[Federal Register Volume 67, Number 93 (Tuesday, May 14, 2002)]
[Notices]
[Pages 34481-34502]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-11871]


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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 April 19, 2002 through May 2, 2002. The last 
biweekly notice was published on April 30, 2002 (67 FR 21283).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room (PDR), located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By June 13, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    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

[[Page 34482]]

prehearing conference scheduled in the proceeding, but such an amended 
petition must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. 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 PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendment request: April 16, 2002.
    Description of amendment request: The proposed amendments would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 16, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

[[Page 34483]]

surveillance will further minimize possible concerns. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

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

    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: April 16, 2002.
    Description of amendment request: The proposed amendments would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 16, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

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

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

    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: April 16, 2002.
    Description of amendment request: The proposed amendments would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is

[[Page 34484]]

less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement would be added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 16, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

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

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

    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: March 28, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.7, ``Auxiliary Electrical 
Systems,'' and Section 4.6, ``Emergency Power System Periodic Tests,'' 
to relocate the requirements for the gas turbine generators to the 
Updated Final Safety Analysis Report and the plans, programs and 
procedures that document and control the credited functions of these 
systems, structures, and components. The proposed amendment would also 
delete TS 3.7.B.2.b to remove the option that allows power operation 
for up to 72 hours with a gas turbine as the only available 13.8 
kilovolt power source.
    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 of occurrence or consequences of an accident previously 
evaluated.
    The Gas Turbine Generators only provide a Licensing Basis Event 
mitigating function. There is no previously evaluated accident or 
event that is initiated by the Gas Turbine Generators or the 
associated fuel storage system. The ability of the Gas Turbine 
Generators to provide power, as a backup to the Emergency Diesel 
Generators, is not affected by the location of the description of 
their licensing basis.
    Therefore, the proposed changes do 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.
    There is no physical change to the plant. The currently existing 
gas turbine generators and associated fuel oil storage facilities 
will still be used. The only change is to relocate the limiting 
conditions for operations, surveillance requirements and associated 
bases from the Technical Specifications to other licensee controlled 
documents.
    Therefore, the proposed change does not create a new accident 
initiator or precursor, or 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 deletion of the limiting conditions for operation and 
surveillance requirements for the gas turbine generators from the 
Technical Specifications does not alter the method of operation, the 
design requirements or the current licensing basis that the gas 
turbine generators be able to power all the loads required by 10 CFR 
Part 50, Appendix R to place the plant into a safe shutdown 
condition following a fire and maintain safe shutdown for three 
days. It also does not remove the licensing basis requirement of 10 
CFR Part 50, Section 50.63, that the unit must have the capacity to 
withstand and recover from a station blackout. The current licensing

[[Page 34485]]

basis will continue to credit the gas turbine generators as the 
alternate ac (AAC) power source in the event of a station blackout 
unless modified under the control of 10 CFR Part 50, Section 50.59.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in [a] 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: April 11, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation (LCO), following a 
missed surveillance. The delay period would be extended from the 
current limit of ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is less'' to `` * * * up to 24 hours or 
up to the limit of the specified Frequency, whichever is greater.'' In 
addition, the following requirement would be added to SR 3.0.3: ``A 
risk evaluation shall be performed for any Surveillance delayed greater 
than 24 hours and the risk impact shall be managed.''
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated April 11, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in [a] margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO is met. Failure to perform 
a surveillance within the prescribed frequency does not cause 
equipment to become inoperable. The only effect of the additional 
time allowed to perform a missed surveillance on [a] margin of 
safety is the extension of the time until inoperable equipment is 
discovered to be inoperable by the missed surveillance. However, 
given the rare occurrence of inoperable equipment, and the rare 
occurrence of a missed surveillance, a missed surveillance on 
inoperable equipment would be very unlikely. This must be balanced 
against the real risk of manipulating the plant equipment or 
condition to perform the missed surveillance. In addition, parallel 
trains and alternate equipment are typically available to perform 
the safety function of the equipment not tested. Thus, there is 
confidence that the equipment can perform its assumed safety 
function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: March 8, 2002.
    Description of amendment request: The proposed amendment is 
consistent with Technical Specifications Task Force (TSTF) Standard 
Technical Specification (TS) Change Traveler TSTF-360, Revision 1 and 
TSTF-204, Revision 3 and proposes to revise TS 3.8.4, ``DC Sources--
Operating,'' TS 3.8.5, ``DC Sources--Shutdown,'' TS 3.8.6, ``Battery 
Cell Parameters,'' and TS 3.8.8, ``Inverters--Shutdown.'' The changes 
associated with TSTF-360, Revision 1, add new Required Actions and 
extend the Completion Times in TS 3.8.4 and TS 3.8.5 and also include 
the relocation to a licensee-controlled program of a number of 
Surveillance Requirements (SRs) in TS 3.8.4 and TS 3.8.6. The changes 
associated with TSTF-204, Revision 3, revise TS 3.8.5 and TS 3.8.8 to 
change requirements for DC electrical power subsystem and inverters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 34486]]

consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes revise TS 3.8.4, ``DC Sources--Operating,'' 
TS 3.8.5, ``DC Sources--Shutdown,'' TS 3.8.6, ``Battery Cell 
Parameters,'' and TS 3.8.8, ``Inverters--Shutdown.''
    TS 3.8.4, TS 3.8.5, and TS 3.8.6 have been revised to 1) add new 
Required Actions and extend the Completion Time for an inoperable 
battery charger, 2) provide alternate battery charger testing 
criteria for TS 3.8.4 and TS 3.8.5, 3) relocate to a licensee-
controlled program a number of Surveillance Requirements (SRs) in TS 
3.8.4 that perform preventive maintenance on the safety-related 
batteries, 4) relocate TS Table 3.8.6-1, ``Battery Cell Parameters 
Requirements,'' to a licensee-controlled program, 5) add to TS 3.8.6 
specific Required Actions associated with out-of-limits conditions 
for battery cell float voltage, float current, electrolyte level, 
and electrolyte temperature, and 6) add a new administrative TS 
program for the maintenance and monitoring of station batteries 
based on the recommendations of Institute of Electrical and 
Electronics Engineers (IEEE) Standard 450-1995, ``IEEE Recommended 
Practice for Maintenance, Testing, and Replacement of Vented Lead-
Acid Batteries for Stationary Applications.'' In addition, TS 3.8.5 
and TS 3.8.8 have been revised to require only one DC electrical 
power subsystem and two inverters, respectively, during shutdown 
conditions.
    The DC Sources, Battery Cell Parameters, and Inverters are not 
initiators of any accident sequence analyzed in the Byron/Braidwood 
Stations' Updated Final Safety Analysis Report (UFSAR). As such, the 
proposed changes do not involve a significant increase in the 
probability of an accident previously evaluated.
    The initial conditions of Design Basis Accident (DBA) and 
transient analyses in the Byron/Braidwood Stations' UFSAR assume 
Engineered Safety Feature (ESF) systems are operable. The AC and DC 
electrical power distribution systems are designed to provide 
sufficient capacity, capability, redundancy, and reliability to 
ensure the availability of necessary power to ESF systems so that 
the fuel, Reactor Coolant System, and containment design limits are 
not exceeded. The operability of the AC and DC electrical power 
distribution systems in accordance with the proposed TS is 
consistent with the initial assumptions of the accident analyses and 
is based upon meeting the design basis of the plant. Therefore, the 
proposed changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical alteration of 
the units. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
are no setpoints at which protective or mitigative actions are 
initiated that are affected by the proposed changes. The operability 
of the AC and DC electrical power distribution systems in accordance 
with the proposed TS is consistent with the initial assumptions of 
the accident analyses and is based upon meeting the design basis of 
the plant. These proposed changes will not alter the manner in which 
equipment operation is initiated, nor will the function demands on 
credited equipment be changed. No alteration in the procedures, 
which ensure the unit remains within analyzed limits, is proposed, 
and no change is being made to procedures relied upon to respond to 
an off-normal event. As such, no new failure modes are being 
introduced. The proposed changes do not alter assumptions made in 
the safety analyses.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes will not adversely affect operation of 
plant equipment. These changes will not result in a change to the 
setpoints at which protective actions are initiated. Sufficient DC 
capacity to support operation of mitigation equipment is ensured. 
The changes associated with the new administrative TS program will 
ensure that the station batteries are maintained in a highly 
reliable manner. The equipment fed by the AC and DC electrical power 
distribution systems will continue to provide adequate power to 
safety-related loads in accordance with analyses assumptions.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of application for amendments: March 19, 2002.
    Description of amendment request: The proposed amendment would 
allow plant operation to continue if the temperature of the Normal Heat 
Sink (NHS) exceeds the Technical Specification (TS) limit of 90  deg.F 
provided the water temperature, averaged over the previous 24-hour 
period, is at or below 90  deg.F. The proposed operational flexibility 
would only apply if the NHS temperature is between 90  deg.F and 92 
deg.F. The current action time requirements would still apply if the 
NHS temperature exceeds 92  deg.F, or if the 24-hour averaged value 
exceeds 90  deg.F. The current TS Limiting Condition for Operation 
(LCO) limit of 90  deg.F would not be changed. In addition, an 
administrative change would remove references to a temporary TS change 
which had expired on May 31, 2000. The Bases for the associated TS 
would also be modified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed changes will allow plant operation to continue if the 
temperature of the NHS exceeds the TS limit of 90  deg.F provided that: 
(1) The water temperature, averaged over the previous 24 hour period, 
is at or below 90  deg.F, and (2) the NHS temperature is less than or 
equal to 92  deg.F. This increase in NHS temperature will not affect 
the normal operation of the plant to the extent that it would make any 
accident more likely to occur. In addition, there exists adequate 
margin in the safety systems and safety-related heat exchangers to 
assure the design safety functions are met at the higher temperature.
    The proposed administrative change to remove an expired, temporary 
license amendment removes information which is no longer valid.
    Thus, the proposed changes will have no adverse effect on plant 
operation, or the availability or operation of any accident mitigation 
equipment. The plant response to the design-basis accidents will not 
change. In addition, the proposed changes can not cause an accident. 
Therefore, there will be no increase in the probability or consequences 
of an accident previously evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The proposed changes will allow plant operation to continue if the 
temperature of the NHS exceeds the TS

[[Page 34487]]

limit of 90  deg.F provided that: (1) The water temperature, averaged 
over the previous 24-hour period, is at or below 90  deg.F, and (2) the 
NHS temperature is less than or equal to 92  deg.F. This will not alter 
the plant configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. The proposed 
changes will not alter the way any structure, system, or component 
functions and will not significantly alter the manner in which the 
plant is operated. There will be no adverse effect on plant operation 
or accident mitigation equipment. The proposed changes do not introduce 
any new failure modes. Also, the response of the plant and the 
operators following a design-basis accident is unaffected by the 
changes. In addition, the NHS is not an accident initiator and the 
design-basis heat removal capability of the affected safety-related 
components is maintained at the increased NHS temperature limit. The 
proposed administrative change to remove an expired, temporary license 
amendment removes information which is no longer valid. Therefore, the 
proposed changes will not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    The proposed changes will allow plant operation to continue if the 
temperature of the NHS exceeds the TS limit of 90  deg.F provided that: 
(1) The water temperature, averaged over the previous 24-hour period, 
is at or below 90  deg.F, and (2) the NHS temperature is less than or 
equal to 92  deg.F. The licensee performed an evaluation of the safety 
systems to ensure their safety functions can be met with a NHS water 
temperature of 92  deg.F. The higher NHS temperature represents a 
slight reduction in the margins of safety in terms of these systems' 
abilities to remove accident heat loads. As part of its evaluation, 
however, the licensee verified that these safety systems will still be 
able to perform their design-basis functions.
    The proposed administrative change to remove an expired, temporary 
license amendment removes information which is no longer valid.
    The proposed changes will have no adverse effect on plant operation 
or equipment important to safety. The plant response to the design-
basis accidents will not change and the accident mitigation equipment 
will continue to function as assumed in the design-basis accident 
analysis. Therefore, there will be no significant reduction in a margin 
of safety.
    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 and 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 8, 2002.
    Description of amendment request: The amendment would revise the 
safety limit minimum critical power ratio for two-loop and single-loop 
operation.
    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 proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the safety limit for the minimum 
critical power ratio (SLMCPR) for Quad Cities Nuclear Power Station 
(QCNPS), Unit 1 such that the fuel is protected during normal 
operation and during any plant transients or anticipated operational 
occurrences.
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revised the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits will be 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criteria (i.e., that at least 99.9 percent of 
the fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences) is met. Since the 
operability of plant systems designed to mitigate any consequences 
of accidents has not changed, the consequences of an accident 
previously evaluated are not expected to increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. The proposed change does not involve 
any modifications of the plant configuration or allowable modes of 
operation. The proposed change to the SLMCPR assures that safety 
criteria are maintained for QCNPS, Unit 1.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Does the proposed change involve a significant reduction in a 
margin of safety?
    The value of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1 percent of the rods are expected to 
be in boiling transition if the MCPR limit is not violated. The 
proposed change will ensure the appropriate level of fuel 
protection. Additionally, operational limits will be established 
based on the proposed SLMCPR to ensure that the SLMCPR is not 
violated during all modes of operation. This will ensure that the 
fuel design safety criteria (i.e., that at least 99.9 percent of the 
fuel rods do not experience transition boiling during normal 
operation as well as anticipated operational occurrences) are met.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 18, 2002.
    Description of amendment request: The proposed amendment would 
revise

[[Page 34488]]

Surveillance Requirement (SR) 3.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated April 18, 2002.
    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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

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

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

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Acting Section Chief: Thomas Koshy.

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

    Date of amendment requests: April 11, 2002.
    Description of amendment requests: The proposed amendments would 
revise the Surveillance Requirements for containment leakage rate 
testing in Technical Specification (TS) 4.6.1.2 to allow a one-time 
extension of the interval between integrated leakage rate tests (ILRTs) 
from 10 to 15 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    Response: No.
    Probability of Occurrence of an Accident Previously Evaluated--
    The proposed change to extend the ILRT interval from 10 to 15 
years does not affect any accident initiators or precursors. The 
containment liner function is purely mitigative. There is no design 
basis accident that is initiated by a failure of the containment 
leakage mitigation function. The extension of the ILRT will not 
create any adverse interactions with other systems that could result 
in initiation of a design basis accident. Therefore, the probability 
of occurrence of an accident previously evaluated is not 
significantly increased.
    Consequences of an Accident Previously Evaluated--
    The potential consequences of the proposed change have been 
quantified by analyzing the changes in risk that would result from 
extending the ILRT interval from 10 to 15 years. The increase in 
risk in terms of person rem per year within 50 miles resulting from 
design basis accidents was estimated to be of a magnitude that 
NUREG-1493 indicates is imperceptible. I&M has also analyzed the 
increase in risk in terms of the frequency of large early releases 
from accidents. The increase in the large early release frequency 
resulting from the proposed extension was determined to be within 
the guidelines published in Regulatory Guide 1.174. Additionally, 
the proposed change maintains defense in depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. I&M has determined 
that the increase in conditional containment failure probability 
from reducing the ILRT frequency from 1 test per 10 years to 1 test 
per 15 years would be small. Continued containment integrity is also 
assured by the history of successful ILRTs, and the established

[[Page 34489]]

programs for local leakage rate testing and inservice inspections 
which are unaffected by the proposed change. Therefore, the 
consequences of an accident previously analyzed are not 
significantly increased.
    In summary, the probability of occurrence and the consequences 
of an accident previously evaluated are not significantly increased.

2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    The proposed change to extend the ILRT interval from 10 to 15 
years does not create any new or different accident initiators or 
precursors. The length of the ILRT interval does not affect the 
manner in which any accident begins. The proposed change does not 
create any new failure modes for the containment and does not affect 
the interaction between the containment and any other system. Thus, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The risk-based margins of safety associated with the containment 
ILRT are those associated with the estimated person-rem per year, 
the large early release frequency, and the conditional containment 
failure probability. I&M has quantified the potential effect of the 
proposed change on these parameters and determined that the effect 
is not significant. The non-risk-based margins of safety associated 
with the containment ILRT are those involved with its structural 
integrity and leak tightness. The proposed change to extend the ILRT 
interval from 10 to 15 years does not adversely affect either of 
these attributes. The proposed change only affects the frequency at 
which these attributes are verified. Therefore, the proposed changes 
do not involve a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: March 22, 2002.
    Description of amendment request: The proposed amendments change 
Seabrook Station Technical Specification (TS) 3/4.9.13, Spent Fuel 
Assembly Storage, and associated TS Figures and Index. The licensee 
will also revise the Bases to reflect the license amendment. The 
proposed changes reflect a revised criticality safety analysis 
supporting a two-zone spent fuel pool, consisting of BORAFLEX 
and Boral fuel assembly storage racks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and 
TS Figure 3.9-2 do not adversely affect accident initiators or 
precursors nor alter the design assumptions, conditions, and 
configuration of the facility. In addition, the proposed changes do not 
affect the manner in which the plant responds in normal operation, 
transient, or accident conditions. The changes reflect the design 
capability of the BORAL storage racks to safely store spent 
fuel.
    The 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 Updated Final Safety Analysis Report (UFSAR). 
Furthermore, 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, the proposed changes 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.
    The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and 
TS Figure 3.9-2 do not change the operation or the design basis of any 
plant system or component during normal or accident conditions. The 
proposed changes do not include any physical changes to the plant. In 
addition, 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 do not modify the 
facility nor do they affect the plant's response to normal, transient, 
or accident conditions. The changes do not introduce a new mode of 
plant operation. The changes reflect the design capability of the 
BORAL storage racks to safely store spent fuel. The plant's 
design and design basis are not revised and the current safety analyses 
remains in effect. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    The proposed changes to TS Index, TS 3/4.9.13, TS Figure 3.9-1, and 
TS Figure 3.9-2 do not adversely affect the safety margins established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings, and Safety Limits as specified in the Technical 
Specifications nor is the plant design revised by the proposed changes. 
The safety margins established through Limiting Conditions for 
Operation, Limiting Safety System Settings, and Safety Limits as 
specified in the Technical Specifications are not revised nor is the 
plant design or its method of operation revised by the proposed 
changes. The changes reflect the design capability of the 
BORAL storage racks to safely store spent fuel. 
Administrative control measures (e.g., procedures) will continue to be 
in place to ensure the safe placement of fuel assemblies within the 
spent fuel pool so as to remain less than or equal to 0.95 
Keff as required by TS 5.6.1.1 for spent fuel storage. 
Therefore, the proposed changes do not involve a significant reduction 
in a margin of safety.
    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: William J. Quinlan, Esq., Assistant General 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

[[Page 34490]]

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

    Date of amendment request: December 21, 2001.
    Description of amendment request: The proposed amendment would 
revise the Containment Systems Section of the Technical Specification 
(TS) to clarify existing requirements, make wording improvements, 
revise existing limiting condition for operations (LCO) and 
surveillance requirements (SR), and add an additional TS LCO to the 
Monticello TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee, Nuclear 
Management Company, LLC (NMC) has provided its analysis of the issue of 
no significant hazards consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS changes do not introduce new equipment or new 
equipment operating modes, nor do the proposed changes alter 
existing system relationships. Providing additional time to correct 
a situation in which suppression pool water level may be outside the 
established limits, deleting an unnecessary TS regarding suppression 
pool water level instrumentation, adding a time limit in which to 
restore oxygen concentration in the containment to within limits, 
and clarifying specific use and actions for Primary Containment 
Isolation Valves, are not initiators of any accident previously 
evaluated. Consequently, the probability of an accident previously 
evaluated is not significantly increased. The equipment referenced 
in the proposed changes is still required to be operable and capable 
of performing its accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected.
    Therefore, the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not involve physical alterations of the 
plant, no new or different type of equipment will be installed. Nor, 
are there significant changes in the methods governing normal plant 
operation. Providing additional time to correct a situation in which 
suppression pool water level may be outside the established limits, 
deleting an unnecessary TS regarding suppression pool water level 
instrumentation, restructuring the TS to provide clear Action 
Statements where needed; adding a time limit in which to restore 
oxygen concentration in the containment to within limits; and 
clarifying specific use and actions for Primary Containment 
Isolation Valves will not lead to an accident beyond those 
previously evaluated.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Providing additional time to correct a situation in which 
suppression pool water level may be outside the established limits, 
deleting an unnecessary TS regarding suppression pool water level 
instrumentation, restructuring the TS to provide clear Action 
Statements where needed; adding a time limit in which to restore 
oxygen concentration in the containment to within limits; and 
clarifying specific use and actions for Primary Containment 
Isolation Valves does not result in a significant reduction in the 
margin of safety. Allowing up to 2 hours to restore level, is 
acceptable because the suppression pool water level does not change 
rapidly during normal operation, and during operations that do 
create changes to the suppression pool water level, the level of the 
pool is closely monitored. The changes that provide specific LCO 
action statements for allowed time to place the reactor in a 
condition in which the LCO is no longer applicable are acceptable 
based on industry practices and engineering judgements. Adding an 
additional LCO which places a specified time limit on oxygen 
concentration greater than or equal to 4% by volume is acceptable 
because it provides a TS requirement which limits additional oxygen 
in the containment. Providing a revision to the LCO for inoperable 
primary containment isolation valves is acceptable because it 
clarifies what is specifically required for this method of 
isolation, and changing the interval at which deactivated and 
isolated valves must be recorded from daily to monthly is acceptable 
because the devices are operated under administrative controls and 
the probability of their misalignment is low. Relocating TS 
requirements is acceptable because it places the requirement for 
limiting the use of the purge and vent valves in a more appropriate 
TS and rewording the LCO is acceptable because it provides 
clarification for use of the purge and vent valves.
    Therefore, these proposed changes 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: L. Raghavan.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: March 20, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.8, ``Service Water (SW) 
System,'' which is applicable in Modes 1, 2, and 3, to allow the SW 
system to be operable with five operable SW pumps, provided one Unit is 
in Mode 5 or Mode 6, or defueled, and the SW system is capable of 
providing required cooling water flow to required equipment. The 
proposed amendment would change the existing TS requirement which now 
requires that both units be in Mode 5 (cold shutdown) within 36 hours 
if five of the total of six SW pumps are operable.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The SW System is primarily a support system for systems required 
to be operable for accident mitigation. Failures within the SW 
System are not an initiating condition for any analyzed accident.
    The SW System removes the required heat from the containment fan 
coolers and residual heat removal heat exchangers ensuring 
containment pressure and temperature profiles following an accident 
are as evaluated in the [Final Safety Analysis Report] FSAR. This in 
turn ensures that environmental qualification of equipment inside 
containment is maintained and thus function as required post-
accident. Single Unit operation with five operable SW pumps will 
continue to be capable of supplying the required cooling water flow 
to systems required for accident mitigation.
    Therefore, the consequences of an accident previously evaluated 
will not be significantly increased as a result of the proposed 
change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. The evaluation of the effects of the proposed changes 
indicate that the SW System will be able to perform all of its 
design basis functions within the design limits of the system. These 
changes do not introduce any new or different normal

[[Page 34491]]

operation or accident initiators. Therefore, operation of the SW 
System as proposed will not create any new failure mechanisms.
    Equipment important to safety will continue to operate as 
designed. The changes do not result in any event previously deemed 
incredible being made credible. The changes do not result in more 
adverse conditions or result in any increase in the challenges to 
safety systems. Therefore, operation of the Point Beach Nuclear 
Plant in accordance with the proposed amendment will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The SW System functions to mitigate the effects of accidents. 
There are no new or significant changes to the initial conditions 
contributing to accident severity or consequences. The proposed 
amendment will not otherwise affect the plant protective boundaries, 
will not cause a release of fission products to the public, nor will 
it degrade the performance of any other SSCs [structure, system and 
components] important to safety. Therefore, reducing the required 
number of operable SW pumps from six to five with one Unit in Mode 5 
or 6, or defueled, while maintaining the capability of required flow 
to required equipment, will not result in a significant reduction in 
the margin of safety.

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

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

    Date of amendment request: March 27, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 1.3.1, ``Limiting Safety Systems 
Settings, Reactor Protective System,'' to change the high power trip 
setpoint from 107.0% to 109.0%. This complies with the regulatory 
requirements in 10 CFR part 50 Appendix A, Criterion 10 and 20 by 
continuing to protect the fuel from exceeding the design basis limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The change does not result in a high power trip setpoint that 
will cause the analysis value of 112.0% to be exceeded. There is no 
change in the analysis value of 112.0% for the high power trip 
setpoint used in the evaluation of the transients and accidents. All 
of the evaluated transients and accidents currently show acceptable 
results and will not be affected by this change. Changing the high 
power trip setpoint will not affect the probability of an accident, 
since that circuit is not a transient or accident initiator. The 
change to the setpoint will not change the failure possibilities for 
this circuit. The effect of the proposed change is the reduction in 
the probability of an undesired safety system challenge initiated by 
an erroneous high power trip during a flow streaming event.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change to the RPS [reactor power system] high power trip 
setpoint does not provide the possibility of the creation of a new 
or different type of accident. Changing the setpoint does not change 
the method of operation of the high power trip circuit or its 
expected response once the setpoint is reached. The trip will occur 
within previously analyzed limits.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed setpoint change does not constitute a significant 
reduction in the margin of safety due to the fact that the transient 
and accident analyses contained in the Updated Safety Analysis 
Report have been evaluated using an analysis trip setpoint of 112.0% 
with the event initiated from the appropriate power level and have 
been shown to produce acceptable results.
    The acceptance criteria used in the analysis have been developed 
for the purpose of use in design basis accident analyses such that 
meeting these limits demonstrates adequate protection of public 
health and safety. An acceptable margin of safety is inherent in 
these licensing limits. Therefore, the proposed changes do not 
involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: March 29, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to allow the use of the pressure-
temperature curves approved in Amendment No. 131 for an additional 
cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The staff's evaluation of the licensee's analysis is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment to revise the technical specifications to 
extend the use of the pressure-temperature (P-T) limits does not affect 
the operation or configuration of any plant equipment. Thus, no new 
accident initiators are created by this change. The proposed change 
extends the use of the P-T limits for an additional cycle. The P-T 
limits are based on the projected reactor vessel neutron fluence at 32 
effective full power years (EFPY) of operation. At the end of cycle 10, 
Hope Creek Generating Station (HCGS) was at approximately 12.2 EFPY of 
operation (38.1% of the 32 EFPY). At the end of cycle 12 there will 
remain sufficient margin to ensure that the current 32 EFPY fluence 
projections will not be exceeded. This ensures that the basis for 
proposed applicability of the current P-T limits is conservative for 
use until the end of cycle 12 ensuring that the reactor vessel 
integrity is protected under all operating conditions. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed amendment revises the technical specifications to 
extend the use of the pressure-temperature (P-T) limits. It does not 
change the design function or operation of any systems, structures, or 
components. Plant operation will not be affected by the proposed 
amendments and no new failure mechanisms, malfunctions or accident 
initiators will be created. The current P-T limits will remain valid 
and conservative during the proposed extension period. The proposed 
change, therefore, does not create the possibility of a new or 
different kind of accident from any previously evaluated.

[[Page 34492]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change extends the use of the current P-T limits for 
an additional cycle of operation. The P-T limits are based on the 
projected reactor vessel neutron fluence at 32 EFPY of operation. At 
the end of cycle 10 in April 2000, HCGS was at approximately 12.2 EFPY 
of operation (38.1% of the 32 EFPY). At the end of cycle 12, HCGS will 
have obtained less than 50% of the 32 EFPY operating time which 
provides significant margin to ensure that the current 32 EFPY fluence 
projection will not be exceeded. The current margin of safety for plant 
operations is established by the P-T curves analyzed at 32 EFPY. 
Because the proposed change will not exceed this fluence, the current 
margin of safety is maintained. The proposed change, therefore, does 
not involve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: April 3, 2002.
    Description of amendment request: The proposed amendment would 
relocate parts of Technical Specification (TS) 3/4.4.4, ``Reactor 
Coolant System--Chemistry,'' from the TS to the Updated Final Safety 
Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is administrative in nature and does not 
involve the modification of any plant equipment or affect basic 
plant operation. Conductivity, chloride, and pH limits are not 
assumed to be an initiator of any analyzed event, nor are these 
limits assumed in the mitigation of consequences of accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve the modification of any 
plant equipment and does not change the method by which any safety-
related system performs its function. The current safety analysis 
assumptions are not altered as a result of this change.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change represents the relocation of current TS 
requirements to the UFSAR based on regulatory guidance and 
previously approved changes for other stations. The proposed change 
is administrative in nature, does not negate any existing 
requirement, and does not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. Margins of safety are unaffected by 
requirements that are retained but relocated from the TS to the 
UFSAR.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: March 25, 2002, as supplemented by the 
letter dated April 23, 2002.
    Brief description of amendments: The proposed change would revise 
the current Technical Specification (TS) 3.7.3 to adopt the version of 
the same TS in NUREG-1431, ``Standard Technical Specifications for 
Westinghouse Plants,'' Revision 2, to add, among other things, 
operability requirements for Feedwater Control Valves (FCV) and 
Associated Bypass Valves, and would allow for the extended out-of-
service time for one or more Feedwater Isolation valves (FIVs). In 
addition, a footnote, which allowed a one-time extension for Condition 
A Completion Time, is being deleted because it is no longer applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the Completion Time for one or more 
Feedwater Isolation Valves (FIVs) inoperable from 4 hours to 72 
hours. Extending the Completion Time is not an accident initiator 
and thus does not change the probability that an accident will 
occur. However, it could potentially affect the consequences of an 
accident if an accident occurred during the extended unavailability 
of the inoperable FIV. The increase in time that the FIV is 
unavailable is small and the probability of an event occurring 
during this time period, which would require isolation of the Main 
Feedwater flow paths, is low. Moreover, the redundancy provided by 
the Feedwater Control Valves, which have [the] same actuation 
signals and closure time requirements as the FIVs, provides adequate 
assurance that automatic feedwater isolation will occur if called 
upon.
    The deletion of the footnote, which is no longer applicable, is 
an administrative change and does not affect the probability or 
consequences of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Closure of the FIVs is required to mitigate the consequences of 
a Main Steam Line Break and Main Feedwater Line Break accidents. The 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The deletion of the footnote, which is no longer applicable, is 
an administrative change and does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not change any Technical Specification 
Limit or accident analysis assumption. Therefore they do not involve 
a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 34493]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: March 27, 2002.
    Brief description of amendments: The proposed change would revise 
Technical Specification (TS) 5.3.1 to require that each member of the 
unit staff, with the exception of Licensed Reactor Operators (RO) and 
Licensed Senior Reactor Operators (SRO), shall meet or exceed the 
minimum qualifications of Regulatory Guide 1.8, Revision 2, 1987. Also, 
a new TS 5.3.2 would be added to require that the Licensed RO and 
Licensed SRO shall meet or exceed the minimum qualifications of 
Regulatory Guide 1.8, Revision 3, May 2000 and the current TS 5.3.2 
would be renumbered to TS 5.3.3.
    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:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program. These changes conform to the 
current requirements of 10 CFR [Part] 55. The TS requirements for 
all other unit staff qualifications remain unchanged.
    Although licensed operator qualifications and training may have 
an indirect impact on accidents previously evaluated, the NRC 
[Nuclear Regulatory Commission] considered this impact during the 
rulemaking process, and by promulgation of the revised 10 CFR [Part] 
55 rule, concluded that this impact remains acceptable as long as 
the licensed operator training program is certified to be accredited 
and is based on a systems approach to training. TXU Energy's [TXU 
Generation Company LP] licensed operator training program is 
accredited by INPO [Institute of Nuclear Power Operations] and is 
based on a systematic approach to training.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
[the] licensed operator training program, and to conform to the 
revised 10 CFR [Part] 55. The TS requirements for all other unit 
staff qualifications remain unchanged.
    As noted above, although licensed operator qualifications and 
training may have an indirect impact on the possibility of a new or 
different kind of accident from any accident previously evaluated, 
the NRC considered this impact during the rulemaking process, and by 
promulgation of the revised rule, concluded that this impact remains 
acceptable as long as the licensed operator training program is 
certified to be accredited and based on a systems approach to 
training. As previously noted, TXU Energy's licensed operator 
training program is accredited by INPO and is based on a systems 
approach to training.
    Additionally, the proposed TS change does not affect plant 
design, hardware, system operation, or procedures. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS change is an administrative change to clarify 
the current requirements applicable to licensed operator 
qualifications and licensed operator training program. This change 
is consistent with the requirements of 10 CFR [Part] 55. The TS 
qualification requirements for all other unit staff remain 
unchanged.
    Licensed operator qualifications and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by promulgation of 
the revised 10 CFR [Part] 55, determined that this impact remains 
acceptable when licensees maintain a licensed operator training 
program that is accredited and based on a systems approach to 
training. As noted previously, TXU Energy's licensed operator 
training program is accredited by INPO and is based on a systems 
approach to training.
    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by INPO in their training 
accreditation program are equivalent to those put forth or endorsed 
by the NRC. As a result, maintaining an INPO-accredited, systems 
approach-based licensed operator training program is equivalent to 
maintaining [an] NRC-approved licensed operator training program 
which conform[s] with applicable NRC Regulatory Guides or NRC-
endorsed industry standards. The margin of safety is maintained by 
virtue of maintaining an INPO-accredited licensed operator training 
program.
    In addition, the NRC has recently published NRC Regulatory Issue 
Summary 2001-01, ``Eligibility of Operator License Applicants,'' 
dated January 18, 2001, ``* * * to familiarize addressees with the 
NRC's current guidelines for the qualification and training of 
reactor operator (RO) and senior operator (SO) license applicants.'' 
This document again acknowledges that the INPO National Academy for 
Nuclear Training (NANT) guidelines for education and experience, 
outline acceptable methods for implementing the NRC's regulations in 
this area.
    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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somerville County, Texas

    Date of amendment request: April 1, 2002.
    Brief description of amendments: This proposed amendment would 
include topical report ERX-2001-005, ``ZIRLOTM Cladding and 
Boron Coating Models for TXU Electric's Loss of Coolant Accident 
Analysis Methodologies,'' in the list of approved methodologies for use 
in generating the Core Operating Limits Report in Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR).'' In 
addition, the proposed change would include ZIRLOTM clad in 
the description of the fuel assemblies in TS 4.2.1, ``Fuel 
Assemblies.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Administrative changes to the Technical Specifications that do 
not affect the accident analyses cannot change the probability of an 
accident previously evaluated, nor will it increase radiological 
consequences predicted by the analyses of record. Controlling the 
use of fuel assemblies within limitations previously approved by the 
NRC [U.S. Nuclear Regulatory Commission] constrains fuel performance 
to within limits bounded by

[[Page 34494]]

existing design basis accident and transient analyses.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Use of ZIRLOTM clad fuel assemblies in accordance 
with NRC approved methodologies and of a design approved by the NRC 
ensures that their effect on core performance remains within 
existing design limits. Use of fuel assemblies whose design has been 
previously approved by the NRC is consistent with current plant 
design bases, does not adversely affect any fission product barrier, 
and does not alter the safety function of safety significant 
systems, structures and components or their roles in accident 
prevention or mitigation. Currently licensed design basis accident 
and transient analyses of record remain valid.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. This proposed change to TSs 4.2 and 5.6.5 
is bounded by existing limits on reactor operation. It leaves 
current limitations for use of fuel assemblies in place, conforms to 
plant design bases, is consistent with the safety analyses as 
accepted in the topical report, and limits actual plant operation 
within analyzed and NRC approved boundaries.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: February 15 and November 7, 2001, and 
March 1, 2002.
    Description of amendment request: The proposed amendment would 
revise paragraph d.1.j (2) in Technical Specification (TS) 5.5.9, 
``Steam Generator (SG) Tube Surveillance Program.'' The revision would 
(1) delete the requirement that all SG tubes containing an 
Electrosleeve, a Framatome proprietary process, be removed from service 
within two operating cycles following installation of the first 
Electrosleeve; (2) add the requirement that Electrosleeves will not be 
installed in the outermost periphery tubes of the SG bundles where 
potentially locked tubes would cause high axial loads; (3) revise the 
references describing electrosleeving; and (4) add the requirement that 
all sleeves with detected inside diameter (ID) flaw indications will be 
removed from service upon detection. The requirement to remove SG tubes 
containing electrosleeves in two operating cycles was incorporated in 
TS 5.5.9 in Amendment No. 132 issued May 21, 1999. The first 
Electrosleeve tube was installed in the fall of 1999 and the two-cycle 
allowance will expire in the fall of 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would remove the restriction that requires 
all steam generator tubes repaired with Electrosleeves to be removed 
from service at the end of two operating cycles following 
installation of the first Electrosleeve. This would allow all steam 
generator tubes repaired with Electrosleeves to remain in service. 
Reference 2 [licensee's letter dated October 27, 1998] concluded 
that there was no significant increase in the probability or 
consequences of an accident previously evaluated when using the 
Electrosleeve repair method. The two operating cycle restriction was 
invoked because the NRC staff concluded that the UT [ultrasonic] 
methods used to perform NDE [nondestructive examination] for 
inservice inspections of the Electrosleeved tubes could not reliably 
depth size stress corrosion cracks to ensure that structural limits 
are maintained.
    Revision 4 to topical report BAW-10219P [nonproprietary version 
is attached to the application] has addressed the concerns that 
resulted in the restriction of two operating cycles and 
consequently, the probability of an accident previously evaluated is 
not significantly increased. As a result, the consequences of any 
accident previously evaluated are not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing plant operation. Reference 2 
concluded that the use of the Electrosleeve repair method did not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated when using this method to repair 
steam generator tubes. This proposed change removes the two 
operating cycle limit for the Electrosleeved tubes based on the 
evaluations and justifications of the NDE techniques used to perform 
inservice examinations of the Electrosleeved steam generator tubes 
provided in Revision 4 of the topical report.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
an analyzed event. The margin of safety presently provided by the 
structural integrity of the steam generator tubes remains unchanged. 
Reference 2 concluded that the use of the Electrosleeve repair 
method did not involve a significant reduction in a margin of safety 
when using this method to repair steam generator tubes. The proposed 
change removes the two operating cycle limit based on the 
evaluations and justifications presented in Revision 4 of the 
topical report.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The reference to ``Reference 2'' in the criteria above is a 
reference to the licensee's letter dated October 27, 1998, and the no 
significant hazards consideration (NHSC) in that letter, which was 
published in the Federal Register (63 FR 66604) on December 2, 1998. 
This NHSC is applicable to the current application letters because it 
applies to the use of Electrosleeved steam generator tubes, the subject 
of the current application letters.
    The NRC staff published an earlier Notice of Consideration for the 
application dated February 15, 2001, in the Federal Register on March 
21, 2001 (66 FR 15931).
    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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

[[Page 34495]]

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

    Date of amendment request: February 26, 2002.
    Description of amendment request: Revise the definition of Operable 
in Technical Specification (TS) 1.0.K with respect to support system 
requirements for AC power sources. Conforming changes are made to 
specific support system TSs in Sections 3/4.5, ``Core and Containment 
Cooling Systems,'' 3/4.7, ``Station Containment Systems,'' and 3/4.10, 
``Auxiliary Electrical Power Systems,'' and associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed changes will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The revised definition of ``Operable'' redefines the AC power 
source requirements to allow either normal or emergency power available 
for equipment requiring AC power to be considered operable and provides 
conforming changes to specific supported system Technical 
Specifications. None of the proposed changes affects any parameters or 
conditions that could contribute to the initiation of any accident. The 
proposed change does not affect the ability of the AC power sources to 
perform their required safety functions nor does the proposed change 
affect the ability of the systems requiring AC power to perform their 
respective safety functions. As a result, the ability of these systems 
to mitigate accident consequences is unchanged. As such, these changes 
do not impact initiators of analyzed events, nor the analyzed 
mitigation of design basis accident or transient events.
    More stringent requirements for the inoperable AC power source 
action provisions that ensure availability of all TS required systems, 
subsystems, trains, components, and devices and the purely 
administrative changes do not affect the initiation of any event, nor 
do they negatively impact the mitigation of any event.
    The elimination of some explicit requirements to verify the 
operability of remaining equipment (i.e., to verify which TS action is 
required to be entered and taken) does not affect the initiation of any 
event, nor does it negatively impact the mitigation of any event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve any physical modification to 
the plant, change in Technical Specification setpoints, change in plant 
design basis, or a change in the manner in which the plant is operated. 
No new or different type of equipment will be installed. No safety-
related equipment or safety functions are altered as a result of these 
changes. In addition, there are no changes in methods governing normal 
plant operation. No new accident modes are created since plant 
operation is unchanged. None of the proposed changes affects any 
parameters or conditions that could contribute to the initiation of any 
accident. The changes do not introduce any new accident or malfunction 
mechanism that could create a new or different kind of accident, thus, 
no new failure mode is created. Therefore, the proposed changes will 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed changes will not involve a significant reduction in 
a margin of safety.
    The manner in which plant systems relied upon in the safety 
analyses to provide plant protection is not changed. Plant safety 
margins continue to be maintained through the limitations established 
in the Technical Specifications Limiting Conditions for Operation and 
Actions. These changes do 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 the ability of the 
plant to respond as assumed in the accident analyses 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, this change does not involve a significant reduction in 
a margin of safety.
    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.

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

    Date of amendment request: March 19, 2002.
    Description of amendment request: The proposed Technical 
Specification changes involve the removal of the existing scram 
function and Group 1 isolation valve closure functions of the Main 
Steam Line Radiation Monitors (MSLRM). An explicit requirement for 
periodic functional test and calibration of the MSLRM is added to 
maintain operability of the mechanical vacuum pump trip function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The scram and Group 1 isolation functions of the MSLRMs do not 
serve as initiators for any of the accidents evaluated in the Updated 
Final Safety Analysis Report (UFSAR). The MSLRM scram function is not 
credited in the UFSAR, and the Group 1 isolation trip function of the 
MSLRMs was only assumed in one design-basis event which was the control 
rod drop accident. Because these functions are not initiators of 
accidents, their removal does not increase the probability of 
occurrence of previously evaluated accidents.
    There is no accident analysis that relies on the high radiation 
scram of the reactor protection system and its removal has no impact on 
the consequences of accidents previously evaluated. The results of the 
control rod drop accident analysis remain within approved guidelines.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility for a new or different kind of accident 
from any previously evaluated.

[[Page 34496]]

    The proposed changes to the plant involve limited changes to 
protective circuitry, but do not involve any plant hardware changes 
that could introduce any new failure modes. The changes will not affect 
non-MSLRM scram and isolation functions. In addition, the MSLRMs will 
remain active for other trip/isolation functions, and these monitors 
will still alarm in the control room to alert operators to off-normal 
conditions. The reconstituted design-basis control rod drop accident 
analysis does not rely upon the trip functions that are being 
eliminated.
    Therefore, the removal of the Group 1 isolation valve closure and 
scram functions of the MSLRMs does not create the possibility of a new 
or different kind of accident than those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change involves the elimination of the scram and Group 
I isolation signal from the MSLRMs. Operation under the proposed change 
will not change any plant operation parameters, nor any protective 
system setpoints other than removal of these functions. The effects of 
the control rod drop accident without the MSLRM scram and isolation 
signal results in doses which remain well within 10 CFR Part 100, 
``Reactor Site Criteria,'' limits. The proposed changes will reduce the 
chances of unnecessary plant trips occurring as a result of an 
inadvertent MSLRM scram or Group I isolation.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    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.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendments request: March 28, 2002.
    Description of amendments request: This requested amendment would 
permit Virginia Electric and Power Company (VEPCO) to replace the 
existing Westinghouse fuel with Framatome ANP Advanced Mark-BW fuel at 
North Anna Power Station, Units 1 and 2. The accompanying requested 
exemptions from 10 CFR 50.44 and 10 CFR 50.46 will be processed 
separately.
    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 probability of occurrence or the consequences of an 
accident previously evaluated is not significantly increased. The 
Advanced Mark-BW fuel is very similar in design to the Westinghouse 
fuel that is being replaced in the core. The reload core designs for 
North Anna cycle will meet all applicable design criteria. [VEPCO] 
will use the NRC-approved standard reload design models and methods 
to demonstrate that all applicable design criteria and all pertinent 
licensing basis criteria will be met. Evaluations will be performed 
as part of the cycle specific reload safety analysis to confirm that 
the existing safety analyses remain applicable for operation of the 
Framatome Advanced Mark-BW fuel.
    Operation of the Advanced Mark-BW fuel will not result in a 
measurable impact on normal operating plant releases, and will not 
increase the predicted radiological consequences of accidents 
postulated in the UFSAR [Updated Final Safety Analysis Report]. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created. The Framatome 
Advanced Mark-BW fuel is very similar in design (both mechanical and 
composition of materials) to the resident Westinghouse fuel. The 
North Anna core in which the fuel operates will be designed to meet 
all applicable design criteria and ensure that all pertinent 
licensing basis criteria are met. Demonstrated adherence to these 
standards and criteria precludes new challenges to components and 
systems that could introduce a new type of accident. North Anna 
safety analyses have demonstrated in Section 6.0 of [the March 28, 
2002 submittal] that the use of Advanced Mark-BW fuel is acceptable. 
All design and performance criteria will continue to be met and no 
new single failure mechanisms will be created. The use of the 
Advanced Mark-BW fuel does not involve any alteration to plant 
equipment or procedures which would introduce any new or unique 
operational modes or accident precursors. Therefore, the possibility 
for a new or different kind of accident from any accident previously 
evaluated is not created.
    3. The margin of safety is not significantly reduced. The 
operation of Advanced Mark-BW fuel does not change the performance 
requirements on any system or component such that any design 
criteria will be exceeded. The normal limits on core operation 
defined in the North Anna Technical Specifications will remain 
applicable for the use of Advanced Mark-BW fuel. The reload core 
designs for the cycles in which the Advanced Mark-BW fuel will 
operate will specifically evaluate any pertinent differences between 
the Advanced Mark-BW fuel product and the current Westinghouse fuel 
product, including both the mechanical design differences and the 
past irradiation history. The use of Advanced Mark-BW fuel will be 
specifically evaluated during the reload design process using 
[VEPCO's] reload design models and methods approved by the NRC. 
North Anna safety analyses have demonstrated in Section 6.0 of [the 
March 28, 2002 submittal] that the use of Advanced Mark-BW fuel is 
acceptable. Therefore, the margin of safety as defined in the Bases 
to the North Anna Units 1 and 2 Technical Specifications is not 
significantly reduced.

    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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

[[Page 34497]]

made a determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: January 31, 2002, as 
supplemented on March 27, 2002.
    Brief description of amendments: The amendment allows a one-time 5-
year extension, for a total of 15 years, for the performance of the 
next Unit 1 integrated leak rate test (ILRT). The amendment also 
exempts Unit 1 from the requirement to perform a post-modification 
containment ILRT associated with the steam generator replacement.
    Date of issuance: May 1, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 252.
    Renewed Facility Operating License No. DPR-53: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7413). The March 27, 2002, supplemental letter provided clarifying 
information that did not change the scope of the original notice or the 
initial proposed no significant hazards consideration. The Commission's 
related evaluation of these amendments is contained in a Safety 
Evaluation dated May 1, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 25, 2001, as supplemented 
by letter dated January 24, 2002.
    Brief description of amendments: The amendments eliminated response 
time testing requirements for selected sensors and specified 
instrumentation loops for the Engineered Safety Features and the 
Reactor Trip System.
    Date of issuance: April 22, 2002.
    Effective date: As of the date of issuance and shall be impl
    emented within 30 days from the date of issuance.
    Amendment Nos.: 197, 190.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64290). The supplement dated January 24, 2002, provided clarifying 
information that did not change the scope of the May 25, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 20, 2001, as 
supplemented by letters dated February 14, and March 26, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications to incorporate NRC-approved Technical 
Specification Task Force (TSTF) Traveler TSTF-51, ``Revise containment 
requirements during handling irradiated fuel and core alterations,'' 
Revision 2. The amendments selectively adopted the Alternate Source 
Term specifically for a fuel handling accident and a weir gate drop 
accident at Catawba Nuclear Station, Units 1 and 2.
    Date of issuance: April 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 198/191.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7415). The supplements dated February 14, and March 26, 2002, 
provided clarifying information that did not change the scope of the 
December 20, 2001, application nor the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 23, 2002.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: December 28, 2000, as 
supplemented by letters dated February 15, April 26, June 26, and 
October 31, 2001, and March 4, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications related to controls to ensure acceptable 
margins of subcriticality in the spent fuel pools to account for 
Boraflex degradation.
    Date of Issuance: April 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 323, 323, 324.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9382). The supplements dated February 15, April 26, June 26, and 
October 31, 2001, and March 4, 2002, provided clarifying information 
that did not change the scope of the December 28, 2000, application nor 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: December 20, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to eliminate the use of the term 
``unreviewed safety question,'' and replace the word ``involve'' with 
the word ``require'' as it applies to changes made to the updated Final 
Safety Analysis Report and the TS Bases.

[[Page 34498]]

    Date of Issuance: April 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 324, 325.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2923). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 22, 2002, as supplemented 
by letter dated March 28, 2002.
    Brief description of amendment: The amendment modifies Technical 
Specification Surveillance Requirement (SR) 3.6.1.3.6 to add a footnote 
specifying that the isolation time of each main steam isolation valve 
(MSIV) include circuit response time and valve motion time until the 
next outage greater than 72 hours.
    Date of issuance: April 25, 2002.
    Effective date: April 25, 2002.
    Amendment No.: 175.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (67 FR 16767 dated April 8, 2002). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by May 8, 2002, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, consultation with the State of Washington and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated April 25, 2002.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendment: September 7, 2001 as revised 
December 17, 2001.
    Brief description of amendment: The amendment revised the Post 
Accident Monitoring Instrumentation Technical Specifications to ensure 
that licensee commitments to Regulatory Guide 1.97 are properly 
reflected.
    Date of issuance: April 25, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5328). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 21, 2001, as 
supplemented by letter dated January 31, 2002.
    Brief description of amendments: The amendments revise the reactor 
core safety limit for peak fuel centerline temperature from less than 
or equal to 4700  deg.F (i.e., the current technical specifications 
limit) to the design-basis fuel centerline melt temperature of less 
than 5080  deg.F, for unirradiated fuel, decreasing by 58  deg.F per 
10,000 Megawatt-Days per MetricTonne Uranium (MWD/MTU) burnup. 
Additionally, the licensee is allowed to irradiate four ZIRLO clad rods 
to 69,000 MWD/MTU that are currently in Byron Unit 2 reactor. The staff 
denied a portion of the amendment request regarding extending burnup 
limit up to 75,000 MWD/MTU for future lead test assembly (LTA) 
campaigns. A separate Notice of Partial Denial of Amendment to Facility 
Operating License and Opportunity for Hearing has been published in the 
Federal Register.
    Date of issuance: April 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 127 and 122.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59505). The supplemental letter dated January 31, 2002, contained 
clarifying information and did not change the initial no significant 
hazards consideration determination and did not expand the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: November 30, 2001.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less,'' to, ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: April 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 205 and 201.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7417). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 19, 2002.
    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: January 25, 2002.
    Brief description of amendments: These amendments revised the 
Technical Specifications requirement for pressure testing diesel fuel 
oil

[[Page 34499]]

system piping. The elevated pressure test will be replaced by a test at 
normal system operating conditions in accordance with the inservice 
inspection program.
    Date of Issuance: April 23, 2002.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 181 and 124.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7419). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 23, 2002.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: November 16, 2001, as 
supplemented March 12, 2002.
    Brief description of amendment: The amendment revises TS Table 3.3-
4, ``Engineered Safety Feature Actuation System Instrumentation Trip 
Setpoints.'' The changes are required as part of a planned design 
change to replace the existing 4kV offsite power transformers, loss of 
voltage relays, and degraded voltage relays with components of an 
improved design to increase the reliability of offsite power for 
safety-related equipment.
    Date of issuance: April 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 268.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001, (66 
FR 64298). The supplemental letter contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 9, 2002, as supplemented April 25, 
2002.
    Description of amendment request: The amendment revises Technical 
Specification Surveillance Requirement 4.8.2.3.c.1 for the Train AB and 
CD batteries. The amendment modifies the requirement to verify that the 
Train AB and CD battery cells, cell plates, and racks show no visual 
indication of physical damage or abnormal deterioration. The amendment 
allows batteries exhibiting damage or deterioration to be determined 
operable by an evaluation. The amendment is consistent with an NRC-
approved change to the Standard Technical Specifications for 
Westinghouse plants (NUREG 1431, Revision 1), as documented in 
Technical Specification Task Force Standard Technical Specification 
Change Traveler-38, ``Revise visual surveillance of batteries to 
specify inspection is for performance degradation.''
    Date of issuance: April 26, 2002.
    Effective date: As of the date of issuance, to be implemented 
immediately.
    Amendment No.: 249.
    Facility Operating License No. DPR-74: Amendment revise the 
technical specifications. Public comments requested as to proposed no 
significant hazards consideration (NSHC): Yes. April 25, 2002 (67 FR 
20552).
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated April 26, 
2002.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: April 1, 2002, as supplemented by 
letters dated April 10 and April 15, 2002.
    Brief description of amendment: This amendment adds an exception to 
the technical specifications to perform the surveillance test of Table 
3-2, Item 20 (Recirculation Actuation Logic Channel Functional Test) 
under administrative controls while components in excess of those 
allowed by Conditions a, b, d, and e of TS 2.3(2) are maintained 
operable by dedicated operator action and are required to be returned 
to operable status within one hour. This exception will apply only to 
the remainder of Cycle 20 and the entirety of Cycle 21.
    Date of issuance: April 19, 2002.
    Effective date: April 19, 2002, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 206.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (67 FR 16130 dated April 4, 
2002). The notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by May 6, 2002, but indicated 
that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendment, finding of exigent circumstances, consultation with the 
State of Nebraska and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated April 19, 
2002.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 14, 2001, as supplemented by 
letter dated February 13, 2002.
    Brief description of amendment: The amendment deletes technical 
specification (TS) Figures 2-1A (Reactor Coolant System (RCS)--
Temperature Limits for Heatup) and 2-1B (RCS Pressure--Temperature 
Limits for Cooldown) and replaces them with a single Figure 2-1. 
Additionally, the amendment changes the lowest service temperature from 
182  deg. F to 164  deg. F to be in compliance with Reference 4, 
American Society of Mechanical Engineers (ASME) Section III, NB-2332 
and the basis for the minimum boltup temperature to be in compliance 
with Reference 5, ASME Section XI, Appendix G. The Bases for TS 2.1 is 
being updated to reflect the use of ASME Code Case N-640 and the 
Westinghouse Electric Company/Combustion Engineering (W/CE) pressure 
temperature (P-T) limit curve methodology as applicable. Finally, based 
on the replacement of Figures 2-1A and 2-1B with a single Figure 2-1, 
the following TS are changed: 2.1.1(8), 2.1.2(1), 2.1.2(2), 2.1.2(6), 
2.1.2(6)(a), 2.1.2(6)(c), 2.1.2(6)(d), and 2.1.6(4) as they reference 
the deleted curves.
    Date of issuance: April 22, 2002.

[[Page 34500]]

    Effective date: April 22, 2002, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 207.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2928). The February 13, 2002, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 21, 2001, as supplemented by 
letter dated February 13, 2002.
    Brief description of amendment: The amendment reformats and revises 
Technical Specifications (TSs) 2.15(5) and (6), ``Instrumentation and 
Control Systems.'' The new TSs clarify the scope of the alternate 
shutdown panels (ASPs). The change resulted from a corrective action 
needed to address the regulatory requirements for the ASPs and the 
associated auxiliary feedwater panel, as documented in Licensee Event 
Report 97-002, Revision 0, dated May 14, 1997.
    Date of issuance: April 25, 2002.
    Effective date: April 25, 2002, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 208.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66470). The February 13, 2002, supplemental letter provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 25, 2002.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: November 16, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification Section 5.5.16, ``Containment Leakage Rate Testing 
Program,'' to allow a one-time extension of the 10 CFR Part 50, 
Appendix J, Type A integrated leak rate test interval from the required 
10 years to a test interval of 15 years.
    Date of issuance: April 22, 2002.
    Effective date: April 22, 2002, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 1-150; Unit 2-150.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
930). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: September 13, 2001, and 
supplemental letter dated March 14, 2002.
    Brief description of amendments: The amendments revise TS Section 
5.5.9, ``Steam Generator Tube Surveillance Program,'' to allow the 
extension of the steam generator tube W star (W*) alternate repair 
criteria (ARC) through Cycles 12 and 13. This extension will allow the 
licensee additional time to validate the W* leak rate model through 
performance of additional in-situ pressure testing of W* indications.
    Date of issuance: April 29, 2002.
    Effective date: April 29, 2002, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 1-151; Unit 2-151.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55021). The March 14, 2002, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 13, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specification Surveillance Requirement 3.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. The delay period is extended from the 
current limit of ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is less'' to ``* * * up to 24 hours or 
up to the limit of the specified Frequency, whichever is greater.'' In 
addition, the following requirement is added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    Date of issuance: April 23, 2002.
    Effective date: April 23, 2002, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2-186; Unit 3-177.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12605). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 21, 2001, as supplemented 
by letters dated October 24, 2001 and March 14, 2002.
    Brief description of amendments: The amendments revise TS 5.5.2.12, 
``Ventilation Filter Testing Program.'' Specifically, the reference to 
the American Society of Mechanical Engineers (ASME) Code N510-1989 was 
changed to the American National Standards Institute Standard N510-
1975. This change was requested to

[[Page 34501]]

ensure the clarity of the methodology used to test the Control Room 
Emergency Air Cleanup System and Post-Accident Cleanup Filter System 
High Efficiency Particulate Air (HEPA) filters. Although the test 
methodology is slightly different than that in N510-1989, the 
acceptance criteria are the same. Also, in Subsection 5.5.2.12.d the 
references to Regulatory Guide (RG) 1.52, Revision 2, and ASME N510-
1989 were deleted. This section is concerned with pressure drop testing 
across HEPA filters.
    Date of issuance: April 30, 2002.
    Effective date: April 30, 2002, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2--187; Unit 3--178.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7421). The March 14, 2002, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 30, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: June 5, 2001.
    Brief Description of amendments: The amendments revise Technical 
Specifications (TS) Surveillance Requirement 3.4.14.1 to clarify the 
frequency of performance with regard to Reactor Coolant System Pressure 
Isolation Valves in the Residual Heat Removal System flow path. Also, 
related TS Bases and editorial changes are part of this TS change.
    Date of issuance: April 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 155/147.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55025). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 19, 2001, as 
supplemented by letter dated March 11, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications to state that a representative sample of 
reactor instrumentation excess flow check valves (EFCVs) will be tested 
every 18 months such that each EFCV will be tested at least once every 
10 years. Prior to issuance of these amendments; the EFCVs were 
required to be tested every 18 months.
    Date of issuance: April 11, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 230/171.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2001 (66 FR 
57125). The supplement dated March 11, 2002, provided clarifying 
information that did not change the scope of the September 19, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: June 27, 2001, as supplemented 
by letter dated January 23, 2002.
    Brief description of amendments: The amendments revise the 
frequency for Surveillance Requirement (SR) 3.8.1.13 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 change allows this SR to be 
performed following the diesel generator inspection/maintenance, which 
is performed at a 24-month interval in accordance with the 
manufacturer's recommendations.
    Date of issuance: April 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 126, 104.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38767). The supplement dated January 23, 2002, provided clarifying 
information and reduced the scope of the June 27, 2001, application, 
but did not change the initial proposed no significant hazards 
consideration determination for this approval.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 17, 2001 (TS-366).
    Brief description of amendments: The amendments removed the low-
scram pilot air header pressure switches.
    Date of issuance: April 8, 2002.
    Effective date: As of date of issuance, to be implemented within 
120 days following completion of the Unit 2 Cycle 12 refueling outage 
scheduled for the spring 2003, and the Unit 3 Cycle 10 refueling outage 
scheduled for the spring 2002.
    Amendment Nos.: 276 and 235.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57126). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 8, 2002.
    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 amendment: November 15, 2001, as 
supplemented March 11, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) and the facility operating licenses (FOLs) to 
reflect an increase in the authorized maximum steady-state core power 
levels at the

[[Page 34502]]

Sequoyah Nuclear Plant, Units 1 and 2, from 3411 megawatts thermal 
(MWt) to 3455 MWt, an increase of approximately 1.3 percent.
    Date of issuance: April 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days for Unit 1 and 120 days for Region 2.
    Amendment Nos.: 275 and 264.
    Facility Operating License No. DPR-79: Amendment revises the TSs 
and FOLs.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64303). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 7th day of May 2002.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-11871 Filed 5-13-02; 8:45 am]
BILLING CODE 7590-01-P