[Federal Register Volume 67, Number 53 (Tuesday, March 19, 2002)]
[Notices]
[Pages 12597-12615]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-6230]


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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 February 22, 2002 through March 7, 2002. The 
last biweekly notice was published on March 5, 2002 (67 FR 10006).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room (PDR), located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By April 18, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first 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

[[Page 12598]]

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

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

    Date of amendment request: December 19, 2001.
    Description of amendment request: The proposed amendment provides 
clarifications and substantive changes to the decay heat removal (DHR) 
Technical Specifications (TSs). It is intended, in part, to fulfill a 
commitment made by the licensee to the NRC during a pre-decisional 
enforcement conference on April 23, 1999. Specifically, the proposed 
changes would: (1) define and clarify the emergency feedwater (EFW) 
flowpath redundancy as described in the Bases; (2) provide operability 
requirements for the redundant steam supply paths to the turbine-driven 
EFW pump; (3) provide a 72-hour allowed outage time (AOT) with any EFW 
pump or flowpath inoperable; (4) provide a 24-hour AOT with one steam 
supply path to the turbine-driven EFW pump and one motor-driven EFW 
pump inoperable; (5) provide a requirement to initiate action to 
immediately restore at least 2 EFW pumps and one flowpath to each once-
through steam generator (OTSG) if more than one EFW pump or both 
flowpaths to either OTSG were inoperable; (6) provide a statement 
suspending actions requiring shutdown or changes in reactor operating 
conditions until at least 2 EFW pumps and one EFW flowpath to each OTSG 
are restored to operable status; and (7) revise, relocate and clarify 
EFW pump and flowpath operability requirements during surveillance 
testing. Minor administrative and editorial changes are also proposed, 
including relocation of some requirements for clarity. A note is added 
to TS 4.9.1.1 and its related Bases to indicate that the surveillance 
is not applicable to the turbine driven EFW pump until 24 hours after 
exceeding 750 psig. A change to TS Table 3.5-2 and the Bases for TS 
3.5.5, ``Accident Monitoring Instrumentation,'' regarding the 
description of the pressurizer level instrument channels to reflect the 
replacement of Bailey transmitters was also included. Unrelated 
editorial changes to the Table of Contents were also included.
    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 the operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. This change incorporates the concept of EFW flowpath 
redundancy throughout the TS[s], which takes into consideration the 
redundancy provided by the EFW System modifications made in the mid-
1980s after the accident at TMI-2 [Three Mile Island Nuclear 
Station, Unit 2]. This change incorporates appropriate Limiting 
Conditions of [for] Operation (LCOs) and required action times and 
clarifies the design basis of the EFW System technical specification 
requirements in the LCOs and Surveillance Standards. These changes 
will not result in any change to the configuration of the EFW System 
as described in the SAR [Safety Analysis Report] or used in plant 
specific analyses. The reliability of EFW System components is 
unaffected. With less than the minimum EFW capability, this change 
incorporates the STS [standard technical specification] requirement 
to initiate action immediately to restore EFW components and suspend 
all actions requiring shutdown or changes in reactor operating 
conditions. The seriousness of this condition requires that action 
be started immediately to restore EFW components to operable status 
prior to power reductions that could result in a plant trip with no 
safety related means for conducting a cooldown. This change will not 
significantly affect any accident initiation sequence or the off 
site dose consequences of accidents that have been analyzed.
    The current surveillance standard contains EFW flowpath 
operability requirements being moved to the Limiting Conditions of 
[for] Operation (LCO) section in Chapter 3 and combined with the 
notes to define the EFW System operability requirements for EFW 
pumps and flowpaths during surveillance testing. The revised 
specification incorporates consideration of EFW flowpath redundancy 
consistent with HSPS [heat sink

[[Page 12599]]

protection system] train operability requirements and continues to 
require that compensatory measures be implemented to promptly 
restore components if EFW is needed during surveillance testing when 
more than one pump or both flowpath[s] to an OTSG are inoperable. 
The intent of this surveillance standard has been retained, which 
assures that the minimum number of EFW flowpaths to the OTSGs will 
be available with minimal operator action. The addition of a note, 
currently provided in the Standard Technical Specifications which 
permits a delay in performing the surveillance of the turbine-driven 
EFW Pump is needed to assure sufficient main steam pressure is 
available for performance of the test and does not significantly 
affect the reliability of the pump or the consequences of accidents 
previously evaluated.
    This change provides further assurance that EFW System design 
basis requirements will be met and does not affect EFW system 
configuration, setpoints, or reliability. These changes will not 
affect any accident initiation sequence and do not affect off site 
dose consequences of accidents that have been analyzed. The revised 
Accident Monitoring Instrumentation specification for the EFW flow 
instruments is needed to reflect the revised flowpath definition and 
does not change the intent or interpretation of this specification. 
The editorial changes included in this LCA [license change 
application] are intended to improve the clarity, consistency and 
readability of the TS[s], [and] do not change the intent or 
interpretation.
    Therefore, operation of the facility in accordance with this 
proposed change will 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. As a result of this change, no additional hardware is being 
added; and there will be no affect on EFW System design, operation 
as described in the SAR, or assumptions used in plant specific 
analyses. The requirement for three EFW Pumps and [associated] 
flowpaths to be operable for continuous plant operation is not 
affected by this change. Events involving the EFW System operation 
have been reviewed and determined to have no impact from these 
changes. The additional operability requirements, revised LCOs and 
surveillance standards, clarifications and changes to define EFW 
flowpath redundancy ensures minimum EFW component operability as 
credited in plant analyses. There are no changes included that could 
affect the plant beyond those accidents that have been evaluated. 
The editorial changes included in this LCA are intended to improve 
the clarity, consistency, and readability of the TS[s] and Bases, 
[and] do not change the intent or interpretation.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    No. This change does not affect EFW System design or 
instrumentation setpoints. The requirement for three operable EFW 
pumps and associated flowpaths is not affected by this change. This 
change revises the Limiting Conditions of [for] Operation (LCOs) for 
the EFW System, revises the required actions, impose[s] additional 
required action times, and provide[s] clarification of the LCO and 
Surveillance Standards. The revised LCO requires that at least one 
flowpath to each OTSG must be operable. The 8 hour action time 
currently allowed for pump inoperability during surveillance testing 
is also applied to flowpath inoperability during testing. The 
revised LCO continues to require compensatory measures during EFW 
testing when HSPS is required to be operable and an OTSG is 
isolated, retaining the provision that EFW flowpath valves can be 
realigned promptly from their test mode to their operational 
alignment if EFW flow is needed. None of these changes affect a 
margin of safety. The revised Accident Monitoring Instrumentation 
specification for the EFW flow instruments is needed to reflect the 
revised flowpath definition and does not change the intent or 
interpretation of this specification. The editorial changes included 
in this LCA are intended to improve the clarity, consistency, and 
readability of the TS[s], [and] do not change the intent or 
interpretation.
    Therefore, operation of the facility in accordance with this 
proposed change 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: Joel T. Munday, Acting.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: November 19, 2001.
    Description of amendment request: The amendment would revise the 
Technical Specification 5.5.16 to eliminate the requirement to perform 
post-modification containment integrated leakage rate testing following 
replacement of Unit 2 steam generators.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The steam generator replacement activities do not affect the 
containment structure or the actual containment liner. Access for 
the replacement steam generators as well as removal of the old steam 
generators will be through the equipment hatch. However, the outer 
shell of the steam generators, the inside containment portions of 
the main steam line, the feedwater lines, the auxiliary feedwater 
lines, and the steam generator blowdown lines are all part of the 
primary reactor containment boundary that will be impacted by the 
replacement activities.
    Calvert Cliffs Nuclear Power Plant Technical Specification 
5.5.16 states, ``A program shall be established to implement the 
leakage testing of the containment as required by 10 CFR 50.54(o) 
and 10 CFR Part 50, Appendix J, Option B. This program shall be in 
accordance with the guidelines contained in Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program,'' dated September 
1995, including errata.'' Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' endorses Nuclear Energy 
Institute (NEI)94-01, Revision 0 for methods acceptable to comply 
with the requirements of Option B. Prior to returning the 
Containment to operation, NEI 94-01 requires leakage rate testing 
(Type A testing or local leakage rate testing), following repairs 
and modification that affect the containment leakage integrity.
    The affected area of the primary containment boundary is also 
part of the pressure boundary of an American Society of Mechanical 
Engineers (ASME) Class 2 component/piping system and, as such, the 
planned replacement of the steam generators are subject to the 
repair and replacement requirements of ASME Section XI. The ASME 
Section XI surface examination, volumetric examination, and system 
pressure test requirements are more stringent than the Appendix J, 
Option B testing requirements. The acceptance criteria for ASME 
Section XI system pressure testing of welded joints in ``zero 
leakage.'' In addition, the test pressure for the system pressure 
test will be approximately 17 times that of Appendix J, Option B 
test.
    The objective of the Type A test is to assure the leak-tight 
integrity of the area affected by the modification. Although the 
leak test is in a direction reverse to that of the design basis 
accident environment, the ASME Section XI inspection and testing 
requirements more than fulfill the intent of the requirements of 
Appendix J, Option B with the exception of secondary side access 
manways. Section 9.2.1, NEI 94-01, Revision 0 allows reverse testing 
if justified. Section XI pressure test applies a sealing pressure to 
the secondary manway due to the inward door swing configuration. 
Hence, a Type B local leak rate test will be performed for the 
secondary

[[Page 12600]]

manways. For all other affected components, reverse testing is 
justified since the acceptance criteria for ASME Section XI system 
pressure testing of welded joints is ``zero leakage,'' and the test 
pressure for the system pressure test will be approximately 17 times 
that of Type A test. Hence, the probability or consequences of 
design basis accidents previously evaluated are unchanged.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment integrated leakage rate testing following replacement of 
Unit 2 steam generators will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    The proposed revision does not involve a physical change to the 
plant and there are no changes to the operation of the plant that 
could introduce a new failure mode. As described above in Item 1, 
the objective of the Appendix J, Option B test is to assure the 
leak-tight integrity of the area affected by the modification. The 
ASME Section XI inspection and testing requirements are more 
stringent than the Appendix J, Option B testing requirements.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment leakage integrated rate testing following replacement of 
Unit 2 steam generators will not create the possibility of a new or 
different [kind] of accident from any accident previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    As described above in Item 1, the ASME Section XI surface 
examination, volumetric examination, and system pressure test 
requirements are more stringent than the Appendix J, Option B 
testing requirements. The acceptance criteria for ASME Section XI 
system pressure testing of welded joints is ``zero leakage.'' In 
addition, the test pressure for the system pressure test will be 
approximately 17 times that of Appendix J, Option B test.
    Therefore, the proposed revision to Technical Specification 
5.5.16 to eliminate the requirement to perform post-modification 
containment integrated leakage rate testing following replacement of 
Unit 2 steam generators 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 E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Joel Munday, Acting.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: January 31, 2002.
    Description of amendments request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation (LCO), following a 
missed surveillance. The delay period would be extended from the 
current limit of ``... up to 24 hours or up to the limit of the 
specified Frequency, whichever is less'' to ``...up to 24 hours or up 
to the limit of the specified Frequency, whichever is greater.'' In 
addition, the following requirement would be added to SR 3.0.3: ``A 
risk evaluation shall be performed for any Surveillance delayed greater 
than 24 hours and the risk impact shall be managed.''
    The 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 January 31, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

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

[[Page 12601]]

proposes to determine that the amendments request involves no 
significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Joel Munday, Acting.

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

    Date of amendment request: December 20, 2001.
    Description of amendment request: The amendments would revise the 
Technical Specification (TS) 3.3.2, Engineered Safety Feature Actuation 
System (ESFAS) Instrumentation, and TS 3.3.5, Loss of Power Diesel 
Generator Start Instrumentation for Catawba Nuclear Station, Units 1 
and 2. These amendments would modify the subject TS as summarized 
below.
    1. Add a new MODE 3 operability requirement within ESFAS Function 5 
(Turbine Trip and Feedwater Isolation) as shown on TS Table 3.3.2-1; 
reformat TS Table 3.3.2-1 in regard to ESFAS Function 5; modify 
identified Conditions and Required Actions applicable within ESFAS 
Function 5; and modify the content and footnotes applicable to ESFAS 
Functions 5 and 6 (Auxiliary Feedwater).
    2. Delete ESFAS Functions 5e (Dog House Water Level--High High) and 
5f (Turbine Trip and Feedwater Isolation, Trip of all Main Feedwater 
Pumps).
    3. Modify the Conditions and Required Actions for ESFAS Function 6d 
(Auxiliary Feedwater, Loss of Offsite Power).
    4. Modify the Conditions and Required Actions for ESFAS Function 6e 
(Auxiliary Feedwater, Trip of all Main Feedwater Pumps).
    5. Modify the Conditions and Required Actions for ESFAS Function 6f 
(Auxiliary Feedwater, Auxiliary Feedwater Pump Train A and Train B 
Suction Transfer on Suction Pressure--Low).
    6. Make an editorial change to ESFAS Function 8 (ESFAS Interlocks, 
Tavg--Low Low, P-12).
    7. Add a new TS Surveillance Requirement (SR 3.3.2.12) for ESFAS 
Function 10 (Nuclear Service Water Suction Transfer--Low Pit Level).
    8. Add a note to Condition A of TS 3.3.5 which allows one channel 
per bus to be bypassed for surveillance testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes 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

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Approval of this amendment will have no effect 
on accident probabilities or consequences. For the proposed changes 
to Technical Specifications (TS) 3.3.2, Engineered Safety Features 
Actuation System (ESFAS); and 3.3.5, Loss of Power Diesel Generator 
Start Instrumentation; the equipment referenced in these TS is not 
accident initiating equipment. Therefore, there will be no impact on 
any accident probabilities caused by the NRC approval of this 
amendment. Additionally, since the design of the equipment is not 
being adversely modified by these proposed changes, there will be no 
impact on any accident consequences.

Second Standard

    Implementation of 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 the NRC approval of this license amendment request. 
No changes are being made to the plant which will introduce any new 
accident causal mechanisms. This amendment request does not impact 
any plant systems that are accident initiators; therefore, no new 
accident types are being created.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a 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. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The equipment referenced in the proposed change 
to TS 3.3.2 and 3.3.5 will remain capable of performing as designed. 
No safety margins will be impacted.

Conclusion

    Based upon the preceding discussion, Duke Energy Corporation has 
concluded that this proposed amendment 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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: February 6, 2002.
    Description of amendment request: The proposed amendment would 
remove the requirement for Main Steam Isolation Valve isolations on 
certain area temperatures from Technical Specifications Section 
3.3.6.1, ``Primary Containment and Drywell Isolation Instrumentation.''
    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.
    There is no credit taken in any licensing basis analysis for the 
main steam line isolation valve (MSIV) closure on the turbine area 
high temperature and there are no calculations that credit the 
subject isolation function as a mitigative feature. A review of 
Chapters 5, 6, 11, 12, and 15 of the USAR [Updated Safety Analysis 
Report] confirmed that the subject isolation function was not 
credited in any analysis for mitigating fuel cladding damage, 
mitigating challenges to vessel integrity, or mitigating dose to 
plant staff or the general public. This conclusion is consistent 
with the discussion of the function in the current Technical 
Specification [TS] Bases (B 3.3.6.1). Removing this requirement from 
the TS will allow the licensee to make changes to the design or 
function of the instrumentation provided the changes meet the 10 CFR 
50.59 criteria. Entergy intends to make changes that will reduce 
unwarranted challenges to the MSIVs, associated isolation and 
actuation logic, and minimize the likelihood of an unwarranted plant 
transient due to increased ambient temperatures for reasons other 
than a steam leak. Therefore, the proposed change does not involve a

[[Page 12602]]

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 removes the automatic MSIV isolation 
function associated with high temperatures in certain Turbine 
Building areas from the requirements of the Technical 
Specifications. Relocating requirements for this isolation function 
to licensee control does not introduce any new failure mechanisms or 
introduce any new accident precursors. Any subsequent changes to the 
design or function of the instrumentation must meet the criteria of 
10 CFR 50.59.
    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.
    There is no credit taken in any licensing basis analysis for the 
main steam line isolation (MSIV closure) on the turbine area high 
temperature. Therefore, since the MSIV isolation function on the 
Turbine Building Area High Temperature is not credited as a 
mitigating feature in any analysis which establishes thermal limits, 
evaluates peak vessel pressure, evaluates peak containment/drywell 
pressure, or evaluates radiological consequences (on and off site), 
there is no adverse impact on any margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment(s) present 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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    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: January 31, 2002.
    Description of amendment request: The proposed amendment would 
modify the Arkansas Nuclear One, Unit 2 (ANO-2) Facility Operating 
License (FOL) and Technical Specifications (TSs) to reorganize the 
Administrative Controls section (Section 6.0) to be consistent with 
NUREG-1432, ``Standard Technical Specifications Combustion Engineering 
Plants'' and provide consistency with the Arkansas Nuclear One, Unit 1 
(ANO-1) TS Section 6.0. This change would result in moving several 
surveillance requirements, currently contained in the Surveillance 
Requirements section of the ANO-2 TSs, and programs contained in the 
FOL, to Section 6.0. The change would also result in the deletion of 
several TSs currently contained in Section 6.0. A Bases Control Program 
would also be added to Section 6.0. The TS actions related to the 
Control Room Ventilation System would also be modified as part of the 
proposed amendment. The ventilation system (emergency and air 
conditioning system) for the control room is shared with ANO-1 and, 
thus, the TSs for this system are maintained consistent between the 
units where appropriate.
    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 modifies the Administrative Controls section 
of the ANO-2 TSs to be consistent with NUREG-1432. [The requirements 
of ] 10 CFR 50.36, ``Technical Specifications'' defines the 
Administrative Controls section as follows: ``Administrative 
controls are the provisions relating to organization and management, 
procedures, recordkeeping, review and audit, and reporting necessary 
to assure operation of the facility in a safe manner.'' Therefore, 
by definition the specifications contained in the Administrative 
Controls section are not specifications related to systems that are 
used to mitigate any types of accidents. The proposed changes to the 
Administrative Controls section therefore do not impact the ability 
of a plant system to perform its intended function.
    The proposed changes to the Control Room Ventilation System 
specifications do not result in any type of plant modification to 
this system. The system's intended function is to provide heating, 
ventilation, and air conditioning to ensure a suitable environment 
for equipment and station operator comfort and safety.
    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 will re-organize the ANO-2 Administrative 
Controls section and modify the actions related to the Control Room 
Ventilation System. The changes to the Administrative Controls 
section by definition of the type of specifications, which are 
included in the Administrative Controls section, will not create any 
new or different types of accidents.
    The modifications to the Control Room Ventilation System 
specifications result in providing clarity to existing actions and 
the addition of new actions. The addition of the new actions results 
in consistency between the ANO-1 and ANO-2 TSs. No design changes 
are proposed to the Control Room Ventilation System.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes result in the relocation of several 
surveillance requirements to the Administrative Controls section as 
well as the re-organization of the Administrative Controls Section 
of the ANO-2 TSs. In addition, clarification is added to the Control 
Room Ventilation System action statements that result in consistency 
between the ANO-1 and ANO-2 TSs. These changes do not affect the 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 20, 2002.
    Description of amendment request: The proposed amendment would 
revise Section 3.6.5 of the St. Lucie Unit 1 and 2 Technical 
Specifications (TS) to extend the allowed outage time for the 
Containment vacuum relief lines from 4 hours to 72 hours, in order to 
facilitate compliance with the Inservice Testing Program without 
placing the plants at risk for unnecessary shutdowns. The extended 
allowed outage time would provide sufficient time to perform the 
required surveillance tests and make any required adjustments on the

[[Page 12603]]

Containment vacuum relief valves. The proposed changes are consistent 
with NUREG-1432, ``Standard Technical Specifications 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) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not create any new system interactions 
and have no impact on operation or function of any system or 
equipment in a way that could cause an accident. The primary 
containment to annulus vacuum relief valves are part of the 
containment vacuum relief system and are not initiators of any 
events nor affect any accident initiators of any events previously 
analyzed in Chapters 6 or 15 of the UFSAR [Updated Final Safety 
Analysis Report].
    The primary containment to annulus vacuum relief valves are 
designed to mitigate the consequences of an inadvertent containment 
spray system actuation during normal plant operation. The UFSAR 
analysis determined that with one of the two containment vacuum 
lines failed, the resultant peak calculated external pressure load 
on the containment was less than the design external pressure 
loading of 0.7 psi. These proposed changes do not affect any of the 
assumptions used in the analysis. Hence, the consequences of the 
design basis accident previously evaluated do not change.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes do not alter the design, configuration, or 
method of operation of the plant. There is no change being made to 
the parameters within which the plant is operated. The setpoints at 
which the protective or mitigating actions are initiated are 
unaffected by this change. As such, no new failure modes are being 
introduced that would involve any potential initiating events that 
would create any new or different kind of accident.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not affect the bases used in or the 
results of the analysis to establish the margin of safety. The 
margin of safety is established through equipment design, operating 
parameters, and the setpoints at which automatic actions are 
initiated. None of these are impacted by the proposed change. The 
proposed change is acceptable because it assures at least one vacuum 
relief line will remain available in the event of a single failure. 
This further assures the ability to actuate upon demand for the 
purpose of mitigating the consequences of the design basis accident 
(inadvertent actuation of the containment spray system during normal 
operation). The remaining vacuum relief line provides sufficient 
vacuum relief capacity to prevent exceeding the design external 
pressure loading on containment of 0.7 psi.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment requests: February 22, 2002.
    Description of amendment requests: The proposed amendments would 
relocate technical specifications (TSs) 3/4.9.6, ``Refueling 
Operations--Manipulator Crane Operability'' and TSs 3/4.9.7, 
``Refueling Operations--Crane Travel--Spent Fuel Storage Pool 
Building,'' with associated Bases to the D. C. Cook 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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes are administrative in nature in that they 
result in relocation of requirements from TS 3/4.9.6 and 3/4.9.7, 
with associated Bases, to the CNP UFSAR. Changes to the UFSAR are 
controlled by 10 CFR 50.59. Regulation 10 CFR 50.59 requires that 
NRC approval be obtained prior to any change to the UFSAR that would 
result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated. Accordingly, the 
relocation of requirements from TS 3/4.9.6 and 3/4.9.7, with 
associated Bases to the CNP UFSAR provides continued protection from 
changes involving unapproved increases in the probability of 
occurrence of an accident. The relocation of the requirements of TS 
3/4.9.6 and 3/4.9.7 would not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, 
configuration of CNP or the manner in which it is operated. 
Therefore, the proposed change does not significantly increase the 
probability of occurrence of an accident previously evaluated.
    The proposed change to relocate TS 3/4.9.6 and 3/4.9.7, with 
associated Bases to the CNP UFSAR does not impact the consequences 
of an accident because there is no effect on the structures, systems 
and components that mitigate the effects of an accident, or the 
manner in which they are operated. In accordance with 10 CFR 50.59, 
if any proposed change to the UFSAR results in more than a minimal 
increase in the consequences of an accident previously evaluated, 
NRC review and approval is required prior to the change being made. 
Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/
4.9.7, with associated Bases to the CNP UFSAR provides continued 
protection from changes involving unapproved increases in the 
probability of in the consequences of an accident. Therefore, the 
relocation of requirements will not affect offsite doses, and the 
consequences of an accident previously evaluated are not 
significantly increased.
    The format changes improve the appearance of the affected pages 
but do not affect any requirements.
    Therefore, the probability of occurrence and the consequences of 
an accident previously evaluated are not significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to relocate TS 3/4.9.6 and 3/4.9.7, with 
associated Bases, to the CNP UFSAR does not create new accident 
causal mechanisms. Plant operation will not be affected by the 
proposed change and no new failure modes will be created. Regulation 
10 CFR 50.59 requires that NRC approval be obtained prior to any 
change to the UFSAR that would create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/
4.9.7, with associated Bases to the CNP UFSAR provides continued 
protection from unapproved changes involving new or different kinds 
of accidents.
    The format changes improve the appearance of the affected pages 
but do not affect any requirements.
    Therefore, the possibility of a new or different kind of 
accident from any previously evaluated is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to relocate the requirements from the TS to 
the UFSAR does not impact equipment design or operation and no 
changes are being made to the TS

[[Page 12604]]

required safety limits, safety system settings, or any safety 
margins associated with TS 3/4.9.6 and 3/4.9.7. Changes to the UFSAR 
are controlled under the 10 CFR 50.59 process, which requires a 
safety evaluation to be performed. If any proposed change to the 
UFSAR results in a design basis limit for a fission product barrier, 
as described in the UFSAR, being exceeded or altered or results in a 
departure from a method of evaluation described in the UFSAR used in 
establishing the design bases or in the safety analyses, NRC review 
and approval will be required prior to the change being made. 
Accordingly, the relocation of requirements from TS 3/4.9.6 and 3/
4.9.7, with associated Bases to the CNP UFSAR provides continued 
protection from changes involving a reduction in the margin of 
safety. The format changes improve the appearance of the affected 
pages but do not affect any requirements.
    Therefore, there is no 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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting Section Chief.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: August 16, 2001.
    Description of amendment request: The proposed amendment would 
terminate license jurisdiction for a portion of the Maine Yankee Atomic 
Power Station (Maine Yankee) site, thereby releasing these lands from 
Facility Operating License No. DPR-36. In part, the release of these 
lands will facilitate the donation of a portion of this property to an 
environmental organization pursuant to a Federal Energy Regulatory 
Commission approved settlement between Maine Yankee Atomic Power 
Company and its ratepayers. The lands donated will be used to create a 
nature preserve and an environmental education center.
    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 proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The requested license amendment involves release of land 
presently considered part of the Maine Yankee plant site under 
license DPR-36. The land in question is not used for any licensed 
activities. No radiological materials have historically been used on 
this land and the land will not be used to support ongoing 
decommissioning operations and activities.
    Most of the land to be released is outside the Exclusion Area 
Boundary and therefore is not affected by the consequences of any 
postulated accident. A small portion of the land is within the 
Exclusion Area Boundary. Maine Yankee will retain sufficient control 
over activities performed within this land through rights granted in 
the legal land conveyance documents to ensure that there is no 
impact on consequences from postulated accidents. Therefore, the 
release of the land from the [10 CFR] Part 50 license will not 
increase the probability or the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The requested amendment involves release of land presently 
considered part of the Maine Yankee plant site under license DPR-36. 
The land is not used for any licensed activities or decommissioning 
operations. The proposed action does not affect plant systems, 
structures or components in any way. The requested release of the 
land does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety defined in the statements of consideration 
for the final rule on the Radiological Criteria for License 
Termination is described as the margin between the 100 mrem/yr 
public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use. 
This margin of safety accounts for the potential effect of multiple 
sources of radiation exposure to the critical group. Additionally, 
the State of Maine, through legislation, has imposed a 10 mrem/yr 
all pathways limit, with no more than 4 mrem/yr attributable to 
drinking water sources. Since the area is non-impacted, there will 
be no additional dose to the average member of the critical group. 
Furthermore, the survey results described in Attachment III [of the 
August 16, 2001, application] demonstrate that residual 
radioactivity, if any, in the area is indistinguishable from 
background. Therefore, this proposed license 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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power 
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 21, 2002.
    Description of amendment request: The proposed amendment would 
relocate specific pressure, differential pressure, and flow values, as 
well as specific test methods, associated with certain Engineered 
Safeguards Features (ESF) pumps from the Technical Specifications to 
the Seabrook Station Technical Requirements Manual.
    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 to relocate the specific ESF pump pressure 
and flow criteria in the aforementioned Technical Specifications 
surveillance requirements to the Seabrook Station Technical 
Requirements Manual are administrative in nature and do not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, configuration of the facility or the 
manner in which it is operated. The proposed changes do not alter or 
prevent the ability or structures, systems, or components to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the 
Seabrook Station Updated Final Safety Analysis Report (UFSAR).
    The subject surveillance requirement criteria relocated to the 
Seabrook Station Technical Requirements Manual will continue to be 
administratively controlled. The Seabrook Station Technical 
Requirements is a licensee-controlled document, which contains 
certain technical requirements and is the implementing manual for 
the Technical Specification Improvement Program. Changes to these 
requirements are reviewed and approved in accordance with Seabrook 
Station Technical Specifications, Section 6.7.1.i, and as outlined 
in the Seabrook Station Technical Requirements Manual.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 12605]]

    The proposed changes do not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated. There are no changes to the source term or 
radiological release assumptions used in evaluating the radiological 
consequences in the Seabrook Station UFSAR. The proposed changes 
have no adverse impact on component or system interactions. The 
proposed changes will not adversely degrade the ability of systems, 
structures and components important to safety to perform their 
safety function nor change the response of any system, structure or 
component important to safety as described in the UFSAR. The 
proposed changes are administrative in nature and do not change the 
level of programmatic and procedural details of assuring operation 
of the facility in a safe manner. Since there are no changes to the 
design assumptions, conditions, configuration of the facility, or 
the manner in which the plant is operated and surveilled, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. The proposed changes are 
administrative in nature and do not reduce the level of programmatic 
or procedural controls associated with the activities presently 
performed via the aforementioned surveillance requirements.
    Future changes to the subject technical requirements will be 
reviewed and approved in accordance with Seabrook Station Technical 
Specifications, Section 6.7, and as outlined in North Atlantic 
[Energy Service Corporation]'s programs. Specifically, changes to 
the Seabrook Station Technical Requirements Manual require an 
evaluation pursuant to the provisions of 10 CFR 50.59 and review and 
approval by the Station Operation Review Committee (SORC) prior to 
implementation.
    Therefore, relocation of the specific pump pressure and flow 
criteria contained in the aforementioned Technical Specifications 
Surveillance Requirements to the Seabrook Station Technical 
Requirements Manual does not involve a significant reduction in the 
margin of safety provided in the existing specifications.

    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: William J. Quinlan, Esq., Assistant General 
Counsel, Northeast Utilities Service Company, PO Box 270, Hartford CT 
06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 11, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.6.4, ``Containment Pressure,'' to 
reduce the maximum allowable pressure from 3 pounds per square inch 
gauge (psig) to 2 psig. The licensee requests these proposed amendments 
to address a non-conservatism that was identified during reviews of the 
Point Beach, Units 1 and 2, accident analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The operability of containment ensures that radionuclides are 
contained within allowable limits during and following all credible 
accident conditions. The inoperability or failure of containment is 
not a design basis accident initiator or precursor. Therefore, the 
probability of an accident previously evaluated will not be 
significantly increased as a result of the proposed change. Because 
design limitations continue to be met and the integrity of the 
containment system pressure boundary is not challenged, the 
assumptions employed in the calculation of the offsite radiological 
doses remain valid. In addition, the radiological consequence 
analysis for the main steam line break (MSLB) is performed assuming 
the MSLB is outside of the containment. Therefore, the operability 
of the containment structure does not affect the results of the 
offsite dose or control room dose consequences.
    Therefore, the consequences of an accident previously evaluated 
will not be significantly increased as a result of the proposed 
change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. The evaluation of the effects of the proposed changes 
indicate that all design standards and applicable safety criteria 
limits are met. These changes, therefore, do not cause the 
initiation of any new or different accident nor create any new 
failure mechanisms.
    Equipment important to safety will continue to operate as 
designed. Component integrity is not challenged. The changes do not 
result in any event previously deemed incredible being made 
credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems. 
Therefore, operation of the Point Beach Nuclear Plant in accordance 
with the proposed amendments will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The containment functions to mitigate the effects of accidents. 
There are no new or significant changes to the initial conditions 
contributing to accident severity or consequences. The proposed 
modification will not otherwise affect the plant protective 
boundaries, will not cause a release of fission products to the 
public, nor will it degrade the performance of any other SSCs 
[structures, systems, and components] important to safety. Reducing 
the maximum allowed containment pressure limit is conservative in 
that it reduces the peak containment pressure that could result in 
the event of an accident. Therefore, reducing the maximum allowed 
containment pressure limit will not reduce the margin of safety. The 
added conservatism provides improvement to the design pressure 
margin resulting from the proposed change and will enhance 
protection against conditions resulting from a design basis 
accident, which will therefore provide a net benefit to radiological 
health and reactor safety.

Conclusion

    Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of any accident previously analyzed; 
will not result in a new or different kind of accident from any 
accident previously analyzed; and, does not result in a significant 
reduction in any margin of safety. Therefore, operation of PBNP 
[Point Beach Nuclear Plant] in accordance with the proposed 
amendments does not result in a significant hazards determination.

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

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

    Date of amendment requests: February 13, 2002.

[[Page 12606]]

    Description of amendment requests: The proposed amendments would 
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated February 13, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

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

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

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

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

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

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

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

    Date of amendment requests: February 22, 2002.
    Description of amendment requests: The proposed amendment would 
revise technical specifications (TSs) for San Onofre Nuclear Generating 
Station (SONGS), Units 2 and 3 relating to spent fuel storage. 
Specifically, TS 3.7.17, ``Fuel Storage Pool Boron Concentration'', TS 
3.7.18, ``Spent Fuel Assembly Storage'', and TS 4.3, ``Fuel Storage'' 
would be revised to remove credit for use of Boraflex, and to take 
credit for soluble boron, and to increase the required concentration of 
soluble boron in the spent fuel storage pool. Additionally, new TS 
5.5.2.16, ``Fuel Storage Program'' would be added to create a TS to 
control the Fuel Storage Program.
    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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No.

Dropped Fuel Assembly

    There is no significant increase in the probability of a fuel 
assembly drop accident in the spent fuel pool when assuming a 
complete loss of the Boraflex panels in the spent fuel pool racks 
and considering the presence of soluble boron in the spent fuel pool 
water for criticality control.
    The presence of soluble boron in the spent fuel pool water for 
criticality control does not increase the probability of a fuel 
assembly drop accident. The handling of the fuel assemblies in the 
spent fuel pool has always been performed in borated water, and the 
quantity of Boraflex remaining in the racks has no affect on the 
probability of such a drop accident.
    Southern California Edison (SCE) has performed a criticality 
analysis which shows that the consequences of a fuel assembly drop 
accident in the spent fuel pool are not affected when considering a 
complete loss of the Boraflex in the spent fuel racks and the 
presence of soluble boron. The rack Keff remains less 
than or equal to 0.95.

Fuel Misloading

    There is no significant increase in the probability of the 
accidental misloading of spent fuel assemblies into the spent fuel 
racks when assuming a complete loss of the Boraflex panels and 
considering the presence

[[Page 12607]]

of soluble boron in the pool water for criticality control. Fuel 
assembly placement will continue to be controlled pursuant to 
approved fuel handling procedures and will be in accordance with the 
Technical Specification Section 5.5.2.16, ``Fuel Storage Program,'' 
which will specify spent fuel rack storage configuration 
limitations.
    There is no increase in the consequences of the accidental 
misloading of a spent fuel assembly into the spent fuel racks. The 
criticality analysis, performed by SCE, demonstrates that the pool 
Keff will be maintained less than or equal to 0.95 
following an accidental misloading by the boron concentration of the 
pool. The proposed Technical Specification 3.7.17 will ensure that 
an adequate spent fuel pool boron concentration is maintained.

Significant Change in Spent Fuel Pool Temperature

    There is no significant increase in the probability of either 
the loss of normal cooling to the spent fuel pool water or a 
decrease in pool water temperature from a large emergency makeup 
when assuming a complete loss of the Boraflex panels and considering 
the presence of soluble boron in the spent fuel pool water. A high 
concentration [> 2000 parts per million (ppm)] of soluble boron has 
always been maintained in the spent fuel pool water. The proposed 
minimum boron concentration of 2000 ppm in Technical Specification 
3.7.17 will ensure that an adequate spent fuel pool concentration is 
maintained in the spent fuel pools.
    A loss of normal cooling to the spent fuel pool water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in water density, and when 
coupled with the assumption of a complete loss of Boraflex, may 
result in a positive reactivity addition. However, the additional 
negative reactivity provided by the boron concentration limit in the 
proposed Technical Specification 3.7.17 will compensate for the 
increased reactivity which could result from a loss of spent fuel 
pool cooling. Because adequate soluble boron will be maintained in 
the spent fuel pool water to maintain Keff less than or 
equal to 0.95, the consequences of a loss of normal cooling to the 
spent fuel pool will not be increased.
    A decrease in pool water temperature causes an increase in water 
density and may result in an increase in reactivity when the 
Boraflex panels are present in the racks. However, the additional 
negative reactivity provided by the boron concentration limit in the 
proposed Technical Specification 3.7.17, determined based on the 
conservative assumption of a complete loss of the Boraflex, will 
compensate for the increased reactivity which could result from a 
decrease in pool water temperature.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No.
    Criticality accidents in the spent fuel pool are not new or 
different. They have been analyzed in the Updated Final Safety 
Analysis Report (UFSAR) and in previous submittals to the Nuclear 
Regulatory Commission (NRC). Specific accidents considered and 
evaluated include fuel assembly drop, fuel assembly misloading in 
the racks, and spent fuel pool water temperature changes.
    The possibility for creating a new or different kind of accident 
is not credible. Neither Boraflex or soluble boron are accident 
initiators. The proposed change takes credit for soluble boron in 
the spent fuel pool while maintaining the necessary margin of 
safety. Because soluble boron has always been present in the spent 
fuel pool, a dilution of the spent fuel pool soluble boron has 
always been a possibility. However, this accident was not considered 
credible. For this proposed amendment, SCE performed a spent fuel 
pool dilution analysis, which demonstrated that a dilution of the 
boron concentration in the spent fuel pool water which could 
increase the rack Keff to greater than 0.95 (constituting 
a reduction of the required margin to criticality) is not a credible 
event. The requirement to maintain boron concentration in the spent 
fuel pool water for reactivity control will have no effect on normal 
pool operations and maintenance. There are no changes in equipment 
design or in plant configuration. This new requirement will not 
result in the installation of any new equipment or modification of 
any existing equipment.
    Therefore, the proposed change will not result in 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?
    No.
    The Technical Specification changes proposed by this License 
Amendment request and the resulting spent fuel storage operation 
limits will provide adequate safety margin to ensure that the stored 
fuel assembly array will always remain subcritical. Those limits are 
based on a San Onofre Nuclear Generating Station (SONGS), Units 2 
and 3 plant specific analysis performed in accordance with a 
methodology previously approved by the NRC.
    The proposed change takes partial credit for soluble boron in 
the spent fuel pool. SCE's analyses show that spent fuel storage 
requirements meet the following NRC acceptance criteria for 
preventing criticality outside the reactor:
    (1) The neutron multiplication factor, Keff, 
including all uncertainties, shall be less than 1.0 when flooded 
with unborated water, and,
    (2) The neutron multiplication factor, Keff, 
including all uncertainties, shall be less than or equal to 0.95 
when flooded with borated water.
    The criticality analysis utilized credit for soluble boron to 
ensure Keff will be less than or equal to 0.95 under 
normal circumstances, and storage configurations have been defined 
using a 95/95 Keff calculation to ensure that the spent 
fuel rack will be less than 1.0 with no soluble boron. Soluble boron 
credit is used to provide safety margin by maintaining 
Keff less than or equal to 0.95 including uncertainties, 
tolerances and accident conditions in the presence of spent fuel 
pool soluble boron. The loss of a substantial amount of soluble 
boron from the spent fuel pool water which could lead to 
Keff exceeding 0.95 has been evaluated and shown to not 
be credible.
    Also, the spent fuel rack Keff will remain less than 
1.0 with the spent fuel pool flooded with unborated water.
    Decay heat, radiological effects, and seismic loads are 
unchanged by the absence of Boraflex.
    Therefore, the proposed change does not involve a significant 
reduction in the plant's margin of safety.

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

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

    Date of amendment request: January 28, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TS) 4.4.5.3a, ``Steam Generator Surveillance 
Requirements,'' inservice inspection frequency requirements for Unit 1 
immediately after the first refueling outage (1RE09) and Unit 2 after 
the second refueling outage (2RE10). The change would allow a 40-month 
inspection interval after one inspection resulting in C-1 
classification, rather than two consecutive inspection resulting in C-1 
classification. The change is proposed to eliminate steam generator 
inspections, which will result in significant dose, schedule and cost 
savings.
    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.
    There is no direct increase in SG [steam generator] leakage 
because the proposed change does not alter the plant design. The 
scope of inspections performed during 1RE10, the first refueling 
outage following SG replacement, exceeded the TS requirements for 
the first two refueling outages after

[[Page 12608]]

replacement combined. That is, more tubes were inspected than were 
required by TS. Currently, South Texas Project Unit 1 does not have 
an active SG damage mechanism and will meet the current industry 
examination guidelines without performing inspections during the 
next refueling outage. The results of the Condition Monitoring 
Assessment after 1RE10 demonstrated that all performance criteria 
were met during 1RE10. The results of the 1RE10 Operational 
Assessment show that all performance criteria will be met over the 
proposed operating period. The results from 2RE10 inspections are 
expected to be the same. 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 will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections performed during 1RE10 and planned for 
2RE10, the first refueling outage for each unit following SG 
replacement, significantly exceed the TS requirements for the scope 
of the first two refueling outages after SG replacement combined.
    The proposed change does not affect the design of the SGs, the 
method of operation, or reactor coolant chemistry controls. No new 
equipment is being introduced and installed equipment is not being 
operated in a new or different manner. The proposed change involves 
a one-time extension to the SG tube inservice inspection frequency, 
and therefore will not give rise to new failure modes. In addition, 
the proposed change does not impact any other plant system or 
components. 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.
    Steam generator tube integrity is a function of design, 
environment, and current physical condition. Extending the SG tube 
inservice inspection frequency by one operating cycle will not alter 
the function or design of the SGs. Inspections conducted prior to 
placing the SGs into service (pre-service inspections) and 
inspection during the first refueling outage following SG 
replacement demonstrate that the SGs do not have fabrication damage 
or an active damage mechanism. The scope of those inspections 
significantly exceeded those required by the TS. These inspection 
results were comparable to similar inspection results for the same 
model of RSGs [replacement steam generators] installed at other 
plants, and subsequent inspections at those plants yielded results 
that support this extension request. The improved design of the 
replacement SGs also provides reasonable assurance that significant 
tube degradation is not likely to occur over the proposed operating 
period. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania Avenue, NW., 
Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 18, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3/4.6.1.7, ``Containment Ventilation 
System,'' to extend the intervals between operability tests of the 
normal and supplementary containment purge valves.
    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.
    Operability and leakage control effectiveness of the containment 
purge isolation valves have no effect on whether or not an accident 
occurs. Consequently, increasing the interval between surveillances 
of isolation valve effectiveness does not involve a significant 
increase in the probability of an accident previously evaluated.
    The consequences of a non-isolated reactor containment building 
at the time of a fuel-handling accident or LOCA [loss-of-coolant 
accident] is release of radionuclides to the environment. Analyses 
have conservatively assumed that a purge system line is open at the 
time of an accident, and release to the environment continues until 
the isolation valves are closed. In addition, LOCA analyses assume 
containment leakage of 0.3 percent per day for the first 24-hours 
and 0.15 percent per day thereafter. Consequently, increasing the 
interval between surveillances of isolation valve effectiveness does 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. The function of the containment 
purge systems is not altered by this change. Therefore, this 
proposed change does not create the possibility of an accident of a 
different kind than previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    This proposed change only increases the interval between 
surveillance tests of the containment purge valves. Analyses have 
conservatively assumed that the normal purge valves are open at the 
time of a fuel handling accident, and that purging by the 
supplementary purge system is in progress at the time of a loss of 
coolant accident. In addition, LOCA analyses assume containment 
leakage of 0.3 percent per day for the first 24-hours and 0.15 
percent per day thereafter. The radiological consequences of both a 
fuel handling accident and a LOCA are unchanged and remain within 
the 10 CFR 100 limits. Therefore, the proposed change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania Avenue, NW., 
Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: June 21, 2001, as supplemented on 
February 8, 2002.
    Description of amendment request: This amendment request proposes 
to revise the control rod block instrumentation requirements contained 
in Technical Specification (TS) 2.1.B, Figure 2.1.1, and Tables 3.2.5 
and 4.2.5. Some of the control rod block trip functions are being 
relocated to the Vermont Yankee Technical Requirements Manual and some 
of the requirements for the retained trip functions are being 
clarified. Two trip functions are added to the TSs and Note 9 to Table 
3.2.5 is changed to reflect one or two Rod Block Monitor channels 
inoperable. This proposed no significant hazards consideration 
determination replaces in its entirety the notice published in the 
Federal Register on July 25, 2001 (66 FR 38769).
    Basis for proposed no significant hazards consideration 
determination:

[[Page 12609]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration which is presented 
below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The relocated trip functions are not assumed as initial 
conditions for, nor are they credited in the mitigation of, any 
design basis accident or transient previously evaluated. Since 
reactor operation with these revised and relocated Specifications is 
fundamentally unchanged, no design or analytical acceptance criteria 
will be exceeded. As such, this change does not impact initiators of 
analyzed events, or the analyzed mitigation of design basis accident 
or transient events.
    More stringent requirements that ensure operability of equipment 
and purely administrative changes do not affect the initiation of 
any event, nor do they negatively impact the mitigation of any 
event. The addition of remedial actions to address a condition when 
both channels of the Rod Block Monitor (RBM) are inoperable also 
ensures that the RBM function is met. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    None of the proposed changes affects any parameters or 
conditions that could contribute to the initiation of any accident. 
No new accident modes are created since plant operation is unchanged 
in that required protective features remain operable. No safety-
related equipment or safety functions are altered as a result of 
these changes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    This change does not impact plant equipment, nor does it involve 
operation with loss of any safety function. There are no changes 
being made to safety limits or safety system settings that would 
adversely affect plant safety as a result of the proposed changes. 
Since the changes have no effect on any safety analysis assumptions 
or initial conditions, the margins of safety in the safety analyses 
are maintained. In addition, administrative changes that do not 
change technical requirements or meaning, and the imposition of more 
stringent or equivalent remedial requirements to ensure operability, 
have no negative impact on margins of safety. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: September 11, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification Section 3.9.5, ``Shutdown Cooling (SDC) and Coolant 
Circulation--Low Water Level,'' by adding a note that allows one SDC 
loop to be inoperable for a period of 2 hours provided the other loop 
is operable while in Mode 6.
    Date of issuance: March 1, 2002.
    Effective date: March 1, 2002, and shall be implemented within 45 
days of the date of issuance.
    Amendment Nos.: Unit 1-139, Unit 2-139, Unit 3-139.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59501).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2002.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: November 19, 2001.
    Brief description of amendments: The amendments authorized 
revisions to the Calvert Cliffs Nuclear Power Plant Updated Final 
Safety Analysis Report to incorporate revisions to the loss of 
feedwater flow analysis.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
in conformance with the scheduling requirements specified in 10 CFR 
50.71e.
    Amendment Nos.: 248, 224.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised Appendix C of the licenses.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
925).
    The Commission's related evaluation of these amendments is 
contained in a

[[Page 12610]]

Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: December 20, 2000, as 
supplemented on July 12, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specifications to incorporate changes required to support 
operation with replacement steam generators.
    Date of issuance: March 1, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to restart following replacement of the steam generators.
    Amendment Nos.: 249 and 225.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13799).
    The July 12, 2001, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 1, 2002.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: July 27, 2001.
    Brief description of amendments: The amendments modify the 
conditions and required actions for the control room emergency 
ventilation system (CREVS) and control room emergency temperature 
system (CRETS) of Technical Specifications (TS) 3.7.8 and 3.7.9. A Note 
is added to TS 3.7.8 and the Note for TS 3.7.9 is revised to specify 
train operability requirements during the movement of irradiated fuel 
assemblies.
    Date of issuance: March 4, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 250, 226.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46475).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 6, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) by decreasing the CNS Unit 1 
Overtemperature Delta Temperature Allowable Value and the CNS Units 1 
and 2 Overpower Delta Temperature Allowable Values in TS Table 3.3.1-1. 
In addition, the amendments make two minor editorial changes in the TS 
Table of Contents and Bases Page 3.3.1-10.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 195/188.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2920). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: March 5, 2001, as supplemented 
by letter dated September 4, 2001.
    Brief description of amendments: The amendment revised the 
Technical Specifications to allow a one-time extension to the interval 
for conducting the 10 CFR part 50, Appendix J containment integrated 
leak rate test.
    Date of Issuance: February 28, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 321.
    Renewed Facility Operating License No. DPR-55: Amendment revised 
the Technical Specification.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50466). The supplement dated September 4, 2001, provided clarifying 
information that did not change the scope of the March 5, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 31, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications by replacing the peak linear heat rate safety limit with 
a peak fuel centerline temperature safety limit.
    Date of issuance: March 4, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 238.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 2002 (67 
FR 6279). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 10, 2001, as supplemented by letter 
dated December 20, 2001.
    Brief description of amendment: Technical Specification (TS) 
Surveillance Requirement (SR) 4.8.1.1.2.e requires certain emergency 
diesel generator (EDG) surveillances be performed during shutdown. This 
change modifies this SR to allow performance of specific surveillances 
during any mode of plant operation. This provides the flexibility in 
the scheduling of testing activities consistent with online maintenance 
activities and improves EDG availability during plant shutdown periods.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 180.
    Facility Operating License No. NPF-38: The amendment revised the 
TS.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44168).
    The December 20, 2001, supplemental letter contained clarifying 
information that did not change the scope of the July 10, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.

[[Page 12611]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 31, 2002.
    Brief description of amendment: The amendment replaces the 
Technical Specification (TS) Safety Limit 2.1.1.2, ``Peak Linear Heat 
Rate,'' (PLHR) with a Peak Fuel Centerline Temperature Safety Limit and 
updates the Index accordingly. The associated TS Bases changes have 
been made to appropriately reflect the proposed new Safety Limit.
    Date of issuance: March 5, 2002.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 181.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 2002 (67 
FR 6281).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 14, 2001.
    Brief description of amendments: The amendments deleted Technical 
Specification (TS) Figures 5.1-1, ``Site Area Map''; and 5.1-2, ``Plant 
Area Map''; and replaced TS 5.1, ``Site,'' with a site location 
description. Conforming changes also deleted TS 5.1.1, ``Exclusion 
Area''; TS 5.1.2, ``Low Population Zone''; and TS 5.1.3, ``Map Defining 
Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid 
Effluents''; from TS 5.1 and the TS Index.
    Date of issuance: February 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 219 and 213.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34284). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 12, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: July 18, 2001.
    Brief description of amendments: The amendments revised Turkey 
Point Units 3 and 4 Technical Specifications, Section 6.0, 
``Administrative Controls.'' The revision consists of changing the 
title of the corporate executive responsible for overall nuclear plant 
safety from ``President--Nuclear Division'' to ``Chief Nuclear 
Officer.''
    Date of issuance: February 21, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 220 and 214.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41622). 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 21, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 1, 2001.
    Brief description of amendments: The amendments revise technical 
specification (TS) surveillance requirements (SR) 4.8.2.3.2.c.2 and 
4.8.2.5.2.c.2 and associated TS bases concerning the safety-related 
batteries to make them more consistent with the Westinghouse Standard 
TSs.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 266 and 247.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64296). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: October 5, 2001, as revised on 
January 4, 2002.
    Brief description of amendment: The amendment imposes a new license 
condition in Operating License NPF-69 to approve a change in the 
licensing basis regarding post-safety-injection hydrogen monitoring. 
Specifically, the amendment changes the permissible delay from 30 
minutes to 90 minutes.
    Date of issuance: February 25, 2002.
    Effective date: As of the date of issuance to be implemented during 
Refueling Outage 8.
    Amendment No.: 102.
    Facility Operating License No. NPF:-69 Amendment revises the the 
operating license.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2925). The staff's related evaluation of the amendment is contained in 
a Safety Evaluation dated February 25, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: March 29, 2001, as supplemented 
October 30, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS), Section 3.8.5, ``DC [Direct Current] Sources--
Shutdown,'' restoring the operability requirement to what it was before 
the TS was converted to the Improved Standard Technical Specifications 
format (i.e., Amendment No. 91).
    Date of issuance: March 1, 2002.
    Effective date: As of the date of issuance, to be implemented prior 
to Refueling Outage 8.
    Amendment No.: 103.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29359).
    The licensee's October 30, 2001, letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The staff's related evaluation of 
the amendment is contained in a Safety Evaluation dated March 1, 2002.
    No significant hazards consideration comments received: No.

[[Page 12612]]

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

    Date of application for amendment: August 15, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) to extend the channel calibration surveillance 
frequency for the automatic depressurization system timers from 18 
months to 24 months.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 245.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50469). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 2, 2001, as supplemented 
March 29, September 14, and December 27, 2001.
    Brief description of amendment: The amendment changes the Technical 
Specifications to increase the limits on stored fuel enrichments and 
provide other more flexible fuel loading constraints for the storage 
racks for new and spent fuel.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 1, 2001 (66 FR 
29844). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 21, 2001.
    Brief description of amendment: The amendment adds three topical 
report references to Technical Specification (TS) 5.9.5, ``Core 
Operating Limits Report.''
    Date of issuance: March 4, 2002.
    Effective date: March 4, 2002, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 203.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66471).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 18, 2001, as 
supplemented February 5, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications surveillance requirement 3.4.3.1 for testing 
of the main steam safety relief valves to permit the setpoint tolerance 
for ``as-found'' testing to be changed from 1 percent to 
3 percent. An editorial change will also be made to remove 
a note regarding an associated relief request.
    Date of issuance: March 7, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented during the spring 2002, and spring 2003, refueling and 
inspection outages for Units 1 and 2, respectively.
    Amendment Nos.: 201, 175.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59511).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 5, 2001, as supplemented 
on December 17, 2001.
    Brief description of amendment: The amendment revises License 
Condition 2.E in the Facility Operating License (FOL) to reflect 
Nuclear Regulatory Commission (NRC) staff approval of a change to the 
Salem-Hope Creek Security Plan and the Salem-Hope Creek Security 
Training and Qualification Plan. The specific change reviewed and 
approved by the NRC staff will allow illumination levels to be 
maintained at a minimum of 0.2 footcandle in the isolation zone while 
allowing lighting in the remainder of the protected area to be 
sufficient as determined by the licensee, rather than requiring a 
minimum 0.2 footcandle illumination level in the entire protected area.
    Date of issuance: February 22, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 138.
    Facility Operating License No. NPF-57: This amendment revised the 
FOL.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36343). The letter dated December 17, 2001, withdrew a portion of the 
March 5, 2001, application which would have changed the escort 
requirements for vehicles in the protected area. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated February 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 5, 2001, as supplemented 
on December 17, 2001.
    Brief description of amendments: The amendments revise License 
Condition 2.E in each of the respective Facility Operating Licenses 
(FOLs) to reflect Nuclear Regulatory Commission (NRC) staff approval of 
a change to the Salem-Hope Creek Security Plan and the Salem-Hope Creek 
Security Training and Qualification Plan. The specific change reviewed 
and approved by the NRC staff will allow illumination levels to be 
maintained at a minimum of 0.2 footcandle in the isolation zone while 
allowing lighting in the remainder of the protected area to be 
sufficient as determined by the licensee, rather than requiring a 
minimum 0.2 footcandle illumination level in the entire protected area.
    Date of issuance: February 22, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 250 and 230.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised each of the respective FOLs.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34288). The letter dated December 17, 2001, withdrew a portion of the 
March 5, 2001, application which would have changed the escort 
requirements for vehicles in the protected area. The

[[Page 12613]]

Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 19, 2001.
    Brief description of amendment: This amendment revises V. C. Summer 
Technical Specification 3.4.6.2.f by increasing the allowable 
operational leakage rate for 23 of the 35 reactor coolant system 
pressure isolation valves listed in TS Table 3.4-1. This change 
implements a size-dependent allowable leakage rate of 0.5 gallon per 
minute per nominal inch of valve diameter, up to a maximum of 5 gallons 
per minute per valve.
    Date of issuance: February 14, 2002.
    Effective date: February 14, 2002.
    Amendment No.: 154.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 8, 2001 (66 FR 
41626). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 14, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 19, 2001.
    Brief description of amendment: This amendment revises Technical 
Specifications (TS) Table 3.7-1 by lowering the maximum allowable power 
range neutron flux high setpoints when one or more main steam line 
safety valves are inoperable. The Bases for TS 3/4.7.1.1 is also 
revised to include the algorithm used for determining the new allowable 
values.
    Date of issuance: February 21, 2002.
    Effective date: February 21, 2002.
    Amendment No.: 155.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57125).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 20, 2001.
    Brief description of amendment: This amendment revises the 
Technical Specifications by adding a footnote to Table 3.3-3 regarding 
the Steam Line Isolation and Engineered Safety Feature Actuation System 
(ESFAS) functions. This revision will allow V.C. Summer to exclude 
ESFAS steam line isolation instrumentation operability in Mode 3 when 
the main steam isolation valves, along with associated bypass valves, 
are closed and disabled, and eases the restriction of Specification 
3.0.4 when performing reactor coolant system resistance temperature 
device cross calibrations at temperatures below the ESFAS P-12 
Interlock for Low-Low Tavg.
    Date of issuance: March 5, 2002.
    Effective date: March 5, 2002.
    Amendment No.: 156.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64301). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 5, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 11, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting 
Air,'' to support the use of California Diesel fuel rather than the 
existing Environmental Protection Agency Clear diesel fuel, and reflect 
a change in the diesel generator load profile in Modes 1 through 4.
    Date of issuance: March 5, 2002.
    Effective date: March 5, 2002, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2--183; Unit 3--174.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
932). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 5, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear 
Plant, Unit 1, Appling County, Georgia

    Date of application for amendment: August 31, 2001, supplemented 
January 24, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow a one-time deferral of the Type A Containment 
Integrated Leak Rate test based on the risk-informed guidance in 
Regulatory Guide 1.174.
    Date of issuance: February 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 226.
    Facility Operating License No. DPR-57: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52802). The supplement dated January 24, 2002, provided clarifying 
information that did not change the scope of the August 31, 2001, 
application nor the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated February 20, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 31, 2001, supplemented 
November 15, 2001, and February 21, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications on a one-time basis to extend from 7 days to 
14 days the completion time for the required actions associated with 
restoration of the 1B emergency diesel generator (EDG). The NRC review 
of the August 31, 2001, amendment request to extend the completion 
times for all of the EDGs to

[[Page 12614]]

14 days on a permanent basis is ongoing.
    Date of issuance: February 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 227/169.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52803). The supplemental letters dated November 15, 2001, and February 
21, 2002, provide clarifying information that did not change the scope 
of the August 31, 2001, application nor the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated February 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: December 14, 2001.
    Brief Description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: March 6, 2002.
    Effective date: As of the date of issuance and shall be implemented 
by August 1, 2002.
    Amendment Nos.: 153/145.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications and associated Bases.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2928). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 6, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 10, 2001, as 
supplemented February 11, 2002.
    Description of amendment request: The amendments revised Technical 
Specification 3.9.1, ``Refueling Equipment Interlocks,'' to allow in-
vessel fuel movement to continue with inoperable refueling equipment 
interlocks, provided (1) control rod withdrawals are blocked and (2) 
all control rods are verified to be inserted.
    Date of issuance: March 6, 2002.
    Effective date: March 6, 2002.
    Amendment Nos.: 242, 274, and 232.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57126). The February 11, 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 
March 6, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 20, 2001, as supplemented 
October 29, 2001, and November 14, 2001.
    Brief description of amendment: Amends the Final Safety Analysis 
Report by changing the spent fuel pool (SFP) cooling analysis 
methodology to increase the evaluated heat removal capacity of the SFP 
cooling system.
    Date of issuance: February 21, 2002.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 30 days.
    Amendment No.: 37.
    Facility Operating License No. NPF-90: Amendment does not revise 
the operating license or its appendices.
    Date of initial notice in Federal Register: December 17, 2001 (66 
FR 64998). The supplemental letters provided clarifying information 
that was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 21, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 10, 2000, as supplemented 
by letters dated September 18, 2000, August 22, 2001, November 8, 2001, 
and January 15, 2002.
    Brief description of amendment: The amendment incorporates new 
requirements into the Technical Specifications (TS) associated with 
steam generator (SG) tube inspection and repair, establishing an 
alternate voltage-based SG tube repair criteria.
    Date of issuance: February 26, 2002.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup following the Cycle 4 refueling outage.
    Amendment No.: 38.
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34751). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 18, 2001.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: February 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 92 and 92.

[[Page 12615]]

    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2931). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 11, 2001.
    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 SR 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: March 4, 2002.
    Effective date: March 4, 2002, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 143.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
935). The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 11th day of March 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-6230 Filed 3-18-02; 8:45 am]
BILLING CODE 7590-01-P