[Federal Register Volume 67, Number 247 (Tuesday, December 24, 2002)]
[Notices]
[Pages 78515-78528]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-32081]


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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

[[Page 78516]]

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, November 25, through December 12, 2002. The 
last biweekly notice was published on December 10, 2002 (67 FR 75867).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By January 23, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
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    \1\ The most recent version of title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29, 
2002.
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner 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

[[Page 78517]]

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

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

    Date of amendment request: November 8, 2002.
    Description of amendment request: The proposed amendment would 
delete sections 3.15.3 and 4.12.3, ``Auxiliary and Fuel Handling 
Building Air Treatment System,'' of the TMI Unit 1 Technical 
Specifications (TSs) and their corresponding Bases. Various minor 
typographical corrections and other administrative corrections are also 
proposed.
    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.
    This change will delete the existing Technical Specifications 
3.15.3 and 4.12.3. It does not impact nor change the physical 
configuration of any system, structure or component, nor does it 
change the manner in which any system is operated. Any change to the 
system design will be evaluated in accordance with the requirements 
of [title 10 of the Code of Federal Regulations (10 CFR)] 10 CFR 
50.59. Failure of the AFHBVS [Auxiliary and Fuel Handling Building 
Ventilation System] will neither initiate any type of accident nor 
increase the severity of the consequences of an accident.
    Previously approved analyses of the dose consequences of the 
accidents described in the TMI Unit 1 UFSAR [Updated Final Safety 
Analysis Report] confirmed that potential dose consequences were 
below the limits of 10 CFR 100 or 10 CFR 50.67 without the operation 
of the AFHBVS. These analyses are not affected by the proposed 
Technical Specification change. Thus the AFHBVS is not required for 
mitigation of any accident as described in TMI Unit 1 UFSAR chapter 
14.
    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.
    This activity will delete sections of the Technical 
Specifications applicable to the AFHBVS. This change does not 
physically alter any system, structure or component. Any change to 
the system design will be evaluated in accordance with the 
requirements of 10 CFR 50.59. The proposed change will not cause the 
AFHBVS to operate outside its design basis. There will be no impact 
to any operational feature of the system or any procedures that 
control its operation. The design basis of the AFHBVS as described 
in the UFSAR is not revised.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The deletion of Technical Specification sections 3.15.3 and 
4.12.3 will not impact the operation of the Auxiliary Fuel Handling 
Building Air Treatment System or the Fuel Handling Building ESF 
(engineered safety features) Ventilation system. The proposed change 
will not cause these systems to be placed in a configuration outside 
of their design basis nor will it reduce the margin of safety of 
these systems. The AFHBVS will continue to be operable in accordance 
with the applicable plant operating procedures. The AFHBVS will also 
continue to be tested and maintained under periodic operations 
surveillance and the TMI Unit 1 Preventive Maintenance Program.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 24, 2002, as supplemented 
November 21, 2002.
    Description of amendment request: The proposed amendments would

[[Page 78518]]

revise the Technical Specifications to extend the completion time for 
an inoperable train of low pressure injection from 72 hours to seven 
days. The proposed amendments are risk-informed.
    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:

    Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has 
made the determination that this amendment request involves a No 
Significant Hazards Consideration by applying the standards 
established by the NRC regulations in 10 CFR 50.92. The specific 
responses to the criterion are discussed 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 allows for one train of Low Pressure 
Injection to be inoperable for up to seven days. The Low Pressure 
Injection system is not an initiator for any accident previously 
evaluated and the consequences of an event during the extended 
Completion Time are no more severe than the consequences of the same 
event during the current Completion Time. Therefore, the 
consequences of an event previously analyzed are not increased. 
Consequently, 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 allows for one train of Low Pressure 
injection to be inoperable for up to seven days. 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 of governing normal plant operation. Therefore, the proposed 
changes 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 
the margin of safety?
    Response: No.
    The proposed change allows for one train of Low Pressure 
injection to be inoperable for up to seven days. An evaluation 
presented in Topical Report BAW-2295 and accepted by the NRC 
concluded that the extended Completion Time did not result in a 
significant reduction in the margin of safety. Therefore, the 
proposed changes does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 22, 2002.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the TS for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means or is of little use in the assessment and 
mitigation of accident conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System 
(PASS).'' The NRC staff issued a notice of opportunity for comment in 
the Federal Register on December 27, 2001 (66 FR 66949), on possible 
amendments concerning TSTF-413, 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 March 20, 2002 (67 FR 13027). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated October 22, 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 PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase

[[Page 78519]]

in the consequences of any accident previously evaluated.
    Criterion 2--The proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The proposed change does not involve a significant 
reduction in the margin of safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: August 19, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) 3/4.2, ``Protective 
Instrumentation,'' and TS 3/4.7, ``Containment Systems,'' by changing 
requirements associated with post-accident monitoring (PAM) 
instrumentation. This will reflect the guidance of the U.S. Nuclear 
Regulatory Commission Regulatory Guide 1.97, and adopt standard TS 
requirements for PAM instrumentation. The proposed amendment would also 
modify the associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously analyzed?
    Response: No.
    Post-Accident Monitoring (PAM) Instrumentation is not an 
initiator of any previously evaluated accident because there is no 
credible failure of PAM instrumentation that could initiate 
previously evaluated accidents. Therefore, the proposed changes do 
not involve a significant increase in the probability of an accident 
previously analyzed.
    The availability and use of PAM instrumentation help to ensure 
that the manual operator actions for mitigating an accident will be 
taken, and that the operator will be able to verify that automatic 
actions have occurred. The proposed changes make the requirements in 
the Technical Specifications more consistent with assumed operator 
actions. The proposed required actions, allowed out-of-service 
times, and surveillance intervals are appropriate based on operating 
experience, other instrumentation available, the passive nature of 
the instrument (no critical automatic action is assumed to occur 
from these instruments), and the low probability of an event 
requiring PAM instrumentation. Therefore, the proposed changes do 
not involve a significant increase in the consequences of an 
accident previously analyzed.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    Response: No.
    The proposed change does not involve the physical modification 
of structures[,] systems, or components, plant design basis, or the 
manner in which the plant is operated. PAM instrumentation is 
passive and does not initiate automatic actions. As a result, there 
are no credible failures that could initiate a new or different kind 
of accident from any accident previously evaluated.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    PAM instrumentation performs no automatic functions. PAM 
instruments help to ensure that operators take necessary manual 
actions to mitigate the consequences of an accident, and that 
operators have adequate information to confirm the operation of 
automatic accident mitigation functions have occurred. The proposed 
required actions, allowed out-of-service times, and surveillance 
intervals are appropriate based on operating experience, other 
instrumentation available, the passive nature of the instrument (no 
critical automatic action is assumed to occur from these 
instruments), and the low probability of an event requiring PAM 
instrumentation.
    Therefore, the proposed amendment does not involve a significant 
reduction in margin of safety.

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

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

    Date of amendment request: February 26, 2002, as revised on October 
9, 2002 and supplemented on October 30, 2002. This notice supersedes 67 
FR 34495 published on May 14, 2002, which was based on the licensee's 
application dated February 26, 2002.
    Description of amendment request: Revise the definition of Operable 
in Technical Specification (TS) 1.0.K with respect to support system 
requirements for AC power sources. Conforming changes are made to 
specific support system TSs in sections 3/4.5, ``Core and Containment 
Cooling Systems,'' 3/4.7, ``Station Containment Systems,'' and 3/4.10, 
``Auxiliary Electrical Power Systems,'' and associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed changes will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The revised definition of ``Operable'' redefines the AC power 
source requirements to allow either normal or emergency power 
available for equipment requiring AC power to be considered operable 
and provides conforming changes to specific supported system TSs. 
None of the proposed changes

[[Page 78520]]

affects any parameters or conditions that could contribute to the 
initiation of any accident. The proposed change does not affect the 
ability of the AC power sources to perform their required safety 
functions nor does the proposed change affect the ability of the 
systems requiring AC power to perform their respective safety 
functions. As a result, the ability of these systems to mitigate 
accident consequences is unchanged. As such, these changes do not 
impact initiators of analyzed events, nor the analyzed mitigation of 
design-basis accident or transient events.
    More stringent requirements for the inoperable AC power source 
action provisions that ensure availability of all TS required 
systems, subsystems, trains, components, and devices and the purely 
administrative changes do not affect the initiation of any event, 
nor do they negatively impact the mitigation of any event.
    The elimination of some explicit requirements to verify the 
operability of remaining equipment (i.e., to verify which TS action 
is required to be entered and taken) does not affect the initiation 
of any event, nor does it negatively impact the mitigation of any 
event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve any physical modification to 
the plant, change in TSs setpoints, change in plant design basis, or 
a change in the manner in which the plant is operated. No new of 
different type of equipment will be installed. No safety-related 
equipment or safety functions are altered as a result of these 
changes. In addition, there are no changes in methods governing 
normal plant operation. No new accident modes are created since 
plant operation is unchanged. None of the proposed changes affects 
any parameters or conditions that could contribute to the initiation 
of any accident. The changes do not introduce any new accident or 
malfunction mechanism that could create a new or different kind of 
accident, thus, no new failure mode is created. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes will not involve a significant reduction 
in a margin of safety.
    The manner in which plant systems relied upon in the safety 
analyses to provide plant protection is not changed. Plant safety 
margins continue to be maintained through the limitations 
established in the TSs Limiting Conditions for Operation and 
Actions. These changes do not impact plant equipment design or 
operation, and there are no changes being made to safety limits or 
safety system settings that would adversely affect the ability of 
the plant to respond as assumed in the accident analyses as a result 
of the proposed changes. Since the changes have no effect on any 
safety analysis assumptions or initial conditions, the margins of 
safety in the safety analyses are maintained.
    In addition, administrative changes that do not change technical 
requirements or meaning, and the imposition of more stringent 
requirements to ensure operability, have no negative impact on 
margins of safety.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based on this review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.

    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Andersen, Acting.

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

    Date of amendment request: November 22, 2002.
    Description of amendment request: The proposed amendment would 
allow for a one-time change to revise the steam generator (SG) 
inservice inspection frequency requirements in Technical Specification 
4.4.5.3.a to allow a 40-month inspection interval after one inspection, 
rather than after two consecutive inspections, based on the results 
falling into the C-1 classification.
    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 preciously evaluated?
    Response: No.
    There are no damage mechanisms that are active in the ANO-2 
(Arkansas Nuclear One, Unit 2) SGs that would prematurely create an 
accident or increase SG leakage. The scope of inspections performed 
during 2R15, the first refueling outage following SG replacement, 
exceeded the TS (technical specification) requirements for ensuring 
that the ANO-2 steam generator[s] fell into the C-1 category. The 
ANO-2 steam generator[s] meet the current industry examination 
guidelines without performing inspections during the next refueling 
outage. The results of the Condition Monitoring Assessment performed 
during 2R15 demonstrated that all performance criteria were met. The 
results of the 2R15 Operational Assessment show that all performance 
criteria are being met over the proposed operating period.
    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 accidents resulting from potential SG tube degradation. 
The scope of inspections performed during the 2R15 outage, the first 
refueling outage following steam generator replacement, exceeded the 
TS requirements.
    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 systems or 
components.
    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.
    Steam generator tube integrity is a function of design, 
environment, and current physical condition. Extending the steam 
generator tube inservice inspection frequence by one operating cycle 
will not alter their function or design. Inspections conducted prior 
to placing the SGs into service 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 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.

[[Page 78521]]

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

    Date of amendment request: August 16, 2002.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) Surveillance section 4.0.3 to 
extend the delay time for completion of a missed surveillance to 24 
hours or up to the surveillance frequency, whichever is greater. 
Additionally the proposed change would add a TS Bases Control Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    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.
    The relocation of two sentences from one specification to 
another in TS section 4.0, and the addition of a TS Bases Control 
Program in TS section 6.0, consistent with STS (Standard TS), is 
administrative in nature, does not affect the interpretation or 
execution of the TS, and has no effect on 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.
    The relocation of two sentences from one specification to 
another in TS section 4.0, and the addition of a TS Bases Control 
Program in TS section 6.0, consistent with STS, is administrative in 
nature, does not affect the interpretation or execution of the TS, 
and 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 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 and this change does not involve a 
significant reduction in a margin of safety.
    The relocation of two sentences from one specification to 
another in TS section 4.0, and the addition of a TS Bases Control 
Program in TS section 6.0, consistent with STS, is administrative in 
nature, does not affect the interpretation or execution of the TS, 
and does not involve a significant reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

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

    Date of amendment request: November 15, 2002.
    Description of amendment request: The proposed changes would revise 
the Safety Limit Minimum Critical Power Ratio (SLMCPR) for both two 
recirculation (dual) loop operation and single recirculation loop 
operation in Technical Specification (TS) 2.1.1.2 to reflect results of 
a cycle specific calculation performed for Cycle 22.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established, consistent with NRC 
approved methods, to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the safety limit for the minimum 
critical power ratio (SLMCPR) for Cooper Nuclear Station Cycle 22 
such that the fuel is protected during normal operation and during 
any plant transients or anticipated operational occurrences.
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits (MCPR) are 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criteria (i.e., that at least 99.9% of the 
fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences) is met. Since the 
operability of plant systems designed to mitigate any consequences 
of accidents has not changed, the consequences of an accident 
previously evaluated are not expected to increase.
    Based on the above NPPD [Nebraska Public Power District] 
concludes that the proposed changes do not involve a significant 
increase

[[Page 78522]]

in the probability or consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. The proposed change does not involve 
any modifications of the plant configuration or allowable modes of 
operation. The proposed change to the SLMCPR assures that safety 
criteria are maintained for Cycle 22.
    Based on the above NPPD concludes that the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The value of the proposed SLMCPR provides a margin of safety by 
ensuring that no more than 0.1% of the rods are expected to be in 
boiling transition if the MCPR limits is violated during all modes 
of operation. This will ensure that the fuel design safety criteria 
(i.e., that at least 99.9% of the fuel rods do not experience 
transition boiling during normal operation as well as anticipated 
operational occurrences) are met.
    Based on the above, NPPD concludes that the proposed changes do 
not involve a significant reduction in a margin of safety.
    From the above discussions, NPPD concludes that the proposed 
amendment involves 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: September 24, 2002.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specifications (TS) 3.4.11, ``Pressurizer Power 
Operated Relief Valves (PORVs),'' and the licensing basis to credit 
automatic actuation of the Class 1 power operated relief valves 
(PORVs), instead of the pressurizer safety valves (PSVs), to limit 
reactor coolant system pressure changes for the spurious operation of 
the safety injection system at power event, and other design basis 
accidents. Also, TS 3.4.10, ``Pressurizer Safety Valves,'' would be 
revised to allow PSV loop seal temperatures to be less than the lower 
design temperature during plant heatup and cooldown in Mode 3 and in 
Mode 4 when any reactor coolant system cold leg temperature is greater 
than the low temperature overpressure protection arming temperature 
specified in the pressure temperature limits report, provided at least 
one Class I PORV is available and capable of providing automatic 
pressure relief. This would allow gradual stabilization of the loop 
seal temperatures, and avoid having to partially drain the loop seals 
to establish the proper PSV inlet temperature.
    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.
    Part of the instrumentation for automatic control of the Class 1 
power operated relief valves (PORVs) during power operation is 
Instrument Class II. The automatic actuation circuitry will be 
upgraded to eliminate the Class II actuation circuitry, by providing 
output from the reactor protection system directly to the Class 1 
PORVs. This upgrade does not adversely affect the ability of the 
Class 1 PORVs to function to mitigate a reactor coolant system (RCS) 
overpressure condition, and would not increase the probability of a 
spurious opening of a PORV.
    The spurious operation of the safety injection (SI) system at 
power event is analyzed to assure that the RCS pressure limits are 
not exceeded, and that the departure from nucleate boiling ratio 
(DNBR) limits are met. The event is discussed in Final Safety 
Analysis Report (FSAR) Update Section 15.2.15. The current 
pressurizer overfill analysis takes credit for operation of the 
pressurizer safety valves (PSVs) to relieve a RCS overpressure 
condition. No credit is taken in the current analysis for automatic 
operation of the PORVs, which function to limit undesirable opening 
of the PSVs, since part of the automatic actuation circuitry is 
currently Instrument Class II. The current analysis that verifies 
that the DNBR limits are met remains bounding and was not 
reanalyzed.
    The spurious operation of the SI system at power event was 
reanalyzed for pressurizer overfill using a RETRAN02/Mod005.2 
computer code model of Diablo Canyon Power Plant. The analysis 
credits for automatic actuation of upgraded Class 1 PORVs to prevent 
water relief from the PSVs. Use of the Class 1 PORVs to perform any 
new safety related function would be evaluated in accordance with 10 
CFR 50.59.
    The RETRAN analysis demonstrates that the Class 1 PORVs can be 
expected to mitigate the consequences of a spurious operation of the 
SI system at power event, and that there is sufficient time for the 
operators to take action and open a PORV block valve(s) if closed.
    Crediting the PORVs in the pressurizer overfill case for the 
spurious operation of the SI system at power event does not increase 
the probability of the occurrence of the transient since the 
automatic opening of the PORVs for RCS pressure control is not an 
initiator for the event. This change allows for the acceptance 
criteria to be met for the spurious operation of the SI system at 
power event, ensuring that the consequences of this event remain 
within acceptable levels.
    The probability of a spurious operation of the SI system at 
power event is not affected by this proposed change and the above 
analysis demonstrates that the PORVs will adequately function in the 
automatic mode to mitigate the consequences of the transient. As 
such, there are no changes in the type or amount of any effluent 
released offsite as a result of this change.
    The proposed change would allow the PSV loop seal temperatures 
to be less than the lower design temperature during plant heatup and 
cooldown in Mode 3, and in Mode 4 when any RCS cold leg temperature 
is greater than the low temperature overpressure protection (LTOP) 
arming temperature specified in the pressure temperature limits 
report (PTLR), provided at least one Class 1 PORV is available and 
capable of providing automatic pressure relief. An evaluation of the 
applicable events in these modes indicates one Class 1 PORV is 
capable of preventing water relief from the PSVs and maintaining the 
reactor coolant pressure below 110 percent of its design value.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes would allow for automatic actuation of the 
Class 1 PORVs to be credited instead of the PSVs for the spurious 
operation of the SI system at power event. The proposed changes also 
allow the PSV loop seal temperatures to be less than the lower 
design temperature during plant heatup and cooldown in Mode 3, and 
in Mode 4 when any RCS cold leg temperature is greater than the LTOP 
arming temperature specified in the PTLR, provided at least one 
Class 1 PORV is available and capable of providing automatic 
pressure relief. Operation of the PORVs would prevent water relief 
from the PSVs, reducing the potential for a PSV not to properly 
reseat, and keep reactor coolant pressure below 110 percent of its 
design value. No new system interactions have been created, such 
that there is no increase in the possibility of a new or different 
kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different

[[Page 78523]]

kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes would allow for automatic actuation of the 
Class 1 PORVs to be credited instead of the PSVs for the spurious 
operation of the SI system at power event. The proposed changes 
allow the PSV loop seal temperatures to be less than the lower 
design temperature during plant heatup and cooldown in Mode 3, and 
in Mode 4 when any RCS cold leg temperature is greater than the LTOP 
arming temperature specified in the PTLR, provided at least one 
Class 1 PORV is available and capable of providing automatic 
pressure relief.
    The spurious operation of the SI system at power event is 
analyzed to assure that the RCS pressure limits are not exceeded, 
and that the DNBR limits are met. The current pressurizer overfill 
analysis takes credit for operation of the PSVs to relief a RCS 
overpressure condition. No credit is taken in the current analysis 
for automatic operation of the PORVs, since part of the PORV 
automatic actuation circuitry is currently Instrument Class II. 
Since the PORV function would limit undesirable opening of the PSVs, 
the automatic actuation circuitry will be upgraded so that the PORVs 
can be credited for accident mitigation. This change would 
specifically allow for automatic actuation of the upgraded Class 1 
PORVs to be credited instead of the PSVs in the accident analysis 
for the pressurizer overfill case.
    A reanalysis for pressurizer overfill takes credit for the 
upgraded PORVs and shows that they can be expected to mitigate the 
consequences of a spurious operation of the SI system at power 
event, and that there is sufficient time for the operators to take 
action and open a PORV block valve(s) if closed. The current DNBR 
analysis remains bounding and was not reanalyzed.
    The Class 1 PORVs will actuate to prevent water relief from the 
PSVs and keep reactor coolant pressure below 110 percent of its 
design value for a spurious operation of the SI system at power 
event. The conservative acceptance criteria for the current FSAR 
Update design analysis will continue to be met, and the margins of 
safety established in previous accident and transient analysis are 
not altered. The Class 1 PORVs will also provide overpressure 
protection during the period when the PSV loop seal temperature is 
less than the design limit.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 30, 2002.
    Description of amendment request: The proposed amendments would 
decrease the Control Room Emergency Outside Air Supply System (CREOASS) 
maximum allowed filter train pressure drop from <9.1 inches water gage 
(wg), to <7.3 inches wg in Technical Specification (TS) 5.5.7.d to 
correct an error in the maximum allowed value. The proposed maximum 
allowed pressure drop across a filter train is consistent with current 
design analyses and test acceptance criteria.
    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 decreases the maximum acceptable pressure 
loss through the Control Room Emergency Outside Air Supply System 
(CREOASS) filter train. A limit is placed on the filter train 
pressure loss to assure that the CREOASS can deliver the design 
flowrate assumed in the control room radiological consequence 
analysis presented in the SSES Final Safety Analysis Report (FSAR). 
The proposed change assures the system design flowrate will be met. 
Thus, the consequences of any accident previously evaluated are not 
increased. [The proposed change does not involve a physical 
difference or alteration of plant equipment (no new or different 
type of equipment will be installed) or a change in the methods 
governing normal plant operation. The proposed change does not 
change the design function or operation of the CREOASS.] The maximum 
allowable pressure drop through the CREOASS filter train is not an 
accident initiator thus, the probability of an accident previously 
evaluated is not increased. Therefore, the change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical modification or 
alteration of plant equipment (no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. The proposed change does not change the design function 
or operation of the CREOASS. Thus this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed action does not involve a significant reduction in 
a margin of safety. For the CREOASS, a lower maximum allowed 
pressure drop in TS does not adversely impact theoperation of any 
safety-related component or equipment. The proposed TS value is 
consistent with the design analysis and test acceptance criteria. 
Engineering evaluations concluded that there are no impacts on 
safety-related systems or accident analyses associated with the 
proposed change.
    The margin of safety is established through the design of plant 
structures, systems, and components, the parameters within which the 
plant is operated, and the establishment of setpoints for the 
actuation of equipment relied upon to respond to an event. The 
proposed change does not impact the condition or performance of 
structures, systems, and components relied upon for accident 
mitigation.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: October 31, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to incorporate generic change 
(Technical Specification Task Force) TSTF-306, Revision 2 to NUREG 
1433, ``Standard Technical Specifications for General Electric Plants 
(BWR/4),'' Revision 1, which has been approved by the NRC for adoption 
by licensees. Limiting Condition for Operation (LCO) 3.3.6.1, ``Primary 
Containment Isolation Instrumentation,'' would be revised to add an 
ACTIONS Note allowing intermittent opening, under administrative 
control, of penetration flow paths that are isolated to comply with 
ACTIONS, and to breakout Traversing Incore Probe (TIP) System isolation 
as a separate isolation function with an associated Required Action to

[[Page 78524]]

isolate the penetration within 24 hours rather than immediately 
initiate a unit shutdown. The associated Bases would also be revised in 
accordance with TS 5.5.10, ``TS Bases Control Program,'' to be 
consistent with TSTF-306, Revision 2, and to document the proposed 
changes and provide supporting information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability * * * or consequences of an accident previously 
evaluated?
    The proposed change relaxes Required Actions. Required Actions 
and their associated Completion Times are not initiating conditions 
for any accident previously evaluated. Further, the Required Actions 
in this change have been developed to provide assurance that 
appropriate remedial actions are taken in response to the degraded 
condition considering the operability status of the redundant 
systems of required features, [and] the capacity and capability of 
remaining features, while minimizing the risk associated with 
continued operation. Therefore, the relaxed Required Actions do not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The Required 
Actions and associated Completion Times in this change have been 
evaluated to ensure that no new accident initiators are introduced. 
Thus, this 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?
    The relaxed Required Actions do not involve a significant 
reduction in a margin of safety. As provided in the justification, 
this change has been evaluated to minimize the risk of continued 
operation under the specified Condition, considering the operability 
status off the redundant systems of required features, the capacity 
and capability of remaining features, a reasonable time for repair 
or replacement of required features, and the low probability of a 
design basis accident occuring during the repair period. Therefore, 
this 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 31, 2002.
    Description of amendment request: The proposed amendment would 
revise the SSES Technical Specification (TS) requirements for 
OPERABILITY of the Main Turbine Bypass System (MTBS) bypass valves. 
Specifically, Surveillance Requirement (SR) 3.7.6.1 would be revised to 
verify one complete cycle of only each required turbine bypass valve 
every 31 days. Currently this TS assumes all five main turbine bypass 
valves are required to be operable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability * * * or consequences of an accident previously 
evaluated?
    The proposed change provides LCO [Limiting Condition for 
Operation] requirements for operation of the facility that are 
consistent with the safety analyses. Since the safety analyses do 
not take credit for any margin provided by the fifth main turbine 
bypass valve, these LCO requirements do not result in operation that 
will increase the probability of initiating an analyzed event and do 
not alter assumptions relative to mitigation of an accident or 
transient event. The requirements continue to ensure process 
variables, structures, systems, and components are maintained 
consistent with the current safety analyses and licensing basis. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change does impose different requirements. However, the change is 
consistent with the assumptions in the current safety analyses and 
licensing basis, and has been evaluated to ensure that no new 
accident initiators are introduced. Thus this 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?
    The imposition of less restrictive LCO requirements does not 
involve a significant reduction in a margin of safety. As provided 
in the justification, this change has been evaluated to ensure that 
the current safety analyses and licensing basis requirements are 
maintained. This change does not involve a significant reduction in 
a margin of safety since the required number of main turbine bypass 
valves will be the number assumed in the safety analysis.

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

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

    Date of amendment request: November 19, 2002.
    Brief description of amendments: The proposed amendments would 
revise Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, 
Operating Licenses, Appendix B, ``Environmental Protection Plan,'' to 
revise and replace references to the U.S. Environmental Protection 
Agency's (EPA's) National Pollutant Discharge Elimination System 
(NPDES) permit. The EPA delegated the provisions of the NPDES permit 
for CPSES to the State of Texas, Texas Natural Resource Conservation 
Commission (currently the Texas Commission on Environmental Quality), 
in accordance with the rules and regulations of both agencies. In 
addition, minor administrative changes to the Environmental Protection 
Plan's description are also proposed to be consistent with provisions 
of the current Texas Pollutant Discharge Elmination System (TPDES) 
permit and the Final Environmental Statement for the Operating License.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 78525]]


    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No
    The requested changes involve an administrative correction to 
the Comanche Peak Steam Electric Station (CPSES) Operating Licenses, 
Appendix B ``Environmental Protection Plan'' to replace references 
to the U.S. Environmental Protection Agency's (EPA's) National 
Pollutant Discharge Elimination System (NPDES) permit with 
references to the current Texas Pollutant Discharge Elimination 
System (TPDES) permit. The continuing environmental regulatory 
provisions of the NPDES permit are incorporated and renewed in the 
current State of Texas TPDES permit. The change in permit issuing 
authority was achieved in a manner consistent with the rules and 
regulations of both the EPA and the Texas Natural Resource 
Conservation Commission (TNRCC) (currently the Texas Commission on 
Environmental Quality).
    Other minor changes proposed in the Environmental Protection 
Plan's description are administrative in nature and provide 
consistency with the provisions of the current TPDES permit and the 
NRC's [U.S. Nuclear Regulatory Commission] Final Environmental 
Statement--Operating License Stage.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed change. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed changes and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, the proposed changes 
do not create a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, Reactor 
Coolant System pressure boundary, and containment structure) to 
limit the level of radiation dose to the public. This request 
involves administrative changes only.
    No actual plant equipment or accident analyses will be affected 
by the proposed changes. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, the proposed changes do 
not involve a significant reduction in the margin of safety.

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

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: November 5, 2002.
    Description of amendment request: The proposed changes would revise 
the secondary coolant surveillance test requirements in table 4-2B, 
item 6, of the Technical Specifications (TS).
    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 revision to Technical Specifications deletes the 
secondary coolant sampling requirements for the fifteen minute 
degassed beta and gamma activity test required once per 72 hours and 
for the semiannual dose equivalent I-131 analysis in TS Table 4.1-
2B. The requirement for a dose equivalent I-131 analysis to be 
performed on a monthly basis remains in Table 4.1-2B. In accordance 
with the requirements of 10 CFR 50.92, the enclosed application is 
judged to involve no significant hazards based upon the following 
information:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change revises the sampling surveillance test 
requirements for the secondary coolant. Analyzed events are 
initiated by the failure of plant structures, systems, or 
components. The proposed change does not have a detrimental impact 
on the integrity of any plant structure, system, or component that 
could initiate an analyzed event. The proposed change will not alter 
the design and operation of, or otherwise increase the likelihood of 
failure of, any plant equipment that could initiate an analyzed 
accident.
    The deletion of the 15 minute degassed beta and gamma activity 
test once every 72 hours is a less restrictive change, while the 
deletion of the semiannual equivalent dose I-131 analysis is more 
restrictive. In view of the higher sensitivity of the liquid gamma 
isotopic test used in calculating the dose equivalent I-131, the 
proposed deletion of the 15 minute degassed beta and gamma activity 
test and the proposed monthly performance of the dose equivalent I-
131 analysis is appropriate. The dose equivalent I-131 analysis 
serves to confirm the validity of the safety analysis assumptions.
    As a result, the probability or consequences of any accident 
previously evaluated are not significantly affected by the proposed 
change in surveillance frequencies.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be 
installed) or a change in the method of plant operation. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    A limit on the specific activity of the secondary coolant is 
required in order to limit the radiological consequences of a main 
steam line break to a small fraction of the 10 CFR 100 criteria. The 
proposed sampling surveillance test requirements for the secondary 
coolant will verify that the TS-required specific activity limit is 
satisfied and will serve to confirm the validity of the safety 
analysis assumptions. Hence, the proposed change in sampling 
surveillance test requirements does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station, 
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, 
Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in

[[Page 78526]]

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 6, 2002.
    Brief description of amendments: The amendments replace the peak 
linear heat rate safety limit, in TS 2.1.1.2, ``Reactor Core SLs 
[Safety Limits],'' by a peak fuel centerline temperature safety limit.
    Date of issuance: December 2, 2002.
    Effective date: December 2, 2002, and shall be implemented within 
90 days of the date of issuance.
    Amendment Nos.: Unit 1-145, Unit 2-145, Unit 3-145.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66007). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 2, 2002.
    No significant hazards consideration comments received: No.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: July 7, 2000, as supplemented by letters 
dated June 14, July 31, August 15, August 22, September 6, September 7, 
2001, and May 9, June 26, August 15, August 20, and October 10, 2002.
    Brief description of amendment: The amendment adds a license 
condition which approves the License Termination Plan (LTP) for the 
Haddam Neck Plant, and provides the criteria by which the licensee may 
make changes to the LTP without prior NRC approval.
    Date of issuance: November 25, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 197.
    Facility Operating License No. DPR-61: The amendment adds a 
condition to the Facility Operating License.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77915).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 25, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 10, 2002.
    Brief description of amendment: This amendment revises Technical 
Specifications Surveillance Requirement 3.1.4.2 to extend the control 
rod scram time testing interval from 120 days to 200 days of full power 
operation.
    Date of issuance: December 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 126.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58641).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-10, Dresden Nuclear Power 
Station (DNPS), Unit 1, Grundy County, IL

    Date of amendment request: August 1, 2002.
    Brief description of amendments: The amendment revises the 
Operating License to update references to plant documents, deletes 
Technical Specification (TS) limiting conditions for required equipment 
and surveillance requirements that no longer apply or are being 
relocated to the Dresden Technical Requirements Manual, and deletes or 
revises TS administrative control and staffing requirements that either 
no longer apply or have changed due to the Unit 1 Fuel Storage Pool no 
longer containing spent fuel.
    Date of issuance: December 3, 2002.
    Effective date: December 3, 2002.
    Amendment No.: 41.
    Facility Operating License No. DPR-2: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58642).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 3, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 19, 2002, as supplemented 
by letters dated October 21 and November 8, 2002.
    Brief description of amendments: The amendments would extend the 
use of the current pressure and temperature (P/T) limit curves in 
Technical Specification (TS) 3.4.11, ``RCS Pressure and Temperature (P/
T) Limits,'' until December 15, 2004. The change will allow sufficient 
time for the incorporation of the General Electric Topical Report NEDC-
32983P, ``General Electric Methodology for Reactor Pressure Vessel Fast 
Neutron Flux Evaluation,'' methodology into the P/T curves in TS 
3.4.11.
    Date of issuance: December 3, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 156 & 142.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.

[[Page 78527]]

    Date of initial notice in Federal Register: October 30, 2002 (67 FR 
66170).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 3, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 12, 2002.
    Brief description of amendment: The amendment revised Technical 
Specifications sections 3.1.1 and 4.1.1, ``Control Rod System,'' by 
reducing the power level below which the rod worth minimizer or a 
second independent verification of rod position must be used from 20% 
to 10% rated thermal power.
    Date of issuance: December 9, 2002.
    Effective date: As of the date of issuance to be implemented before 
startup from Refueling Outage 17.
    Amendment No.: 178.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50957).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 9, 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: August 26, 2002.
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 3.0.3 to extend the delay period before entering a 
Limiting Condition for Operation following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: December 12, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61683).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: April 30, 2002, as supplemented 
June 26, August 29, October 3, October 23, and November 11, 2002.
    Brief description of amendments: These amendments increase the 
licensed reactor core power level by 1.4 percent from 1518.5 megawatts 
thermal (MWt) to 1540 MWt.
    Date of issuance: November 29, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 207 and 212.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2002 (67 
FR 57630). The June 26, August 29, October 3, October 23, and November 
11, 2002, supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the Nuclear Regulatory 
Commission staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated November 29, 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: July 23, 2002, as supplemented by 
letters dated October 8 and 28, 2002.
    Brief description of amendment: The amendment revises TS 2.5(1), 
``Steam and Feedwater Systems'' to: (1) remove the requirement to 
demonstrate operability of redundant auxiliary feedwater system 
components, and (2) provide an allowed outage time to restore 
operability of the emergency feedwater storage tank. In addition to 
these revisions, TS 2.5 has been revised to be more consistent with 
NUREG-1432, ``Improved Standard Technical Specification (ISTS) for 
Combustion Engineering Plants, Revision 2.''
    Date of issuance: November 26, 2002.
    Effective date: November 26, 2002, and shall be implemented within 
120 days from the date of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56327). The October 8 and 28, 2002, supplemental letters provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 26, 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: November 7, 2001, as supplemented by 
letter dated October 18, 2002.
    Brief Description of amendments: The amendments revise the 
operating licenses by replacing the license conditions concerning spent 
fuel cask lifting devices with a commitment to the requirements in 
American National Standards Institute N14.6-1978, ``Standard for 
Special Lifting Devices for Shipping Containers Weighing 10,000 lbs 
(4500 kg) or More for Nuclear Materials,'' in the Updated Final Safety 
Analysis Report.
    Date of issuance: December 2, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 158 and 149.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Operating License.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66013). The supplement dated October 18, 2002, provided clarifying 
information that did not change the scope of the November 7, 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 December 2, 2002.
    No significant hazards consideration comments received: No.

[[Page 78528]]

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

    Date of amendments request: May 23, 2002, as supplemented by letter 
dated October 31, 2002. The supplemental information provided 
clarification that did not change the scope or the initial no 
significant hazards consideration determination.
    Brief description of amendments: The amendments revise the 
technical specifications for the end-of-life moderator temperature 
coefficient surveillance requirements.
    Date of issuance: November 26, 2002.
    Effective date: Amendments are effective on the date of issuance 
and shall be implemented within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-144; Unit 2-132.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45572). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 26, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: May 14, 2002, as supplemented 
July 22, 2002.
    Brief Description of amendments: These amendments revise Technical 
Specifications section 4.5 and the associated Bases to change the 
surveillance frequency of the containment spray and recirculation spray 
header nozzles from a periodic surveillance of once every 10 years to a 
performance-based surveillance following maintenance that could cause 
nozzle blockage.
    Date of issuance: December 10, 2002.
    Effective date: December 10, 2002.
    Amendment Nos.: 232 and 232.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42831). The July 22, 2002, supplement contained clarifying information 
only and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 10, 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: September 27, 2001, as supplemented by 
letters dated June 27 and September 19, 2002.
    Brief description of amendment: The amendment revises section 
5.3.1.1, ``Unit Staff Qualifications,'' of the technical specifications 
to state new education and experience eligibility requirements for 
operator license applicants. As stated in the letter dated September 
19, 2002, the new requirements are outlined by the National Academy for 
Nuclear Training in its ``Guidelines for Initial Training and 
Qualification of Licensed Operators,'' which were issued January 2000.
    Date of issuance: November 26, 2002.
    Effective date: November 26, 2002, and shall be implemented within 
30 days of the date of issuance.
    Amendment No.: 150.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48223).
    The September 19, 2002, supplemental letter provided additional 
information that clarified the application, did not change the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 26, 2002.
    No significant hazards consideration comments received: No.


    Dated in Rockville, Maryland, this 16th day of December 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-32081 Filed 12-23-02; 8:45 am]
BILLING CODE 7590-01-P