[Federal Register Volume 67, Number 218 (Tuesday, November 12, 2002)]
[Notices]
[Pages 68728-68748]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-28483]



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Part II





Nuclear Regulatory Commission





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

  Federal Register / Vol. 67, No. 218 / Tuesday, November 12, 2002 / 
Notices  

[[Page 68728]]


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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 October 18, 2002, through October 31, 2002. 
The last biweekly notice was published on October 29, 2002 (67 FR 
66005).

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 Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By December 12, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29, 
2002.
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner

[[Page 68729]]

must provide sufficient information to show that a genuine dispute 
exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendment request: September 30, 2002.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) definition of containment 
integrity to ensure that all power-operated valves, relief valves, and 
check valves are included. The proposed changes would provide 
operability requirements to include the Type III containment isolation 
valves (CIVs), those valves that are in line with a containment 
isolation barrier consisting of a closed system within containment 
(e.g., main steam isolation valves (MSIVs)). The proposed amendment 
would revise the applicability of CIV operability requirements for 
those plant conditions when containment integrity applies and the 
reactor is not critical. The proposed amendment would clarify that the 
exceptions to containment integrity provided in TS 3.6.1 apply equally 
to TS 3.6.2, whenever containment integrity is required. The proposed 
amendment would incorporate provisions for intermittent manual 
operation of the CIVs under administrative controls. The proposed 
amendment would also delete TS 4.8, ``Main Steam Isolation Valves,'' 
along with the reference to TS 4.8 in Table 4.1-2, Item No. 6. This 
change would delete a monthly requirement for a partial stroke test, 
but would not affect testing performed in accordance with the American 
Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME 
Code), which the licensee states would continue to ensure operability 
of the MSIVs. The proposed changes would also revise Figure 5-1, 
``Extended Plot Plan,'' to correct inaccurate information, and Figure 
5-3, ``Gaseous Effluent Release Points and Liquid Effluent Outfall 
Locations,'' and its accompanying table to reflect the modification 
which permanently isolated the liquid outfall associated with emergency 
discharge from Three Mile Island Nuclear Station, Unit 2.
    Additional administrative and clerical changes are also included in 
the proposed TSs to delete obsolete references to TS sections that have 
been deleted, improve the consistency and clarity of the TSs, and 
revise the Bases of TS 3.1.6 to delete the setpoint range for emergency 
core cooling system cubicle leak detection and replace it with a single 
value.
    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 will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Changes to the definition of containment integrity and the 
additional operability requirements for Containment Isolation Valves 
(CIVs) provide additional requirements and add clarity to the 
Technical Specifications. The addition of a provision for permitting 
intermittent opening of normally closed CIVs or manual control of 
power-operated CIVs under administrative control is consistent with 
the Standard Technical Specifications or a similar provision in the 
current TMI Unit 1 Technical Specifications. This assures that the 
containment will be isolated if necessary in the event of an 
accident previously evaluated and offsite dose from an accident will 
not be significantly increased. The additional operability 
requirements provide additional conservatism to the technical 
specifications.
    None of the changes included with this License Amendment Request 
will result in any change to the configuration of plant components, 
affect any accident initiators associated with any accident 
previously evaluated or result in a significant increase

[[Page 68730]]

in the offsite dose consequences of accidents previously evaluated. 
The administrative changes are needed to correct errors and the 
editorial changes will improve the clarity, consistency and 
readability of the Technical Specifications and do not affect the 
intent or interpretation.
    Therefore, operation of the facility in accordance with the 
proposed amendment will 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 will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The changes associated with this proposed amendment do not 
result in any additional hardware or design changes to structures, 
systems, or components (SCCs) of the plant; nor will any of these 
changes affect the ability of an SSC to perform its design function. 
No new failure mechanisms, malfunctions, or accident initiators will 
be introduced that were not considered in the design and licensing 
basis.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    Additional operability requirements provide conservative 
improvements to the Technical Specifications. The addition of a 
provision for permitting intermittent opening of normally closed 
CIVs or manual control of power-operated CIVs under administrative 
control is consistent with the Standard Technical Specifications or 
with similar provisions in the current TMI Unit 1 Technical 
Specifications. This condition assures that the containment will be 
isolated if necessary in the event of an accident. Changes to the 
MSIV [main steam isolation valve] test requirements do not alter the 
Inservice Test requirements in accord[ance with] the American 
Society of Mechanical Engineers (ASME) [Boiler and Pressure Vessel] 
Code, which will continue to assure operability. The administrative 
changes are needed to correct errors and the editorial changes will 
improve the clarity, consistency, and readability of the Technical 
Specifications and do not affect the intent or interpretation.
    None of the changes included with this request have the 
potential to significantly reduce a margin of safety. These changes 
do not affect the design of a plant component or instrument setpoint 
so as to [a]ffect its design basis or affect the controlling 
numerical value for any parameter established in the updated final 
safety analysis report or the license.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: September 30, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 6.8.5, ``Reactor Building 
Leakage Rate Testing Program,'' to allow a one-time deferral of the 
next Type A, Containment Integrated Leak Rate Test (ILRT) from October 
2003 to no later than September 2008.
    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 Technical Specification change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specification Section 6.8.5 
(``Reactor Building Leakage Rate Testing Program'') involves a one-
time extension to the current interval for Type A containment 
testing. The current test interval of ten (10) years would be 
extended on a one-time basis to no longer than fifteen (15) years 
from the last Type A test (1993). The proposed Technical 
Specification change does not involve a physical change to the plant 
or a change in the manner in which the plant is operated or 
controlled. The reactor containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the reactor containment itself and the testing guidelines invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident. Therefore, the proposed Technical 
Specification change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications and NEI 
[Nuclear Energy Institute] 94-01. Industry experience has shown, as 
documented in NUREG-1493, that Type B and C containment leakage 
tests have identified a very large percentage of containment leakage 
paths and that the percentage of containment leakage paths that are 
detected only by Type A testing is very small. TMI, Unit 1 ILRT test 
history supports this conclusion. NUREG-1493 concluded, in part, 
that reducing the frequency of Type A containment leak tests to once 
per twenty (20) years leads to an imperceptible increase in risk. 
Therefore, the proposed Technical Specification change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The proposed Technical Specification change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed revision to [the] Technical Specifications involves 
a one-time extension to the current interval for Type A containment 
testing. The reactor containment and the testing guidelines invoked 
to periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident and do not involve the prevention or identification of 
any precursors of an accident. The proposed Technical Specification 
change does not involve a physical change to the plant or the manner 
in which the plant is operated or controlled. Therefore, the 
proposed Technical Specification change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed Technical Specification change does not involve 
a significant reduction in a margin of safety.
    The proposed revision to [the] Technical Specifications involves 
a one-time extension to the current interval for Type A containment 
testing. The proposed Technical Specification change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific guidelines 
and conditions of the Reactor Building Leakage Rate Testing Program, 
as defined in [the] Technical Specifications, exist to ensure that 
the degree of reactor building containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leakage rate limit specified by 
[the] Technical Specifications is maintained. The proposed change 
involves only the extension of the interval between Type A 
containment leakage tests. Type B and C containment leakage tests 
will continue to be performed at the frequency currently required by 
plant Technical Specifications and NEI-94-01.
    NUREG-1493 concludes that reducing the Type A Integrated Leak 
Rate Test (ILRT) testing frequency to one per twenty (20) years was 
found to lead to imperceptible increase in risk. Additionally, while 
Type B and C tests identify the vast majority (greater than 85%) of 
all potential leak paths, performance-based alternatives are 
feasible without significant risk impacts. Since leakage contributes 
less than 0.1 percent of

[[Page 68731]]

overall risk under existing guidelines, the overall effect is very 
small. The TMI, Unit 1 plant specific risk analysis supports this 
conclusion. Therefore, the proposed Technical Specification change 
does not involve a significant reduction in a margin of safety.

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

Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam 
Electric Plant, Unit 2, Brunswick County, North Carolina

    Date of amendment request: September 16, 2002.
    Description of amendment request: The proposed change revises a 
license condition, contained in Appendix B of the Technical 
Specifications, to reflect a modification to support the implementation 
of an alternative source term (AST) on Unit 2 that would ensure seismic 
ruggedness of the alternate leakage treatment (ALT) piping and 
appendages. As a result of further modification development, it has 
been determined that only one check valve will be installed (i.e., MVD-
V5009) by the Unit 2 ALT piping modification. The proposed license 
amendment revises the affected license condition to require that only 
MVD-V5009 must be added to the facility check valve program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises a license condition, added to 
Appendix B, ``Additional Conditions,'' of the Unit 2 Technical 
Specifications (TSs) in Amendment 246, which approved the 
implementation of Alternative Source Term. This license condition 
currently requires that alternate leakage treatment (ALT) path check 
valves MVD-V5008 and MDV-V5009 be included in the facility check 
valve program. Differences between the Unit 1 and Unit 2 main steam 
line isolation valve drain piping, which will be within the ALT 
pathway pressure boundary after a loss-of-coolant-accident (LOCA), 
obviate the need to install check valve MVD-V5008. This is because 
the Unit 2 steam bypass system was designed for full bypass 
capability and thus has two steam bypass chests; whereas Unit 1 has 
only one steam bypass chest. The Unit 2 design includes a drain line 
from the steam bypass chest, which ties into the same line that on 
Unit 1 was isolated post-LOCA by use of the 1-MVD-V5008 valve. 
Since, for Unit 2, the entire line is required to be seismically 
verified, up to and including the steam bypass chest, there was no 
benefit in installing the new check valve MVD-V5008 on Unit 2.
    CP&L has performed an evaluation of the Unit 2 ALT path 
modification, in accordance with the provisions of 10 CFR 50.59, and 
determined that the modification can be implemented without prior 
NRC approval. As such, the requested amendment merely aligns the 
wording of the current license condition with the design of the Unit 
2 ALT path modification. The original intent of the license 
condition was to ensure that check valves being installed as a 
result of the modification would be included in the facility check 
valve program. This intent is maintained by the proposed license 
condition. 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.
    As stated above, CP&L has performed an evaluation of the Unit 2 
ALT path modification, in accordance with the provisions of 10 CFR 
50.59, and determined that the modification can be implemented 
without prior NRC approval. The requested amendment merely aligns 
the wording of the current license condition with the design of the 
Unit 2 ALT path modification. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises a license condition, added to 
Appendix B, Unit 2 TSs in Amendment 246. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety. This license condition currently requires that ALT path 
check valves MVD-V5008 and MDV-V5009 be include in the facility 
check valve program. The proposed revision to affected Unit 2 
license condition eliminates reference to a CP&L September 27, 2001, 
submittal and the requirement to include MVD-V5008 in the facility 
check valve program. The requested amendment merely aligns the 
wording of the current license condition with the design of the Unit 
2 ALT path modification which has been evaluated, in accordance with 
the provisions of 10 CFR 50.59, and it has been determined that the 
modification can be implemented without prior NRC approval. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, CP&L concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

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

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

    Date of amendment request: September 26, 2002.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3.3.3.1, ``Monitoring 
Instrumentation, Radiation Monitoring,'' TS 3.3.4, ``Instrumentation, 
Containment Purge Valve Isolation Signal,'' TS 3.7.6.1, ``Plant 
Systems, Control Room Emergency Ventilation System,'' TS 3.9.4, 
``Refueling Operations, Containment Penetrations,'' TS 3.9.8.1, 
``Refueling Operations, Shutdown Cooling and Coolant Circulation--High 
Water Level,'' TS 3.9.8.2, ``Refueling Operations, Shutdown Cooling and 
Coolant Circulation--Low Water Level,'' and TS 3.9.15, ``Refueling 
Operations, Storage Pool Area Ventilation System.'' In addition, the TS 
Bases would be revised to address the proposed changes. The basis for 
the proposed changes is a re-analysis of the limiting design basis Fuel 
Handling Accident using an Alternative Source Term in accordance with 
Title 10 of the Code of Federal Regulations (10 CFR) section 50.67 and 
Regulatory Guide 1.183.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve the reanalysis of a Fuel Handling 
Accident (FHA)

[[Page 68732]]

in the Containment, FHA in the Spent Fuel Pool Area, and the Cask 
Drop Accident in the Spent Fuel Pool Area. The new analyses, based 
on the Alternative Source Term (AST) in accordance with 10 CFR 
50.67, will replace the existing analyses which are based on 
methodologies and assumptions derived from Regulatory Guide 1.25, 
Standard Review Plan (SRP) 15.7.4, SRP 15.7.5, and TID-14844. 
Because different methodologies are used, the new calculated doses 
are not directly comparable to the current calculated doses. If a 
consistent basis is used, it is expected that the new analyses 
assumptions in some cases result in a decrease in dose at the site 
boundary or to control room personnel and in some cases result in an 
increase in dose at the site boundary or to control room personnel. 
However, in all cases the analyses results are within the 10 CFR 
50.67 and Regulatory Guide 1.183 acceptance criteria.
    As a result of the new analyses, changes to the Technical 
Specifications are proposed which take credit for the new analyses. 
The proposed changes to the Technical Specifications modify 
requirements regarding Containment closure and Spent Fuel Pool area 
ventilation during movement of irradiated fuel assemblies in 
Containment and in the Spent Fuel Pool area. The proposed changes 
will allow Containment penetrations, including the equipment door 
and personnel airlock door, to be maintained open under 
administrative control. The proposed changes will eliminate the 
requirements for automatic closure of Containment purge during Mode 
6 fuel movement. The technical specifications associated with 
storage pool area ventilation will be deleted. These proposed 
changes do not involve physical modifications to plant equipment and 
do not change the operational methods or procedures used for the 
physical movement of irradiated fuel assemblies in Containment or in 
the Spent Fuel Pool area. As such, the proposed changes have no 
effect on the probability of the occurrence of any accident 
previously evaluated.
    The revised requirements apply only when irradiated fuel 
assemblies are being moved in Containment or the Spent Fuel Pool 
area. Previously evaluated accidents with the plant in other 
conditions including Modes 1 through Mode 5 are not impacted. The 
AST methodology is used to evaluate a FHA that is postulated to 
occur during fuel movement activities in Containment and in the 
Spent Fuel Pool area. The AST analyses follow the guidance of NRC 
Regulatory Guide 1.183 and the acceptance criteria of 10 CFR 50.67. 
The analyses demonstrate that the dose consequences meet the 
regulatory acceptance criteria.
    The FHA Analyses conservatively assume that the Containment 
building and the fuel storage building, including ventilation 
filtration systems for those buildings do not diminish or delay the 
assumed fission product release. The analysis does take credit for, 
and technical specifications enforce, the presence of 23 feet of 
water over the irradiated fuel while fuel movement activities are 
being performed. The analysis also takes credit for, and the 
technical specification bases enforce a fuel decay time of at least 
72 hours. In addition, administrative controls are put in place to 
provide for closure of Containment atmosphere boundary openings in 
the event of a FHA. Use of an alternative analysis method does not 
affect fuel parameters or the equipment used to handle the fuel. The 
above proposed changes to the Technical Specifications reflect 
assumptions made in the FHA Analyses. The other changes to the 
Technical Specifications are also consistent with the revised FHA 
Analyses. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment involves the use of an alternative 
analysis methodology for the evaluation of the dose consequences 
from a FHA that is postulated to occur in either the Containment or 
the Spent Fuel Pool area. The analysis demonstrates that Containment 
closure conditions and automatic closure of the Containment purge 
are not required to maintain dose consequence within regulatory 
limits following a postulated FHA inside Containment. Therefore, the 
new analysis supports proposed changes to requirements for 
Containment closure during movement of irradiated fuel assemblies in 
Containment. The analysis results also demonstrate that operation of 
the Spent Fuel Pool area ventilation system is not required to 
maintain dose consequences within regulatory limits following a 
postulated FHA in the Spent Fuel Pool area. The Containment closure 
components (e.g., equipment door, personnel airlock doors, and 
various Containment penetrations) and filtration systems are not 
accident initiators. The proposed changes do not involve the 
addition of new systems or components nor do they involve the 
modification of existing plant systems. The proposed changes do not 
affect the way in which a FHA is postulated to occur. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The existing dose analysis methodology and assumptions 
demonstrate that the dose consequences of a FHA are within 
regulatory limits for whole body and thyroid doses as established in 
10 CFR 100. The alternative dose analysis methodology and 
assumptions also demonstrate that the dose consequences of a FHA are 
within regulatory limits. The limits applicable to the alternative 
analysis are established in 10 CFR 50.67 in conjunction with the 
TEDE (total effective dose equivalent) acceptance criteria directed 
in Regulatory Guide 1.183. The acceptance criteria for both dose 
analysis methods have been developed for the purpose of evaluating 
design basis accidents to demonstrate adequate protection of public 
health and safety. An acceptable margin of safety is inherent in 
both types of acceptance criteria. 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Andersen (Acting).

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

    Date of amendment request: September 30, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications for the plant's reactor building 
integrity. The proposed amendment would (1) modify the surveillance 
requirement to be consistent with the design of the reactor building 
access openings, (2) modify the frequency of the surveillance 
requirement for visual inspections for the exposed interior and 
exterior surface of the reactor building, and (3) modify the 
administrative controls for the containment leakage rate testing 
program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    As required by 10 CFR 50.91(a)(1), this analysis is provided to 
demonstrate that the proposed license amendment does not involve a 
significant hazard.
    Conformance of the proposed amendment to the standards for a 
determination of no significant hazards, as defined in 10CFR50.92, 
is shown in the following:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed amendment to the Technical Specifications does 
not result in the alteration of the design, material, or 
construction standards that were applicable prior to the change. The 
proposed change will not result in the modification of any system 
interface that would increase the likelihood of an accident since 
these events are independent of the proposed change. The proposed 
amendment will not change, degrade, or prevent actions, or alter any 
assumptions previously made in evaluating the radiological 
consequences of an accident described in the [Updated Final Safety 
Analysis Report] UFSAR. Therefore, the proposed amendment does not 
result in the

[[Page 68733]]

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?
    No. This change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
No new accident causal mechanisms are created as a result of NRC 
approval of this amendment request. No changes are being made to the 
facility which should introduce any new accident causal mechanisms. 
This amendment request does not impact any plant systems that are 
accident initiators, since the containment and reactor building 
function primarily as accident mitigators.
    (3) Does the proposed change involve a significant reduction in 
margin of safety?
    No. Implementation of this amendment would not involve a 
significant reduction in the margin of safety. Margin of safety is 
related to the confidence in the ability of the fission product 
barriers to perform their design functions during and following an 
accident situation, including the performance of the containment and 
reactor building. The ability of the containment and reactor 
building to perform their design function will not be impaired by 
the implementation of this amendment at McGuire Nuclear Station. 
Consequently, no safety margins will be impacted.

Conclusion

    Based on the preceding analysis, it is concluded that the 
proposed license amendment does not involve a Significant Hazards 
Consideration Finding as defined in 10 CFR 50.92.

    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: 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, and Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, Located in Mecklenburg County, North 
Carolina and York County, South Carolina

    Date of amendment request: August 29, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for the plants direct-current 
(DC) system batteries. The Surveillance Requirements for the current TS 
for DC sources require a battery service test to be performed each 18 
months. A note provides that, on a once per 60 month frequency, the 
service test requirement may be met by performing a modified 
performance test. The TS change would remove the once per 60 month 
restriction, thus allowing the requirement for a service test to be met 
by a modified performance test that bounds the conditions of the 
service test. The licensee states that the proposed change will allow 
the use of a consistent battery testing technique in order to provide 
consistent data for trending battery performance.
    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:

    The following discussion is a summary of the evaluation of the 
change contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
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.

First Standard

    Operation of the facilities in accordance with this amendment 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The Class 1E DC 
[direct-current] power system is not an initiator to any accident 
sequence analyzed in the Updated Final Safety Analysis Report. The 
safety features of the batteries will continue to function as 
designed and in accordance with all applicable TS. The design and 
operation of the system is not being modified by this proposed 
amendment. This amendment only revise[s] the requirements for 
testing the batteries. Therefore, there will be no impact on any 
accident probabilities or consequences.

Second Standard

    Operation of the facilities in accordance with this amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated. No new accident 
causal mechanisms are created as a result of this proposed 
amendment. No changes are being made to any structure, system, or 
component which will introduce any new accident causal mechanisms. 
This amendment request does not impact any plant systems that are 
accident initiators and does not impact any safety analysis.

Third Standard

    Operation of the facilities in accordance with this amendment 
would not involve a significant reduction in a margin of safety. The 
change to the battery surveillance will ensure each station's 
batteries are maintained in a highly reliable manner. The batteries 
will continue to be tested every 18 months with the modified 
performance test enveloping the service test. The equipment powered 
by the batteries will continue to provide adequate power to safety 
related loads in accordance with analysis assumptions.
    Based on the preceding discussion, Duke Energy has concluded 
that the proposed amendment does not involve a significant hazard 
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: 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.

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

    Date of amendment request: September 19, 2002.
    Description of amendment request: The proposed amendment would 
extend the allowable outage time (AOT) for the emergency diesel 
generators (EDGs) from 72 hours to a maximum of 14 days.
    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 TS [technical specification] change does not affect 
the design, operational characteristics, function or reliability of 
the EDGs. The EDGs are not the initiators of previously evaluated 
accidents. The EDGs are designed to mitigate the consequences of 
previously evaluated accidents including a loss of offsite power. 
Extending the AOT for a single EDG would not affect the previously 
evaluated accidents since the remaining EDG supporting the redundant 
Engineered Safety Features (ESF) systems and the AACDG [alternate 
alternating current diesel generator], which has the capability to 
support either train of ESF systems, would continue to be available 
to perform the accident mitigating functions.

[[Page 68734]]

    The duration of a TS AOT is determined considering that there is 
a minimal possibility that an accident will occur while a component 
is removed from service. A risk-informed assessment was performed 
which concluded that the increase in plant risk is small and 
consistent with the guidance contained in Regulatory Guide 1.177.
    The current TS requirements ensure that redundant systems 
relying on the remaining EDG are operable. In addition to these 
requirements, administrative controls will be established to provide 
assurance that the AOT extension is not applied during adverse 
weather conditions that could potentially affect offsite power 
availability. Administrative controls are also implemented to avoid 
or minimize risk significant plant configurations during the time 
when an EDG is removed from service.
    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 involved a change in the design, 
configuration, or method of operation of the plant that could create 
the possibility of a new or different kind of accident. The proposed 
change extends the AOT currently allowed by the TS to 14 days. It 
also provides for a reduction to 72 hours, not to exceed 14 days, 
should the AACDG become inoperable during the extended AOT.
    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 ESF systems required to mitigate the consequences of 
postulated accidents consist of two independent trains. The ESF 
systems on either of the two trains provide for the minimum safety 
functions necessary to shut down the unit and maintain it in a safe 
shutdown condition. Each of the two trains can be powered from one 
of the offsite power sources of its associated EDG. In addition, the 
AACDG is available to provide power to either or both of the two 
trains. This design provides adequate defense in depth to ensure 
that diverse power sources are available to accomplish the required 
safety functions. Thus, with one EDG out of service, there are 
sufficient means to accomplish the safety functions and prevent the 
release of radioactive material in the event of an accident.
    The proposed change does not affect any of the assumptions or 
inputs to the Final Safety Analyses Report and does not erode the 
decrease in severe accident risk achieved with the issuance of the 
Station Blackout (SBO) Rule, 10 CFR 50.63, ``Loss of All Alternating 
Current Power.''
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 19, 2002.
    Description of amendment request: The proposed amendment would 
extend the allowed outage time (AOT) for a single inoperable low 
pressure safety injection (LPSI) train from 72 hours to 7 days. In 
addition, an AOT of 72 hours would be included for other conditions 
where the equivalent of a single emergency core cooling system (ECCS) 
subsystem flow is still available to both the LPSI and high pressure 
safety injection (HPSI) trains. Also, if 100% of ECCS flow is 
unavailable due to two inoperable HPSI or LPSI trains, an action 
statement would been added to restore at least one of each HPSI and 
LPSI train to operable status within one hour.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The HPSI and LPSI trains are part of the ECCS subsystem. 
Inoperable HPSI or LPSI components are not accident initiators in 
any accident previously evaluated. Therefore, this change does not 
involve an increase in the probability of an accident previously 
evaluated. Both the HPSI and LPSI systems are primarily designed to 
mitigate the consequences of a Loss of Coolant Accident (LOCA). 
These proposed changes do not affect any of the assumptions used in 
the deterministic LOCA analysis. Hence the consequences of accidents 
previously evaluated do not change.
    In order to fully evaluate the LPSI AOT extension, probabilistic 
safety analysis (PSA) methods were utilized. The results of the 
analyses show no significant increase in the core damage frequency. 
As a result, there would be no significant increase in the 
consequences of an accident previously evaluated. The analyses are 
detailed in CE NPSD-995, Combustion Engineering Owners Group Joint 
Applications Report for Low Pressure Safety Injection System AOT 
Extension.
    The proposed change allows a combination of equipment from 
redundant trains to be inoperable provided that at least the 
equivalent flow of a single HPSI and LPSI train of ECCS remains 
operable. Analyzed events are assumed to be initiated by the failure 
of plant structures, systems or components. Allowing equipment from 
redundant trains to constitute a single operable train does not 
increase the probability that a failure leading to an analyzed event 
will occur. The ECCS components are passive until an actuation 
signal is generated. This change does not increase the failure 
probability of the ECCS components. As such, the probability of 
occurrence for a previously analyzed accident is not significantly 
increased.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not change the design or configuration 
of the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which the plant is 
operated, and the setpoints at which protective or mitigative 
actions are initiated are unaffected by this change. No alteration 
in the procedures, which ensure the plant remains within analyzed 
limits, is being proposed and no change is being made to the 
procedures relied upon to respond to an off-normal event. As such, 
no new failure modes are being introduced. The proposed change will 
only provide the plant some flexibility in maintaining the minimum 
equipment required to be Operable to perform the ECCS function while 
in this Condition. The change does not alter assumptions made in the 
safety analysis and licensing basis.
    Therefore, the proposed change will 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 CE NPSD-995 and ANO-2 [Arkansas Nuclear One, Unit 2] PSA 
evaluations demonstrate that the changes are essentially risk 
neutral or risk beneficial. The margin of safety is established 
through equipment design, operating parameters, and the setpoints at 
which automatic actions are initiated. None of these are adversely 
impacted by the proposed change. Sufficient equipment remains 
available to actuate upon demand for the purpose of mitigating a 
transient event. The proposed change, which allows operation to 
continue for up to 72 hours with components inoperable in both ECCS 
subsystems, is acceptable based on the remaining ECCS components 
providing 100% of the required ECCS flow.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.


[[Page 68735]]


    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 12, 2001, as supplemented on 
October 10, 2002.
    Description of amendment request: The proposed amendment would 
change the Technical Specification Tables 3.2.A, 3.2.B, 4.2.A, and 
4.2.B. The proposed changes affect various instrument trip level 
settings and decreases the calibration frequencies for a variety of 
instruments. The proposed changes also involve clarifications to the 
Reactor Water Cleanup system trip configuration and the titles of 
certain trip systems. In addition, the proposed changes would make 
certain editorial and administrative corrections. The proposed setpoint 
changes and calibration frequencies are based on the licensee's 
evaluation.
    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. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The methodology used to determine the proposed trip level 
settings and surveillance intervals ensure adequate performance of 
the affected instrumentation. In addition, the affected instruments 
are not initiators of any accident previously evaluated. Therefore, 
the proposed trip level setting and surveillance intervals will not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed changes to trip level settings and surveillance 
intervals were establish using methodologies subject to 10 CFR 
Appendix B Quality Assurance program and ensure existing 
radiological limits are met. Therefore, the proposed trip level 
settings and surveillance intervals will not involve a significant 
increase in the consequences of an accident previously evaluated.
    Other changes are editorial or administrative in nature and can 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    No new or different [kind] of accidents or malfunctions than 
those previously analyzed in Pilgrim's UFSAR [Updated Final Safety 
Analysis Report] are introduced by this proposed change because 
there are no new failure modes introduced. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Will not involve a significant reduction in the margin of 
safety.
    The proposed changes to trip level settings and surveillance 
intervals were established using approved methodologies subject to a 
10 CFR, Appendix B, Quality Assurance program and existing 
radiological limits are met. These changes do not impact Pilgrim's 
configuration or operation.
    Editorial and administrative type changes do not impact the 
operation or configuration of Pilgrim. For the above reasons the 
proposed change does 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: James W. Andersen, Acting.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: July 5, 2002.
    Description of amendment request: The proposed amendment would 
relocate the ``Primary System Boundary--Shock Suppressors (Snubbers),'' 
Technical Specifications (TS) 3/4.6.I, 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. Snubbers are not assumed to be an 
initiator of any analyzed event, nor are they 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 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 any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change 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. As such, there 
are no changes being made to safety analysis assumptions, safety 
limits or safety system settings that would adversely affect plant 
safety as a result of the proposed change. Margins of safety are 
unaffected by requirements that are retained, but relocated from the 
Technical Specifications 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Andersen, Acting.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: August 16, 2002.
    Description of amendment request: The proposed amendment would 
relocate certain Control Rod Block functions from Technical 
Specifications 3/4.2.C, ``Instrumentation that Initiates Rod Blocks,'' 
to the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 68736]]

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. These control rod blocks are not assumed to be an 
initiator of any analyzed event, nor are they 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 any physical alteration of 
plant equipment and does not change the method by which any safety-
related system performs its function. As such, no new or different 
types of equipment will be installed, and the basic operation of 
installed equipment is unchanged. The methods governing plant 
operation and testing remain consistent with current safety analysis 
assumptions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident form any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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. As such, there are no changes being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by requirements 
that are retained, but relocated from the Technical Specifications 
to the FSAR [Final Safety Analysis Report]. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Andersen, Acting.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: October 4, 2002.
    Description of amendment request: Change the Technical 
Specifications by extending the primary containment integrated leak 
rate testing (ILRT) interval on a one-time basis from 10 years to no 
longer than approximately 10.6 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. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specifications 6.7.C 
``Primary Containment Leak Rate Testing Program'' involves a one-
time extension to the current interval for Type A containment 
testing. The current test interval of 10 years would be extended on 
a one-time basis to no longer than approximately 10.6 years from the 
last Type A test. The proposed Technical Specification change does 
not involve a physical change to the plant or a change in the manner 
in which the plant is operated or controlled. The reactor 
containment is designed to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such, the reactor containment itself 
and the testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident, and do not 
involve the prevention or identification of any precursors of 
anaccident. Therefore, the proposed Technical Specification change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests. Type B and C containment 
leak rate tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG[-]1493, that Type B and 
C containment leakage tests have identified a very large percentage 
of containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. VY's [Vermont Yankee] ILRT test history supports this 
conclusion. NUREG-1493 concluded, in part, that reducing the 
frequency of Type A containment leak tests to once per twenty (20) 
years leads to an imperceptible increase in risk. The integrity of 
the reactor containment is subject to two types of failure 
mechanisms which can be categorized as (1) activity based and (2) 
time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as design change control and procedural requirements 
for system restoration ensure that containment integrity is not 
degraded by plant modifications or maintenance activities. The 
design and construction requirements of the reactor containment 
itself combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI, the Maintenance Rule and Licensing commitments related 
to containment coatings serve to provide a high degree of assurance 
that the containment will not degrade in a manner that is detectable 
only by Type A testing. Therefore, the proposed Technical 
Specification change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed revision to the Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specification change does not involve a physical change to 
the plant or the manner in which the plant is operated or 
controlled. Therefore, the proposed Technical Specification change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed revision to Technical Specifications involves a 
one-time extension to the current interval for Type A containment 
testing. The proposed Technical Specification change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific requirements 
and conditions of the Primary Containment Leak Rate Testing Program, 
as defined in Technical Specifications, exist to ensure that the 
degree of reactor containment structural integrity and leak-
tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by 
Technical Specifications is maintained. The proposed change involves 
only the extension of the interval between Type A containment leak 
rate tests. The proposed surveillance interval extension is

[[Page 68737]]

bounded by the 15 month extension currently authorized within NEI 
[Nuclear Energy Institute] 94-01. Type B and C containment leak rate 
tests will continue to be performed at the frequency currently 
required by plant Technical Specifications. VY's, as well as the 
industries experience, strongly supports the conclusion that Type B 
and C testing detects a large percentage of containment leakage 
paths and that the percentage of containment leakage paths that are 
detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI, the 
Maintenance Rule and the Coatings Program serve to provide a high 
degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. Additionally, the 
on-line containment monitoring capability that is inherent to 
inerted BWR [Boiling Water Reactor] containments allows for the 
detection of gross containment leakage that may develop during power 
operation. The combination of these factors ensures that the margin 
of safety that is inherent in plant safety analysis is maintained. 
Therefore, the proposed Technical Specification change does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: August 16, 2002.
    Description of amendment request: The proposed amendments would 
modify the Unit 3 allowable value, and the Units 2 and 3 surveillance 
requirements for the reactor protection system scram discharge volume 
water level-high function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Dresden Nuclear Power Station (DNPS), Unit 3 plans to implement 
a design change that upgrades the Scram Discharge Volume Water 
Level--High instrumentation from existing float-type level switches 
to electronic analog trip units. Analog trip units are a proven 
technology that is more reliable than existing equipment. The 
proposed design is consistent with a generic design that has been 
previously reviewed and approved by the NRC. Analog trip units are 
used in various applications at DNPS, including the Reactor 
Protection System (RPS) Low Water Level Trip Function.
    The proposed Technical Specifications (TS) changes add new Unit 
3 Channel Check and trip unit calibration Surveillance Requirements 
(SRs) for the new analog trip units associated with the Scram 
Discharge Volume Water Level--High RPS Trip Function. These new Unit 
3 SRs are not applicable to the existing instrumentation because the 
existing float-type level switches are non-indicating and do not 
employ trip units. In addition, the proposed TS changes add a new 
trip unit calibration SR for existing Unit 2 and 3 instrumentation 
that is composed of differential pressure type level transmitter 
switches.
    TS requirements that govern operability or routine testing of 
plant instruments are not assumed to be initiators of any analyzed 
event because these instruments are intended to prevent, detect, or 
mitigate accidents. Therefore, these proposed changes will not 
involve an increase in the probability of an accident previously 
evaluated. Additionally, these proposed changes will not increase 
the consequences of an accident previously evaluated because the 
proposed changes do not adversely impact structures, systems, or 
components. The planned Unit 3 instrument upgrade is a more reliable 
design than existing equipment. The proposed changes establish 
requirements that ensure components are operable when necessary for 
the prevention or mitigation of accidents or transients. 
Furthermore, there will be no change in the types or significant 
increase in the amounts of any effluents released offsite.
    In summary, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes support a planned instrument upgrade on 
Unit 3 by incorporating SRs required to ensure operability. There is 
no change being made to the parameters within which DNPS is 
operated. The proposed changes do not adversely impact the manner in 
which the Scram Discharge Volume Water Level--High RPS 
instrumentation will operate under normal and abnormal operating 
conditions. The proposed changes will not alter the function demands 
on credited equipment. No alteration in the procedures, which ensure 
DNPS remains within analyzed limits, is proposed, and no change is 
being made to procedures relied upon to respond to an off-normal 
event. Therefore, these proposed changes provide an equivalent level 
of safety and will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The changes 
in methods governing normal plant operation are consistent with the 
current safety analysis assumptions. Therefore, these proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
and actions. The proposed changes support a planned instrumentation 
upgrade to enhance the reliability of RPS instrumentation. The 
proposed changes do not affect the probability of failure or 
availability of the affected instrumentation. The revised Allowable 
Value, addition of a Channel Check and trip unit calibration, and 
revision of other SRs for RPS Instrumentation Channel Check and trip 
unit calibration, and revision of other SRs for RPS Instrumentation 
Function 7 (Scram Discharge Volume Water Level--High) are 
conservative changes that align the SRs for proper determination of 
operability with that of similar instrumentation. Therefore, it is 
concluded that the proposed changes do not result in a reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 15, 2002.
    Description of amendment request: The proposed amendment modifies 
the reactor coolant system flow rate from 363,000 gallons per minute 
(gpm) to 355,000 gpm in Saint Lucie Unit 2 Technical Specifications 
(TS) Table 3.3-2 and a footnote for Table 2.2-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment would decrease the value of design 
reactor coolant system flow rate. This reduction in the reactor

[[Page 68738]]

coolant system (RCS) flow requirement will support operation of the 
plant with an increased steam generator (SG) tube plugging. The 
changes to the Technical Specification (TS) bases either support the 
proposed flow reduction or are administrative in nature, consistent 
with the current design basis. The parameters affected by the 
proposed changes are not accident initiators and do not affect the 
frequency of occurrence of previously analyzed transients. 
Additionally, there are no changes to any active plant component.
    This evaluation has demonstrated acceptable results for all the 
accidents previously analyzed. It is concluded that the radiological 
consequences would remain within their established acceptance 
criteria when including effects of the proposed reduction in the RCS 
flow, which would support an increased steam generator tube plugging 
level.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    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 previously evaluated.
    This proposed amendment revises the RCS design flow requirement 
to cover plant operation with increased steam generator tube 
plugging. There are no physical changes to the plant systems or 
system interactions due to the proposed changes. The modes of 
operation of the plant and the design functions of all the safety 
systems remain unchanged.
    Therefore, 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.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The impact of the proposed changes on the design basis accident 
analysis was evaluated and it is concluded that the setpoint and 
safety analyses of all design basis accidents meet the applicable 
acceptance criteria with respect to the radiological consequences, 
specified acceptable fuel design limits (SAFDL), primary and 
secondary overpressurization, peak containment pressure and 
temperature, and 10 CFR 50.46 requirements.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Allen G. Howe.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: October 21, 2002.
    Description of amendment request: The proposed amendment deletes 
the requirements defined in Technical Specification (TS) 3/4.9.3, 
``Refueling Operations, Decay Time,'' and places them in the TS Bases. 
Additionally this amendment proposes to modify the TS Bases definition 
of ``recently irradiated fuel'' will be re-defined as fuel that has 
occupied part of a critical reactor core within the previous 72 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The accident of concern related to the proposed change is 
the fuel handling accident (FHA). This accident assumes a dropped 
fuel assembly. One of the assumptions made in the analysis is that 
fuel movement is delayed for some time period after shutdown to 
accommodate cooldown of the reactor coolant system and disassembly 
of the reactor pressure vessel. This delay period allows for 
radioactive decay of the in-reactor vessel fission product 
inventory. Reducing the analyzed decay time from 100 hours to 72 
hours does not increase the probability of a FHA because the timing 
of fuel movement in the reactor pressure vessel does not alter the 
manner in which fuel assemblies are handled.
    Reducing the analyzed decay time from 100 hours to 72 hours does 
increase the offsite dose and control room dose projections of a FHA 
above those previously reviewed and approved by the NRC for Turkey 
Point Units 3 and 4 per Amendments 216 and 210. However, it has been 
shown by reanalysis of such an accident involving irradiated fuel 
with at least 72 hours of decay that the projected doses remain well 
within applicable regulatory limits. Hence, the proposed change in 
timing of fuel movement in the reactor pressure vessel does not 
involve a significant increase in the consequences of a FHA.
    Additionally, the manner in which the minimum in-reactor vessel 
decay time is controlled will not impact the probability of 
occurrence, or the consequences of a FHA. Relocating the decay time 
requirement from the TS to the TS Bases document and other 
administrative controls will continue to ensure that this key 
accident analysis assumption is upheld. The inherent delay 
associated with completing the required preparatory steps for moving 
fuel in the reactor vessel further ensures that the proposed 72-hour 
decay time will be met for a refueling outage.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    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?
    No. The impact of the proposed change is limited to fuel 
handling operations and spent fuel pool cooling. No physical plant 
changes are proposed to accommodate the timing change for fuel 
movement. Hence, no new failure modes are created that would cause a 
new or different kind of accident from any accident previously 
evaluated. The supporting analysis for the timing change 
demonstrates that the associated increase in decay heat load will 
not cause any spent fuel pool (SFP) component or structure to 
operate outside design limits. Adequate margins to safety are 
maintained with respect to SFP water temperature and structural 
loading.
    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.
    Additionally, the manner which the minimum in-reactor vessel 
decay time is controlled will not impact the operation of any 
structure, system, or component.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    No. The proposed change in plant operation does not 
significantly reduce the margin of safety. It has been shown by 
reanalysis of a FHA involving irradiated fuel with at least 72 hours 
of decay that the projected doses will be well within applicable 
regulatory limits. Additionally, it has been shown by thermal 
hydraulic analysis that operation of the SFP cooling system in 
accordance with the restrictions and limitations identified in the 
amendments application will maintain adequate margins to pool 
boiling. Analysis of transient SFP concrete temperatures similarly 
demonstrates that the integrity of the pool structure will not be 
compromised if the amount of in-reactor vessel fuel assembly decay 
time is reduced from 100 hours to 72 hours.
    The proposed change in the manner in which the minimum in-
reactor vessel decay time will be controlled will not impact plant 
safety. Relocating the decay time requirement from the TS to the TS 
Bases document and other administrative controls will continue to 
ensure that this key accident analysis

[[Page 68739]]

assumption is upheld. The inherent delay associated with completing 
the required preparatory steps for moving fuel in the reactor vessel 
further ensures that the proposed 72-hour decay time will be met for 
a refueling outage.
    Therefore, operation of the facility in accordance with the 
proposed amendments 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Allen G. Howe.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: October 16, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 3.3-4, ``Engineered Safety Feature 
Actuation System Instrumentation Trip Setpoints.'' The proposed changes 
are part of a planned design change to replace the existing 4160 volt 
(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.
    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 changes to the degraded voltage and loss-of-voltage 
setpoints and time delay affect when an emergency bus that is 
experiencing low or degraded voltage will trip from offsite power 
and shift to an emergency diesel generator. While the setpoints that 
initiate this action will be modified, the function remains the 
same. The setpoints have been analyzed to ensure spurious trips will 
be avoided. The proposed changes will not significantly affect any 
accident initiators or precursors. The format changes are intended 
to improve readability, consistency with NUREG-1431, Revision 2, and 
appearance. In addition, they do not alter any requirements. The 
bases change provides explanatory information only. Thus, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.

Consequences of an Accident Previously Evaluated

    The proposed changes to the degraded voltage and loss-of-voltage 
setpoints and time delay affect when an emergency bus that is 
experiencing low or degraded voltage will trip from offsite power 
and shift to an emergency diesel generator. While the setpoints that 
initiate this action will be modified, they are bounded by the 
current safety analysis. The function of the plant equipment remains 
the same. The proposed changes improve the reliability of safety-
related equipment to operate as designed. The format changes are 
intended to improve readability, consistency with NUREG-1431, 
Revision 2, and appearance. In addition, they do not alter any 
requirements. The bases change provides explanatory information 
only. Thus, the consequences of an accident previously analyzed 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 changes to the degraded voltage and loss-of-voltage 
setpoints and time delay do not affect existing or introduce any new 
accident precursors or modes of operation. The relays will continue 
to detect undervoltage conditions and transfer safety loads to the 
emergency diesel generators at a voltage level adequate to ensure 
proper safety equipment performance and to prevent equipment damage. 
The function of the relays remains the same. The format changes are 
intended to improve readability, consistency with NUREG-1431, 
Revision 2, and appearance. In addition, they do not alter any 
requirements. The bases change provides explanatory information 
only. Thus, 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?
    Response: No.
    The proposed changes will allow all safety-related loads to have 
sufficient voltage to perform their intended safety function while 
ensuring spurious trips are avoided. Thus, the results of the 
accident analyses will not be affected as the input assumptions are 
protected. The format changes are intended to improve readability, 
consistency with NUREG-1431, Revision 2, and appearance. In 
addition, they do not alter any requirements. The bases change 
provides explanatory information only. Thus, the proposed changes do 
not involve a significant reduction in a margin of safety.
    In summary, based upon the above evaluation, [Indiana Michigan 
Power Company] I&M has concluded that the proposed changes involve 
no significant hazards consideration under the standards set forth 
in 10 CFR 50.92(c), and accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    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.

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

    Date of amendment request: October 7, 2002.
    Description of amendment request: The proposed amendment would add 
Specification 4.0.3 to address missed surveillances. This new 
specification specifies an initial 24-hour delay period for performing 
a missed surveillance prescribed by Specification 3.0.3. Specification 
4.0.3 will also require: ``A risk evaluation shall be performed for any 
surveillance delayed greater than 24 hours and the risk impact shall be 
managed.'' In addition, the licensee proposed to add wording to each of 
the following existing specifications such that the new Specification 
4.0.3 would apply to them: Specification 6.16, 6.17, 6.18, and 6.19.
    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 October 7, 2002. The NSHC 
determination is restated below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:


[[Page 68740]]



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

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

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

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

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 8, 2002.
    Description of amendment request: The proposed amendment will 
change the Limiting Condition for Operation (LCO) 2.3(2).a, ``Emergency 
Core Cooling Systems,'' for the allowed outage time (AOT) for a single 
train of the low pressure safety injection system. The proposed change 
is based on the Combustion Engineering Owners Group Topical Report CE 
NPSD-995, ``Joint Applications Report for Low Pressure Safety Injection 
System AOT Extension.'' This amendment will permit the licensee to 
extend the AOT for a single low pressure safety injection (LPSI) train 
from the existing 24 hours to 7 days.
    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 allowed outage time is not an initiator of any previously 
evaluated accident. The proposed change to the allowed outage time 
for a single LPSI train will not prevent the safety systems from 
performing their accident mitigation function as assumed in the 
safety analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any 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 only affects the technical specifications 
and does not involve a physical change to the plant. Modifications 
will not be made to existing components nor will any new or 
different types of equipment be installed. The proposed change 
modifies the allowed outage time for a single LPSI train from 24 
hours to 7 days for the purpose of performing preventive or 
corrective maintenance, or surveillance testing. Actions will be 
taken to ensure the increase in LPSI allowed outage time is 
incorporated appropriately into plant procedures.
    Therefore, this 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 modifies the allowed outage time for a 
single LPSI train to permit necessary ECCS [emergency core cooling 
system] maintenance or testing to be performed in a measured, 
deliberate fashion. Results of an integrated assessment of the 
overall plant risk associated with the adoption of the proposed AOT 
extension show a negligible increase in plant risk. The increase in 
allowed outage time will also permit more efficient and more safely 
managed plant operations and can help reduce the risk associated 
with changing plant operating modes.
    An evaluation of the impact of extending the AOT for a single 
LPSI train on plant risk was performed for the conditions of the 
plant being at power. While at power, the incremental conditional 
core damage frequency (ICCDF) was determined to be 1.396E-05 per 
year, with a 5.80E-07 per year incremental increase in the core 
damage frequency attributed to extending the allowed outage time 
from 24 hours to seven days.
    A sensitivity analysis was performed to identify the impact on 
core damage probability over a seven day interval that results from 
performing maintenance on one LPSI train while in a shutdown mode. 
Results of this study show that even small improvements in LPSI 
train reliability will produce a decrease in core damage 
probability, thus the net impact of performing LPSI train preventive 
maintenance while at power is risk-beneficial.
    The unavailability of one LPSI train resulted in a large early 
release frequency of 2.636E-06 per year, with a 2.40E-08 per year 
incremental conditional large early release frequency (ICLERF) 
attributed to extending the allowed outage time from 24 hours to 
seven days.
    Therefore, this technical specification change does not involve 
a significant reduction in the margin of safety.

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

[[Page 68741]]

    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: October 8, 2002.
    Description of amendment request: The proposed amendment would 
relocate the requirements of Technical Specification (TS) 2.13, 
``Nuclear Detector Cooling System,'' to the Fort Calhoun Station 
Updated Safety Analysis Report (USAR). The accident analyses do not 
assume operation of the nuclear detector cooling system; therefore, 
this system does not meet the criteria set forth in 10 CFR 
50.36(c)(2)(ii) for inclusion in the TS. The requirements will be 
relocated to the USAR.
    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 relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in the TS set 
forth in 10 CFR 50.36(c)(2)(ii). The requirements for Nuclear 
Detector Cooling are being relocated from TS to the USAR, which will 
be maintained pursuant to 10 CFR 50.59, thereby reducing the level 
of regulatory control. The level of regulatory control has no impact 
on the probability or consequences of an accident previously 
evaluated. Therefore, the 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 relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in TS set forth 
in 10 CFR 50.36(c)(2)(ii). The change does not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or make changes in the methods governing normal plant 
operation. The change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing 
basis. Therefore, the change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change relocates requirements for Nuclear Detector 
Cooling that do not meet the criteria for inclusion in TS set forth 
in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin of 
safety since the location of a requirement has no impact on any 
safety analysis assumptions. In addition, the relocated requirements 
for Nuclear Detector Cooling remain the same as the existing TS. 
Since any future changes to these requirements or the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
there will be no reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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: October 8, 2002.
    Description of amendment request: The proposed amendment would 
increase the amount of diesel fuel oil required by Technical 
Specification (TS) 2.7, ``Electrical Systems,'' to be kept in the 
auxiliary boiler fuel oil storage tank. A recent calculation determined 
that the amount of diesel fuel oil required by TS 2.7 is slightly 
insufficient (35 gallon shortfall) for 7 days of emergency diesel 
generator (EDG) 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    No changes to the EDG diesel fuel oil storage and distribution 
system configuration or usage is required to achieve the inventory 
increase. This change only increases the current minimum inventory 
requirements listed in TS 2.7 and assures that the inventory will 
meet the capacity requirements of IEEE-308, which requires 
sufficient fuel for 7 days of EDG operation following the most 
severe accident. Increasing the minimum inventory requirement of FO-
10, the auxiliary boiler fuel oil tank by 2000 gallons enables the 
site to meet this criterion and provides an extra margin of 
inventory to prevent any future concerns.
    Therefore, this change does not involve an increase in the 
probability or consequences of any 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.
    No changes to the Emergency Diesel Generator fuel oil storage 
and distribution system configuration or usage are required to 
achieve the inventory increase. FO-10 has a capacity of 18,000 
gallons. Therefore, FO-10 can readily accommodate the additional 
2000 gallons of inventory. Therefore, the proposed changes will 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 will increase the margin of safety by 
requiring that additional diesel fuel oil inventory be kept on-site 
to ensure that the 7 day on-site fuel supply criteria is met.
    Therefore, this technical specification change does not involve 
a reduction in the margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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: October 8, 2002.
    Description of amendment request: The proposed amendment will 
relocate the requirements of Technical Specification (TS) 3.5(5), 
``Surveillance for Prestressing System,'' for testing prestressed 
concrete containment tendons to the Fort Calhoun Station (FCS) Updated 
Safety Analysis Report (USAR). This proposed amendment will also add a 
TS requirement (TS 5.21) for a containment tendon testing program 
consistent with that presented in Section 5.5 of NUREG-1432, ``Improved 
Standard Technical Specification (ITS) for Combustion Engineering 
Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or

[[Page 68742]]

consequences of an accident previously evaluated.
    The proposed change relocates requirements for testing 
Prestressed Concrete Containment Tendons that do not meet the 
criteria for inclusion in the TS set forth in 10 CFR 
50.36(c)(2)(ii). The requirements for testing Prestressed Concrete 
Containment Tendons are being relocated from TS to the USAR, which 
will be maintained pursuant to 10 CFR 50.59, thereby reducing the 
level of regulatory control. The level of regulatory control has no 
impact on the probability or consequences of an accident previously 
evaluated. Therefore, the 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 relocates requirements for testing 
Prestressed Concrete Containment Tendons that do not meet the 
criteria for inclusion in TS set forth in 10 CFR 50.36(c)(2)(ii). 
The change does not involve a physical alteration of the plant (no 
new or different type of equipment will be installed) or make 
changes in the methods governing normal plant operation. The change 
will not impose different requirements, and adequate control of 
information will be maintained. This change will not alter 
assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change relocates requirements for testing 
Prestressed Concrete Containment Tendons that do not meet the 
criteria for inclusion in TS set forth in 10 CFR 50.36(c)(2)(ii). 
The change will not reduce a margin of safety since the location of 
a requirement has no impact on any safety analysis assumptions. In 
addition, the relocated requirements for testing Prestressed 
Concrete Containment Tendons remain the same as the existing TS. 
Since any future changes to these requirements or the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
there will be no reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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: October 8, 2002.
    Description of amendment request: The proposed amendment will 
change Technical Specification 5.19, ``Containment Leakage Rate Testing 
Program,'' to extend the integrated leak rate test (ILRT) surveillance 
interval from 10 to 15 years. The proposed changes are justified based 
on a combination of risk-informed analysis and assessment of the 
containment structural condition utilizing ILRT historical results and 
containment inspection programs.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change adds a one-time extension to the current 
surveillance interval for Type A testing (ILRT). The current test 
interval of 10 years, based on performance history, would be 
extended on a one-time basis to 15 years from the last Type A test. 
The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since the 
containment Type A test is not a modification, nor a change in the 
way that plant systems, structures, or components are operated, and 
is not an activity that could lead to equipment failure or accident 
initiation. The proposed change does not involve a significant 
increase in the consequences of an accident since research in 
Reference 10.3 [NUREG-1493, Performance Based Containment Leak-Test 
Program] has found that generically very few potential leaks are not 
identified in Type B and C tests. Reference 10.3 concluded that an 
increase in the test interval to 20 years resulted in an 
imperceptible increase in risk. FCS provides a high degree of 
assurance through testing and inspection that the containment will 
not degrade in a manner only detectable by Type A testing. 
Inspections required by ASME code and the Maintenance Rule are 
performed in order to identify indications of containment 
degradation that could affect leak tightness. Type B and C testing 
required by 10 CFR 50, Appendix J are not affected by this proposed 
extension to the Type A test interval and will continue to identify 
containment penetration leakage paths that would otherwise require a 
Type A test.
    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 adds a one-time extension to the current 
surveillance interval [* * *] for Type A testing (ILRT). The change 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or make changes in 
the methods governing normal plant operation. This change will not 
alter assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will not result in operation of the facility 
involving a significant reduction in a margin of safety. The 
proposed change adds a one-time extension to the current interval 
for Type A testing. The current test interval of 10 years, based on 
performance history, would be extended on a one-time basis to 15 
years from the last Type A test. Reference 10.3 has found that 
generically very few potential leaks are not identified in Type B 
and C tests. Reference 10.3 concluded that an increase in the test 
interval to 20 years resulted in an imperceptible increase in risk. 
Furthermore, the extended test interval would have a minimal effect 
on such risk since Type B and C testing detect over 95 percent of 
potential leakage paths. A plant specific risk calculation, as part 
of Reference 10.2, [WCAP-15691, Joint Applications Report for 
Containment Integrated Leak Rate Test Interval Extension, Revision 
3, August 2002] on this topic obtained results consistent with the 
generic conclusions of Reference 10.3. The overall increase in risk 
contribution was determined as 0.31%.

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

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

    Date of amendment request: September 24, 2002.
    Description of amendment request: This proposed license amendment 
request would revise Technical Specification (TS) 4.8.1.1, ``AC 
Sources'' and the associated Bases section related to the Emergency 
Diesel Generators (EDG). This change would clarify the requirement for 
the start time test performed on a 184 day and an 18-month frequency. 
The proposed change will revise Surveillance Requirement (SR) 
4.8.1.1.2.f.1 and 4.8.1.1.2.g.5 to more accurately reflect the plant 
conditions during EDG start testing.

[[Page 68743]]

    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 change does not involve a significant increase in 
probability or consequences of an accident previously evaluated?
    This proposed amendment modifies an EDG Surveillance Requirement 
and does not impact the offsite AC distribution system; therefore 
the probability of any LOOP [loss of off-site power], including one 
concurrent with a LOCA [loss-of-coolant accident] is not 
significantly increased.
    The proposed change revises the SR to better match the plant 
conditions during the test. SR 4.8.1.1.2.f.1 and 4.8.1.1.2.g.5 are 
performed with the EDG unloaded and as a result, overshoots its 
target nominal voltage and frequency during the test. In an actual 
event, the EDG would be almost immediately loaded once minimum 
voltage and frequency requirements are satisfied, thereby minimizing 
the overshoot.
    To ensure the EDGs are capable of fulfilling their safety 
function, the proposed SR requires EDG voltage and frequency to 
achieve the specified minimum acceptable valued within 10 seconds, 
and to settle to a steady state voltage and frequency within the 
minimum and maximum values. That is, the upper limits are only 
applicable for steady state operation and do not apply during the 
transient portion of the EDG start. This change revises the 
acceptance criteria of 4.8.1.1.2.f.1 and 4.8.1.1.2.g.5 to clarify 
which voltage and frequency limits are applicable during the 
transient and steady state portions of the EDG start test.
    This change does not affect the EDGs ability to supply the 
minimum voltage and frequency within 10 seconds or the steady state 
voltage and frequency required by the FSAR [Final Safety Analysis 
Report]. The EDGs will continue to perform their intended safety 
function, in accordance with the safety analysis. Thus, the 
consequences of any previously analyzed event are not significantly 
increased by this change.
    The proposed change to 3.8.1.1, Action b.2 will not increase the 
probability or consequences of an accident previously evaluated. The 
change to this requirement to allow determination of no common cause 
failure mechanism has no impact on any accident. This change allows 
for not testing the redundant EDG if it can be demonstrated the 
failure mechanism of the affected EDG is not common cause. The 
normal TS surveillance testing schedule assures that operable EDG(s) 
are capable of performing their intended safety functions. The 
revision to the footnote on page 3/4.8-1 assures the action will be 
completed even if the EDG is restored to operable status within the 
action completion time.
    The proposed revision to the fuel oil surveillance program will 
not preclude the EDGs from fulfilling their design functions. These 
changes provide flexibility to the testing program and continue to 
provide assurances that the fuel oil is acceptable for sustained 
engine operation. Eliminating or revising methodologies for testing 
of the fuel oil will not increase any probabilities or consequences 
to any accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change revises SR 4.8.1.1.2.f.1 requirements to clarify 
which voltage and frequency limits are applicable during the 
transient and steady state portions of the EDG start testing. No 
changes are being made in equipment hardware or software, 
operational philosophy, testing frequency, how the system actually 
operates, or how the system is physically tested. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The elimination of unnecessary surveillance testing does not 
affect the design bases of the EDGs. The EDGs are designed to 
provide electrical power to the equipment important for safety 
during all modes and plant conditions following a loss of offsite 
power. The proposed changes to the Action requirements are 
consistent with NUREG-1431, NUREG-1366, Generic Letter 93-05, 
industry operating experience, and VCS operating experience. These 
changes are intended to improve plant safety, decrease equipment 
degradation, and remove unnecessary burden on personnel resources by 
reducing the amount of testing that the TS requires during power 
operation.
    The revision to the fuel oil testing methodology does not impact 
the capabilities or functions of the EDGs. This testing methodology 
change will continue to assure the EDG is not degraded due to the 
fuel oil used. Existing test methodologies and guidance will 
continue to be followed, unless an evaluation demonstrates another 
methodology is as effective. Since the changes do not adversely 
impact important to safety equipment that is used in mitigating an 
accident, they will not create the possibility of an accident 
different from any previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The EDGs will still perform their intended safety function, in 
accordance with the VCSNS accident analysis. The revised test 
acceptance criteria are a much better match for the tested condition 
(unloaded). The performance of other TS SRs (in particular 4.8.1.1.2 
g.4.b, 4.8.1.1.2g.6 and 4.8.1.1.2g.14) demonstrate EDG operability 
in conditions that are more representative of postulated accident 
conditions (loaded in the actual time sequence assumed in the 
accident analysis). The proposed amendment does not alter any 
acceptance criteria or equipment testing scope, which could impact 
the accident analysis.
    The proposed change to exempt specific surveillance testing, as 
long as potential common cause can be ruled out, and eliminate 
unnecessary mechanical stress and wear on the diesel generator is an 
effort to improve plant reliability and safety. These changes are 
consistent with NUREG-1431, NUREG-1366, industry operating 
experience, and VCS operating experience and do not adversely affect 
the design bases, accident analysis, reliability or capability of 
the EDGs to perform their intended safety function. The revised 
footnote will assure that once the action is initiated, it will be 
completed regardless of when the EDG is restored to operability.
    The proposed change to the fuel oil testing methodology has no 
impact on any safety margin. Accident analysis requires that the 
EDGs provide electric power to selected components during an 
accident scenario. The fuel oil quality will continue to meet 
established acceptance criteria and support the design function of 
the EDGs.
    Since the design and licensing basis of the plant is unaffected, 
the proposed amendment will not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: April 4, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications 5.5.17, ``Containment Leakage rate 
Testing Program,'' to reflect a one-time deferral of the Type-A 
Containment Integrated Leak Rate Test (ILRT). The 10-year interval 
between ILRTs is to be extended to 15 years from the previous ILRTs 
that were completed in March 1994 for Unit 1 and March 1995 for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed revision to Technical Specifications 5.5.17, 
``Containment Leakage Rate Testing Program,'' involves a one time 
extension to the current interval for Type A containment leak 
testing. The current test interval of ten (10) years would be 
extended

[[Page 68744]]

on a one time basis to no longer than fifteen (15) years from the 
last Type A test. The proposed Technical Specifications change does 
not involve a physical change to the plant or a change in the manner 
in which the plant is operated or controlled. The reactor 
containment is designed to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such, the reactor containment itself 
and the testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident, and do not 
involve the prevention or identification of any precursors of an 
accident. Therefore, the proposed Technical Specification change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed change involves only the extension of the interval 
between Type A containment leakage tests. Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by plant Technical Specifications. Industry 
experience has shown, as documented in NUREG-1493, that Type B and C 
containment leakage tests have identified a very large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is very 
small. FNP [Joseph M. Farley Nuclear Plant] test history supports 
this conclusion. NUREG-1493 concluded, in part, that reducing the 
frequency of Type A containment leak tests to once per twenty (20) 
years leads to an imperceptible increase in risk. The integrity of 
the reactor containment is subject to two types of failure mechanism 
which can be categorized as (1) activity based and (2) time based. 
Activity based failure mechanisms are defined as degradation due to 
system and/or component modifications or maintenance. Local leak 
rate test requirements and administrative controls such as design 
change control and procedural requirements for system restoration 
ensure that containment integrity is not degraded by plant 
modifications or maintenance activities. The design and construction 
requirements of the reactor containment itself combined with the 
containment inspections performed in accordance with ASME [American 
Society of Mechanical Engineers] Section XI, the Maintenance Rule 
and the containment coatings program serve to provide a high degree 
of assurance that the containment will not degrade in a manner that 
is detectable only by Type A testing. Therefore, the proposed 
Technical Specifications change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed Technical Specifications change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed revision to Technical Specifications involves a one 
time extension to the current interval for Type A containment leak 
testing. The reactor containment and the testing requirements 
invoked to periodically demonstrate the integrity of the reactor 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident and do not involve the prevention or 
identification of any precursors of an accident. The proposed 
Technical Specifications change does not involve a physical change 
to the plant or the manner in which the plant is operated or 
controlled. Therefore, the proposed Technical Specifications change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed Technical Specifications change does not involve 
a significant reduction in a margin of safety.
    The proposed revision to Technical Specifications involves a one 
time extension to the current interval for Type A containment leak 
testing. The proposed Technical Specifications change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The specific requirements 
and conditions of the Containment Leakage Rate Testing Program, as 
defined in Technical Specifications, exist to ensure that the degree 
of reactor containment structural integrity and leak tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leakage rates limits specified by Technical 
Specifications is maintained. The proposed change involves only the 
extension of the interval between Type A containment leakage tests. 
Type B and C containment leakage tests will continue to be performed 
at the frequency currently required by plant Technical 
Specifications.
    FNP and industry experience strongly support the conclusion that 
Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI, the 
Maintenance Rule and the Coatings Program serve to provide a high 
degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. Therefore, the 
proposed Technical Specifications 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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

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 amendment request: September 24, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) Limiting Conditions for Operation 
3.7.10, Control Room Emergency Filtration/Pressurization System; and 
associated Bases. These changes will allow maintenance on ventilation 
area pressure boundaries (i.e., doors) that cannot be conducted within 
the requirements of existing TS. The changes are based on U.S. Nuclear 
Regulatory Commission (NRC) approved Technical Specification Task 
Force--287, Rev. 5. In addition, the proposed amendments would revise 
TS 3.7.12 to eliminate a requirement to cease power operation if the 
fuel handling accident function of the penetration room filtration 
system is inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The control room emergency filtration/pressurization system 
(CREFS) and the penetration room filtration (PRF) system are not 
initiators of any accident. The proposed changes do not alter the 
physical plant nor do they alter modes of plant operation. 
Therefore, the proposed changes do not affect the probability of any 
accident previously evaluated. Compensatory actions such as the 
availability of self-contained breathing apparatus or iodine filters 
provide additional assurance that the requirements of GDC [General 
Design Criteria] 19 are met. Prohibiting movement of irradiated 
fuel, or loads over irradiated fuel or core alterations when the 
control room boundary is inoperable and limiting movement of 
irradiated fuel or loads over the fuel in the spent fuel pool room 
when its boundary is inoperable will eliminate the potential for 
exceeding GDC 19 due to a fuel handling accident. These actions will 
also prevent an off site dose release in excess of analyzed values. 
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 previously evaluated?
    The CREFS and the PRF systems are not initiators of any analyzed 
accident. The proposed changes do not alter the operation of the 
plant or any of its equipment, introduce any permanent new 
equipment, adversely impact maintenance practices or result in any 
new failure mechanisms or single failures. Any temporary equipment 
utilized for compensatory measures will be subject to existing 
administrative controls that address issues such as fire prevention 
and seismic concerns. Therefore, there is no

[[Page 68745]]

potential for a new accident and no potential for changing the 
progression of an analyzed accident. The proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Do the proposed changes result in a significant reduction in 
a margin of safety?
    The proposed changes do not adversely affect the ability of the 
fission product barriers to perform their functions. Adequate 
compensatory measures are available to mitigate a breach in the 
control room, spent fuel pool room and penetration room pressure 
boundaries. The probability of a loss of coolant accident that would 
place demands on these systems during a period that the ventilation 
system pressure boundaries would be allowed to be inoperable has 
been shown to be very small. In addition, proposed administrative 
controls eliminate the potential for a fuel handling accident, with 
potential to exceed dose limits, while the spent fuel pool room 
boundary room is breached. Therefore, the proposed changes do not 
result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: John A. Nakoski.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of amendment request: September 3, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) by: (1) Modifying the wording 
of the current Surveillance Requirement (SR) 4.0.1 and SR 4.0.3 to be 
consistent with NUREG-1431, Revision 2, Improved Standard Technical 
Specifications (ISTS) wording for SR 3.0.1 and SR 3.0.3; (2) modifying 
the current TS 6.8 by adding a new subsection 6.8.j, which will include 
the NUREG-1431, Revision 2, ISTS wording for TS 5.5.14 that discusses 
the TS Bases Control Program; and (3) modifying the ISTS wording, 
adopted in item 1 above, to allow a delay period of 24 hours or up to 
the surveillance frequency interval, whichever is greater, and to 
require a risk analysis to be performed for any surveillance greater 
than 24 hours.
    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 (CLIIP). 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). Tennessee Valley Authority 
reviewed the following proposed NSHC determination published in the 
Federal Register as part of the CLIIP for Technical Specification Task 
Force (TSTF)-358, and concluded in its application of September 3, 
2002, that the proposed NSHC determination applied to Sequoyah.
    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:

Adoption of TSTF-358, Revision 6--Missed Surveillances

A. 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.

B. 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.

C. 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.

    In addition to the above determination of NSHC, the licensee has 
provided its analysis for the following proposed NSHC determination:
    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 for the adoption of NUREG-1431, Revision 2, for 
Surveillance Requirements 3.0.1 and 3.0.3 wording and for the adoption 
of NUREG-1431, Revision 2, Technical Specification Bases Control 
Program, both of which are presented below:

Adoption of NUREG-1431, Revision 2, for Surveillance Requirements 3.0.1 
and 3.0.3 Wording

A. The Proposed Change Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated

    The proposed change involves rewording of existing Specification 
4.0.1 and 4.0.3 to be consistent with NUREG-1431, Revision 2. These 
modifications involve no technical changes to the existing TS 
[Technical Specifications]. This change is administrative in nature 
and does not affect initiators of analyzed events or assumed 
mitigation of accident or transient events. Therefore, this

[[Page 68746]]

change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

B. The Proposed Change Does Not Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated

    The proposed change involves the rewording of the existing 
Specification 4.0.1 and 4.0.3 to be consistent with NUREG-1431, 
Revision 2. The change does not involve a physical alteration of the 
plant (no new or different type of equipment installed) or changes 
in the methods governing normal plant operation. The change will not 
impose any new or different requirements or eliminate any existing 
requirements. Therefore, the proposed change does not create the 
probability of a new or different kind of accident from any accident 
previously evaluated.

C. The Proposed Change Does Not Involve a Significant Reduction in the 
Margin of Safety

    The proposed change involves rewording of the existing 
Specification 4.0.1 and 4.0.3 to be consistent with NUREG-1431, 
Revision 2. The change is administrative in nature and will not 
involve any technical changes. The change will not reduce a margin 
of safety because it has no impact on any safety analysis 
assumptions. Since this change is administrative in nature, no 
question of safety is involved. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

Adoption of NUREG-1431, Revision 2, Technical Specification Bases 
Control Program

A. The Proposed Change Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated

    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into the SQN 
[Sequoyah Nuclear Plant] Units 1 and 2 TS. This change involves no 
technical change to existing TS, it simply adds wording on how the 
bases section of the TS will be maintained and controlled. This 
change is administrative in nature and does not affect initiators or 
analyzed events or assumed mitigation of accident or transient 
events. Therefore, this change does not involve a significant 
increase in the probability or consequences or an accident 
previously evaluated.

B. The Proposed Change Does Not Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated

    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into the SQN Units 1 
and 2 TS. The change does not involve a physical alteration of the 
plant (no new or different type of equipment installed) or changes 
in the methods governing normal plant operation. The change will not 
impose any new or different requirements or eliminate any existing 
requirements. Therefore, the proposed change does not create the 
probability of a new or different kind of accident from any accident 
previously evaluated.

C. The Proposed Change Does Not Involve a Significant Reduction in the 
Margin of Safety

    The proposed change involves incorporation of the NUREG-1431, 
Revision 2, Bases Control Program requirements into SQN Units 1 and 
2 TS. The change is administrative in nature and will not involve 
any technical changes. The change will not reduce a margin of safety 
because they have not impact on any safety analysis assumptions. 
Since this change is administrative in nature, no question of safety 
is involved. Therefore, the proposed 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: October 1, 2002.
    Description of amendment request: The amendment would add a phrase 
to Limiting Condition for Operation (LCO) 3.1.8, ``Physics Tests 
Exceptions--Mode 2,'' of the technical specifications (TSs). The phrase 
to be added is that the number of required channels for certain 
functions in Table 3.3.1-1 of LCO 3.3.1, ``RTS Instrumentation,'' may 
be reduced from four to three required channels. LCO 3.1.8 applies to 
reactor Mode 2 during physics tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance [for the proposed change] 
will remain within the bounds of the previously performed accident 
analyses since there are no permanent hardware changes. The design 
of the RTS [reactor trip system] instrumentation will be unaffected; 
only the manner in which the system is connected for short duration 
physics testing is being changed to allow the temporary bypass of 
one power range channel. The reactor protection system will continue 
to function in a manner consistent with the plant design basis since 
a sufficient number of power range channels will remain OPERABLE to 
assure the capability of protective functions, even with a 
postulated single failure. [The number of required channels for 
certain functions in Table 3.3.1-1 is only being reduced from 4 to 3 
channels.] All design, material, and construction standards that 
were applicable prior to the request are maintained.
    The proposed change will allow the temporary bypass of one power 
range neutron flux channel during the performance of low power 
physics testing in MODE 2. This results in a temporary change to the 
coincidence logic from one-out-of-three under the current TS (with a 
trip imposed on the channel used for physics testing) to two-out-of-
three under the proposed TS (the channel used for physics testing 
would be in a bypassed state). However, this two-out-of-three 
coincidence logic still supports [the] required protection and 
control system applications, while reducing plant susceptibility to 
a spurious reactor trip.
    The proposed change will not affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters or accident mitigation performance.
    The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the USAR [Wolf Creek Updated Safety Analysis Report].
    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.
    There are no permanent hardware changes nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function. This change will not affect the normal 
method of power operation or change any operating parameters. No 
performance requirements will be affected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    The proposed amendment does not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System (other than as discussed above), or Solid State Protection 
System used in the plant protection systems. [The number of the 
required channels is not an initiator of an accident.]
    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.
    There will be no effect on the manner in which safety limits or 
limiting safety system

[[Page 68747]]

settings are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protective 
functions. There will be no impact on the overpower limit, departure 
from nucleate boiling ratio (DNBR) limits, heat flux hot channel 
factor (FQ), nuclear enthalpy rise hot channel factor (F'H), loss of 
coolant accident peak cladding temperature (LOCA PCT), peak local 
power density, or any other margin of safety. The radiological dose 
consequence acceptance criteria listed in the Standard Review Plan 
will continue to be met.
    The proposed change does not eliminate any RTS surveillance or 
alter the Frequency of surveillances required by the Technical 
Specifications. The nominal RTS and Engineered Safety Features 
Actuation System (ESFAS) trip setpoints (TS Bases Tables B 3.3.1-1 
and B 3.3.2-1), RTS and ESFAS allowable values (TS Tables 3.3.1-1 
and 3.3.2-1), and the safety analysis limits assumed in the 
transient and accident analyses [(USAR Table 15.0-4)] are unchanged. 
None of the acceptance criteria for any accident analysis is 
changed. The potential reduction in the frequency of spurious 
reactor trips would effectively increase the margin of safety or, at 
a minimum, be risk-neutral.
    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: April 26, 2002, as supplemented 
on July 11 and September 12, 2002.
    Brief description of amendment: The amendment revised Sections 2.3, 
``Limiting Safety System Settings,'' 3.1, ``Protective 
Instrumentation,'' and 3.10, ``Core Limits,'' of the Technical 
Specifications, and approved the use of flow control reference cards to 
support implementation of the Boiling Water Reactor Owners Group Option 
II solution for the long-term reactor stability problem.
    Date of Issuance: October 18, 2002.
    Effective date: October 18, 2002, and shall be implemented within 
30 days of issuance.
    Amendment No.: 235.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 28, 2002 (67 FR 
36926).
    The July 11 and September 12, 2002, letters provided clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated October 18, 
2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 19, 2002, as supplemented 
September 6, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Surveillance Requirement (SR) 4.0.3 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'' to ``* * * up to 24 hours or 
up to the limit of the specified surveillance interval, whichever is 
greater.'' In addition, the following requirement is added to SR 4.0.3: 
``A risk evaluation shall be performed for any surveillance delayed 
greater than 24 hours and the risk impact shall be managed.'' The 
amendment also makes administrative changes to SRs 4.0.1 and 4.0.3 to 
be consistent with NUREG-1432, Revision 2, ``Standard Technical 
Specifications, Combustion Engineering Plants.''
    Date of issuance: October 15, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 271.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: August 22, 2002 (67 FR 
54497).
    The September 6, 2002, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 15, 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: July 29, 2002.
    Brief description of amendments: The amendments revised Technical

[[Page 68748]]

Specification Surveillance Requirement 3.7.2.2 to decrease the 
allowable closure time for the turbine stop valves from 15 seconds to 1 
second.
    Date of Issuance: October 24, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 329, 329, 330.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56320).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 24, 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: July 18, 2002.
    Brief description of amendment: The amendment revises 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: October 8, 2002.
    Effective date: October 8, 2002, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 180.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56321).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 8, 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: October 17, 2001, as 
supplemented by letters dated February 26, August 14 and September 13, 
2002.
    Brief description of amendments: The amendments revise (1) 
Technical Specification (TS) Section 1.1, ``Definitions,'' for Dose 
Equivalent I-131, to allow the use of the thyroid dose conversion 
factors, listed in the International Commission on Radiological 
Protection Publication 30, ``Limits for Intakes of Radionuclides by 
Workers,'' and (2) Section 3.9.4, ``Containment Penetrations,'' to 
allow the equipment hatch, personnel air lock doors, and emergency air 
lock doors to remain open during core alterations and movement of 
irradiated fuel assemblies.
    Date of issuance: October 21, 2002.
    Effective date: October 21, 2002, to be implemented within 30 days 
from the date of issuance, including the completion of the 
administrative procedures that ensure that closure of the open 
containment penetrations, with direct access to the outside atmosphere 
during refueling operations with core alterations or irradiated fuel 
movement inside containment, will be initiated immediately in the event 
of a fuel handling accident inside containment, or if severe weather 
warnings are in effect.
    Amendment Nos.: Unit 1--155; Unit 2--155.
    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 
929).
    The supplemental letters dated February 26, August 14 and September 
13, 2002, provided additional clarifying information, 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 amendments is contained 
in a Safety Evaluation dated October 21, 2002.
    No significant hazards consideration comments received: No.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon

    Date of application for amendment: November 15, 2001 as 
supplemented by letters dated January 31, July 31, and October 3, 2002.
    Brief description of amendment: The amendment revises License 
Condition 2.C(10), ``Loading of Fuel into Casks in the Fuel Building,'' 
to license number NPF-1 for the Trojan Nuclear Plant (TNP). 
Specifically, these design changes are the result of the licensee's 
selection of Holtec International's design components (e.g., the Multi-
Purpose Cannister versus the Pressurized Water Reactor Basket. The new 
design basis limits impact the cask loading operations and contingency 
unloading in the Fuel Building.
    Date of issuance: October 21, 2002.
    Effective date: As of the date of issuance to be implemented and 
shall be implemented prior to placing Holtec International MPC's in the 
TNP ISFSI.
    Facility Operating License No. NPF-1: The amendment changes the 
cask loading and contingency unloading operations in the Fuel Building.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15626).
    The January 31, July 31, and October 3, 2002, supplemental letters 
provided clarifying information that did not change the scope of the 
original Federal Register (67 FR 15626) notice or the original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 21, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 4th day of November 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-28483 Filed 11-8-02; 8:45 am]
BILLING CODE 7590-01-P