
for FHA related specifications will not adversely alter plant 
functions or equipment operating practices. The proposed elimination 
of core alteration requirements will not impact accident generation 
because these functions provide for FHA mitigation and are not 
postulated to be an initiator of postulated accidents. Therefore, 
since plant functions and equipment are not adversely affected and 
the availability of FHA mitigation functions do not contribute to 
the initiation of postulated accidents, the proposed revision will 
not create a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The elimination of core alteration requirements for 
specifications that provide mitigation functions for FHAs will not 
affect the ability of these functions to perform as necessary. This 
is based on postulated events during core alteration not having the 
potential to result in fuel cladding damage that is assumed for the 
FHA and therefore, not requiring functions necessary to mitigate the 
FHA event. The proposed revision will continue to provide acceptable 
provisions for activities that could result in an FHA or events 
postulated during core alterations to maintain the necessary margin 
of safety.
    Therefore, the margin of safety provided by specifications 
required for the mitigation of

[[Page 38769]]

FHAs is not significantly reduced by the proposed allowance to 
eliminate core alterations requirements.

    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 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Patrick M. Madden, Acting.

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

    Date of amendment request: June 21, 2001.
    Description of amendment request: This amendment request proposes 
to revise the control rod block instrumentation requirements contained 
in Technical Specification (TS) 2.1.B, Figure 2.1.1, and Tables 3.2.5 
and 4.2.5. Some of the control rod block trip functions are being 
relocated to the Vermont Yankee Technical Requirements Manual and some 
of the requirements for the retained trip functions are clarified. Two 
trip functions are added to the TSs.
    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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The relocated trip functions are not assumed as initial 
conditions for, nor are they credited in the mitigation of, any 
design basis accident or transient previously evaluated. Since 
reactor operation with these revised and relocated Specifications is 
fundamentally unchanged, no design or analytical acceptance criteria 
will be exceeded. As such, this change does not impact initiators of 
analyzed events, nor the analyzed mitigation of design basis 
accident or transient events.
    More stringent requirements that ensure operability of equipment 
and purely administrative changes do not affect the initiation of 
any event, nor do they negatively impact the mitigation of any 
event. Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    None of the proposed changes affects any parameters or 
conditions that could contribute to the initiation of any accident. 
No new accident modes are created since plant operation is 
unchanged. No safety-related equipment or safety functions are 
altered as a result of these changes. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    This change does not impact plant equipment design or operation, 
and there are no changes being made to safety limits or safety 
system settings that would adversely affect plant safety as a result 
of the proposed changes. Since the changes have no effect on any 
safety analysis assumptions or initial conditions, the margins of 
safety in the safety analyses are maintained. In addition, 
administrative changes that do not change technical requirements or 
meaning, and the imposition of more stringent requirements to ensure 
operability, have no negative impact on margins of safety. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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 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].

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

    Date of application for amendment: February 14, 2000, as 
supplemented on May 3, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to correct various editorial errors and make other 
administrative changes. Specifically, the amendment makes 
administrative changes that revise: (a) Tables 3.6-1 and 4.4-1 to 
correct listing and editorial errors, (b) TS 3.8.B.10 to reflect the 
wording in 10 CFR 50.54(m)(2)(iv), (c) Figures 3.10-2 through 3.10-6 to 
remove these figures, (d) Table 4.1-1 to reflect change in level 
indication components, (e) TS 4.19.B and 6.14.1.1 to correct editorial 
errors, (f) TS 6.12.1 to reflect an organizational title change, and 
(g) TS 6.13.2 to correct a typographical error. In the May 3 letter, 
the licensee requested that the proposed changes to TS 6.12.1 regarding 
references to the current sections of 10 CFR Part 20 be withdrawn.

[[Page 38770]]

    Date of issuance: July 5, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 216.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications. Date of initial notice in Federal Register: 
February 21, 2001 (66 FR 11055).
    The May 3, 2000, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 21, 2000.
    Brief description of amendment: The amendment approves a change to 
the licensing basis to allow a 121-second delay in the timing of the 
release of fission products following design-basis accidents.
    Date of issuance: July 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 143.
    Facility Operating License No. NPF-43: Amendment revised the 
Updated Safety Analysis Report.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81914) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 12, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 20, 1999, as supplemented by 
letters dated November 29, 2000, and April 6, May 7, and June 7, 2001.
    Brief description of amendment: The amendment changes River Bend 
Station (RBS) Technical Specification (TS) 3.6.1.3, ``Primary 
Containment Isolation Valves (PCIVs),'' to allow the Inclined Fuel 
Transfer System (IFTS) primary containment isolation blind flange to be 
removed during MODES 1, 2, or 3. In its application, the RBS licensee 
stated that, with the blind flange removed and certain restrictions and 
administrative controls in place, the IFTS penetration would continue 
to be provided through implementation of these additional controls.
    Date of issuance: July 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 116.
    Facility Operating License No. NPF-47: The amendment revised the 
TS.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4273).
    The November 29, 2000, and April 6, May 7, and June 7, 2001, 
supplemental letters provided 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 July 3, 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: May 3, 2001.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) 6.8.4.d, ``Post-accident Sampling, for Waterford 
Steam Electric Station, Unit 3, and thereby eliminates the requirement 
to have and maintain the post-accident sampling system.
    Date of issuance: July 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
by February 28, 2002.
    Amendment No.: 172.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29353).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 3, 2001.
    No significant hazards consideration comments received: No.

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 application for amendments: July 26, 2000, as supplemented 
January 17, 2001, and April 17, 2001.
    Brief description of amendments: Revised the Technical 
Specifications (TSs) Index to delete reference to the BASES since, in 
accordance with 10 CFR 50.36(a), the BASES are not a part of the TSs 
required by 10 CFR 50.36, and to include a ``Technical Specification 
(TS) Bases Control Program'' in the Administrative Controls Section of 
the TS.
    Date of Issuance: July 12, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 176 and 117.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59222). The letters dated January 17, 2001, and April 17, 2001, 
contained clarifying information that did not affect the original 
proposed no significant hazards determination, or expand the scope of 
the request as noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 12, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades 
Plant, Van Buren County, Michigan

    Date of application for amendment: March 5, 2001, as revised by 
letter dated March 30, 2001
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Section 5.5.12, ``Programs and Manuals--Technical 
Specifications (TS) Bases Control Program,'' in accordance with Nuclear 
Energy Institute TS Task Force (TSTF) Standard TS Change Traveler, 
TSTF-364, ``Revision to TS Bases Control Program to Incorporate Changes 
to 10 CFR 50.59,'' Revision 0.
    Date of issuance: July 9, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 9, 2001.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: February 14, 2001.
    Brief description of amendment: The amendment makes minor revisions 
in the Ginna Station Improved Technical Specifications (ITS) format to 
allow for maintaining, viewing, and publishing

[[Page 38771]]

them with a different software package. The amendment also includes a 
revision to ITS Section 5.5.13, ``Technical Specifications (TS) Bases 
Control Program,'' to provide consistency with the changes to 10 CFR 
50.59 as published in (64 FR 53852 dated October 4, 1999).
    Date of issuance: June 26, 2001.
    Effective date: June 26, 2001.
    Amendment No.: 80.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15929).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 26, 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: April 12, 2001.
    Brief description of amendments: Revised the Technical 
Specifications (TS) and associated Bases to change the methodology and 
frequency for sampling the ice condenser ice bed (stored ice) and adds 
a new TS and associated bases to address sampling requirements for all 
ice additions to the ice bed.
    Date of issuance: July 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 269 and 259.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22033).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 12, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 17th day of July 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-18324 Filed 7-24-01; 8:45 am]
BILLING CODE 7590-01-P