[Federal Register Volume 68, Number 23 (Tuesday, February 4, 2003)]
[Notices]
[Pages 5668-5687]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-2415]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, January 10, 2003, through January 23, 2003. 
The last biweekly notice was published on January 21, 2003 (68 FR 
2796).

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, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By March 6, 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, 
Public File Area 01F21, 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

[[Page 5669]]

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 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, Public File Area 01F21, 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, Public File 
Area 01F21, 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, 301-415-4737 or by e-mail to 
[email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois; Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket 
Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle 
County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear 
Power Station, Units 1 and 2, Rock Island County, Illinois; Docket Nos. 
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, 
York County, Pennsylvania; Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: December 20, 2002.
    Description of amendment request: Nuclear Regulatory Commission 
(NRC) Regulatory Issue Summary 2002-05: ``NRC Approval of Boiling Water 
Reactor Pressure Vessel Integrated Surveillance Program,'' provides 
guidance on implementing the boiling water reactor (BWR) reactor 
pressure vessel integrated surveillance program (ISP). The amendment 
will modify the Updated Safety Analysis Reports (USARs) by removing the 
current facility reactor material surveillance capsule removal 
schedules from the facility USARs and specifying that these facilities 
will participate in an ISP developed by the BWR Vessel and Internals 
Project (BWRVIP). In addition, the Limerick Station will remove the 
current facility reactor material

[[Page 5670]]

specimen surveillance schedule from the Technical Specifications.
    With the exception of Oyster Creek, the USARs of each of the listed 
facilities contain a withdrawal schedule for the reactor pressure 
vessel material specimens. For those facilities which are not scheduled 
to remove a material specimen as part of the ISP (i.e., Clinton, Quad 
Cities, and Limerick), the proposed amendment would remove these plant-
specific schedules from the facility USARs and substitute a description 
of the facility's participation in the ISP. For those facilities which 
are scheduled to remove a capsule as part of the ISP (i.e., Dresden, 
LaSalle, and Peach Bottom), the proposed amendment would revise the 
material specimen withdrawal schedule in accordance with the ISP. 
Finally, for Oyster Creek, which is not scheduled to remove any further 
material specimens, the proposed amendment would revise the USAR to 
state that Oyster Creek will participate in the ISP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change adopts an integrated surveillance program 
(ISP) for reactor material specimen surveillances. The ISP ensures 
that the reactor pressure vessel (RPV) will continue to meet all 
applicable fracture toughness requirements. No physical changes to 
the facilities will result from the proposed change. The initial 
conditions and methodologies used in accident analyses remain 
unchanged. The proposed change does not revise or alter the design 
assumptions for systems or components used to mitigate the 
consequences of accidents. Thus, accident analyses results are not 
affected by this proposed change.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change adopts an ISP for reactor material specimen 
surveillances. The ISP ensures that the RPV will continue to meet 
all applicable fracture toughness requirements. No physical changes 
to the facilities will result from the proposed change.
    The proposed change does not affect the design or operation of 
any system, structure, or component (SSC) in the plant. The safety 
functions of the related SSCs are not changed in any manner, nor is 
the reliability of any SSC reduced. The change does not affect the 
manner by which the facility is operated and does not change any 
facility, structure, system, or component.
    No new or different type of equipment will be installed by this 
proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change has no impact on the margin of safety of any 
Technical Specification. There is no impact on safety limits or 
limiting safety system settings. The change does not affect any 
plant safety parameters or setpoints. No physical or operational 
changes to the facility will result from the proposed changes. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 19, 2002, as supplemented by 
letter dated December 19, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by: (1) Modifying the wording 
of the current Surveillance Requirements (SRs) 4.0.1 and 4.0.3 to be 
consistent with NUREG-1431, Revision 2, Improved Standard Technical 
Specifications (ISTS) wording for SR 3.0.1 and SR 3.0.3; and (2) 
modifying the ISTS wording, adopted in item 1 above, to allow a delay 
period of 24 hours or up to the surveillance frequency interval, 
whichever is greater, and to require a risk analysis to be performed 
for any surveillance greater than 24 hours consistent with Technical 
Specification Task Force (TSTF)-358 for missed surveillances.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the Consolidated Line Item Improvement Process (CLIIP). The NRC 
staff subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). Entergy Operations Inc. reviewed 
the following proposed NSHC determination published in the Federal 
Register as part of the CLIIP for TSTF-358, and concluded in its 
application of August 19, 2002, that the proposed NSHC determination 
applied to Waterford Steam Electric Station, Unit 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    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.
    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.

[[Page 5671]]

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

Proposed Changes to SR 4.0.1 and 4.0.3

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration for the 
adoption of NUREG-1431, Revision 2, for the revised SR 4.0.1 and 4.0.3 
wording. 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 change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change involves rewording of the existing SRs 4.0.1 
and 4.0.3 to be consistent with NUREG-1431, Revision 2. These 
modifications involve no technical changes to the existing TS. This 
change is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accident or transient events. 
Therefore, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change involves the rewording of the existing SR 4.0.1 
and 4.0.3 to be consistent with NUREG-1431, Revision 2. The change does 
not involve a physical alteration of the plant (no new or different 
type of equipment installed) or changes in the methods governing normal 
plant operation. The change will not impose any new or different 
requirements or eliminate any existing requirements. Therefore, the 
proposed change does not create the probability of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
the margin of safety.
    The proposed change involves rewording of the existing SRs 4.0.1 
and 4.0.3 to be consistent with NUREG-1431, Revision 2. The change is 
administrative in nature and will not involve any technical changes. 
The change will not reduce a margin of safety because it has no impact 
on any safety analysis assumptions. Since this change is administrative 
in nature, no question of safety is involved. Therefore, the proposed 
change does not involve a significant reduction in a margin of safety.
    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 16, 2002.
    Description of amendment request: The proposed amendment will 
revise the current main steam isolation valve (MSIV) Technical 
Specification (TS) 3/4 7.1.5 to more closely reflect TS 3.7.2 contained 
in NUREG-1432, Revision 2. In addition, this change will remove the 
MSIVs from the scope of containment isolation valve (CIV) TS 3/4 6.3 
such that only TS 3/4.7.1.5 will apply to the MSIVs. These changes will 
provide increased flexibility and clarity regarding the implementation 
of the TSs regarding MSIVs.
    Basis for proposed no significant hazard 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 to the applicability for the main steam line 
isolation valves will not require operability when all MSIVs are 
closed in Modes 2, 3, and 4. Analyzed events are assumed to be 
initiated by the failure of plant structures, systems or components. 
In the closed position the MSIVs are already in their safety 
function position. In this position, there can be no increase in the 
probability or consequences of an accident.
    The consequences of previously analyzed events are dependent on 
the initial conditions assumed for the analysis, and the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event. When the MSIVs are closed 
in Modes 2, 3, and 4 they are performing their design function for 
containment isolation and for main steam line isolation on the 
secondary side of the plant. The proposed change does not alter the 
initial conditions assumed in the safety analyses. The plant 
parameters assumed for the analyses are maintained within assumed 
limits through compliance with the Technical Specifications and 
plant procedures. Additionally, the proposed change does not impose 
any new safety analyses limits. Therefore, the proposed change does 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    The proposed change increases the allowed outage time for an 
inoperable MSIV from 4 hours to 8 hours in Mode 1 and for Modes 2, 
3, and 4; will allow both MSIVs to be inoperable, will allow 
separate action entry for the inoperable valves, and will allow 8 
hours to close each inoperable valve. Analyzed events are assumed to 
be initiated by the failure of plant structures, systems or 
components. Extending the time available to complete repairs of an 
inoperable component does not have a detrimental impact on the 
integrity of plant components nor does it increase the probability 
that these components will fail. The proposed changes are not 
related in any way to the probability of failure of a plant 
structure, system or component which would result in the occurrence 
of an analyzed event. Because the probability of failure of plant 
equipment is not affected, there is no impact on the probability of 
occurrence of a previously analyzed accident.
    The consequences of previously analyzed events are dependent on 
the initial conditions assumed for the analysis, and the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event. The steam line break 
analysis in FSAR [Final Safety Analysis Report] Section 15.1.3 
assumes a failure of one MSIV to close. For the containment 
isolation function, in the event of an inoperable MSIV coincident 
with a LOCA [loss-of-coolant accident], the closed system (i.e., the 
steam generator tubes and main steam line piping) remains intact. 
The closed system is subjected to a Type A containment leakage test, 
is missile protected, and [has] seismic category I piping, and 
typically has flow through it during normal operation such

[[Page 5672]]

that any loss of integrity could be continually observed through 
leakage detection systems within containment and system walkdowns 
outside containment. Therefore, with an inoperable MSIV the safety 
analysis (both LOCA and steam line break) remains valid assuming no 
additional failures. The increase in core damage frequency and large 
early release fraction, resulting from the increased restoration 
time, is negligible. The proposed 8 hour Allowed Outage Time is 
sufficiently short to ensure that the MSIVs are operable when 
required to perform their design function. Even though both MSIVs 
will be allowed under separate condition entry, to be inoperable in 
Modes 2, 3, and 4 the inoperable valves are still required to be 
closed. The 8 hour Allowed Outage Time to close an inoperable valve 
is based on the small likelihood of an accident occurring that will 
need the MSIV isolation function during this time period and the 
fact that the valves are located on a closed system with respect to 
containment integrity. The proposed change does not alter the 
initial conditions assumed in the safety analyses. The plant 
parameters assumed for the analyses are maintained within assumed 
limits through compliance with the Technical Specifications and 
plant procedures. Additionally, the proposed change does not impose 
any new safety analyses limits. Therefore, the proposed change does 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    The proposed change will add a Note to the MSIV surveillance to 
allow entry into Mode 3 for testing at hot conditions. Analyzed 
events are assumed to be initiated by the failure of plant 
structures, systems or components. The addition of this allowance 
for testing is not related in any way to the probability of failure 
of a plant structure, system or component which would result in the 
occurrence of an analyzed event. Because the probability of failure 
of plant equipment is not affected, there is no impact on the 
probability of occurrence of a previously analyzed accident.
    The consequences of previously analyzed events are dependent on 
the initial conditions assumed for the analysis, and the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event. The proposed change will 
allow entry into Mode 3 in order to perform MSIV testing at hot 
conditions. However, prior to this testing, the MSIVs are not known 
to be inoperable from any other cause other than not having 
performed the Surveillance Requirement to demonstrate closure times 
at hot plant conditions, which they are expected to pass. The 
proposed change will allow entry into Mode 3 for the condition where 
both MSIVs may require closure time testing. This testing allowance 
is limited to Mode 3, and must be completed prior to entry into 
Modes 1 or 2. The proposed change does not alter the initial 
conditions assumed in the safety analyses. The plant parameters 
assumed for the analyses are maintained within assumed limits 
through compliance with the Technical Specifications and plant 
procedures. Additionally, the proposed change does not impose any 
new safety analyses limits. Therefore, the proposed change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    The proposed change will require MSIVs, that are closed in 
accordance with the Mode 2, 3, and 4 Action, be verified closed once 
per seven days. Analyzed events are assumed to be initiated by the 
failure of plant structures, systems or components. The addition of 
this requirement is not related in any way to the probability of 
failure of a plant structure, system or component which would result 
in the occurrence of an analyzed event. Because the probability of 
failure of plant equipment is not affected, there is no impact on 
the probability of occurrence of a previously analyzed accident.
    The consequences of previously analyzed events are dependent on 
the initial conditions assumed for the analysis, and the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event. The proposed change adds 
a Surveillance Requirement to Technical Specification 3/4.7.1.5 to 
verify proper MSIV isolation on an actuation signal. This is not a 
new Surveillance Requirement for the Technical Specifications. 
Technical Specification 3.3.2, Engineering Safety Features Actuation 
System Instrumentation, Surveillance Requirement 4.3.2.1 (Table 4.3-
2 Item 4.d) requires a functional test of the actuation relay (K305) 
once per 18 months which verifies automatic closure of the MSIVs on 
a simulated main steam isolation signal. The proposed change does 
not alter the initial conditions assumed in the safety analyses. The 
plant parameters assumed for the analyses are maintained within 
assumed limits through compliance with the Technical Specifications 
and plant procedures. Additionally, the proposed change does not 
impose any new safety analyses limits. Therefore, the proposed 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    Therefore, none of the proposed change[s] described above 
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 alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which the plant is 
operated, or to the setpoints at which protective or mitigative 
actions are initiated. No alteration in the procedures which ensure 
the plant remains within analyzed limits is being proposed, and no 
change is being made to the procedures relied upon to respond to an 
off-normal event. As such, no new failure modes are being 
introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
limitations on operating parameters, and the setpoints at which 
automatic actions are initiated. No equipment design features are 
impacted by this change, no operating parameters are revised, and no 
changes to the actuation setpoints are involved.
    The design safety function of the MSIVs is to close upon receipt 
of a main steam isolation signal. With the MSIVs already closed in 
Modes 2, 3 or 4, the design function is satisfied.
    The proposed change will increase the allowed outage time from 4 
hours to 8 hours in Mode 1, for an inoperable MSIV. The proposed 
change will also relax current allowances for MSIVs in Modes 2, 3, 
and 4; however, the relaxations are in lower modes of operation 
where the potential for an accident that would require the MSIV 
isolation function is reduced. The proposed changes will still 
ensure that the inoperable MSIV(s) are restored or closed in a 
reasonable time of 8 hours. Once closed, the MSIVs meet their design 
safety function.
    The proposed change will add a note indicating the Surveillance 
Requirements must be performed prior to entry into Modes 1 or 2. The 
MSIVs are expected to pass the Surveillance Requirement and are not 
known to be inoperable for any other reason than not having 
performed the valve closure test at hot conditions. The testing is 
limited to Mode 3, when the reactor is subcritical, thus verifying 
the MSIV closure times prior to power operation.
    The proposed change will require MSIVs, which are closed in 
accordance with the Mode 2, 3, and 4 Action, be verified closed once 
per seven days. This requirement provides additional assurance that 
the MSIVs perform their design safety function to close.
    The proposed change adds a Surveillance Requirement to Technical 
Specification 3/4.7.1.5 to verify proper MSIV isolation on an 
actuation signal. This, however, is not a new Surveillance 
Requirement for the Technical Specifications. Technical 
Specification 3.3.2, Engineering Safety Features Actuation System 
Instrumentation, Surveillance Requirement 4.3.2.1 (Table 4.3-2 Item 
4.d) requires a functional test of the actuation relay (K305) once 
per 18 months which verifies automatic closure of the MSIVs on a 
simulated main steam isolation signal.
    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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street, NW., Washington, DC 20005-3502.

[[Page 5673]]

    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 16, 2002.
    Description of amendment request: The proposed amendment will add 
the topical report entitled ``Fuel Rod Maximum Allowable Gas 
Pressure,'' CEN-372-P-A, to the list of analytical methods in Technical 
Specification (TS) 6.9.1.11.1 used to determine the Waterford Steam 
Electric Station, Unit 3 (Waterford 3) core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously analyzed?
    Response: No.
    The proposed change does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident. The proposed 
change adds an NRC [Nuclear Regulatory Commission]-approved topical 
report to the list of analytical methods used to determine the core 
operating limits. The effect of the addition of this new reference 
is to revise the fuel design criterion for internal rod pressure to 
accept rod pressures that may exceed nominal Reactor Coolant System 
operating pressure. The use of this revised criterion continues to 
ensure that the consequences of an accident remain within acceptable 
limits. The change also proposes the administrative deletion of 
report date and revision levels in the list of references. These 
changes do not alter any of the assumptions or bounding conditions 
currently in the Final Safety Analysis Report.
    Waterford 3 performed a large break loss-of-coolant accident 
(LOCA) analysis using bounding fuel performance data as described in 
CEN-372-P-A. This analysis concluded that the peak cladding 
temperature remained within 10 CFR 50.46 limits.
    In addition to the LOCA analysis, an evaluation of the potential 
for departure from nucleate boiling (DNB) propagation was performed 
as described in CEN-372-P-A. The results confirmed that Waterford 3 
is bounded by the results evaluated in the topical report and that 
DNB propagation will not occur.
    Based on these analyses, there is no increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident. Accordingly, no 
new failure modes have been defined for any plant system or 
component important to safety nor has any new limiting failure been 
identified as a result of the proposed change. The intent of the 
proposed change is to reference an NRC-approved topical report in 
the Technical Specifications. The topical report justifies an 
acceptance criterion that allows fuel rod internal pressure to 
exceed RCS [reactor coolant system] pressure. There are no new 
accidents created by this change. An administrative aspect of this 
change, the deletion of date and revision levels, was also 
considered and does not create a new or different accident.
    The impact of fuel rod internal pressure exceeding reactor 
coolant system (RCS) pressure was considered in both an emergency 
core cooling system (ECCS) performance analysis and in a DNB 
propagation evaluation performed for Waterford 3. These two aspects 
were required considerations based on the NRC Safety Evaluation 
review of the topical report. The results demonstrated that 
Waterford 3 continues to meet 10 CFR 50.46 and that there is no 
potential for DNB propagation.
    Based on these analyses, there is no possibility of the creation 
of a new or different kind of accident from those previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds an NRC-approved topical report to the 
list of analytical methods used to determine core operating limits. 
It also deletes the revision number and dates associated with each 
of the topical reports listed. The effect of the addition of the new 
reference is to revise the fuel design criterion for fuel rod 
internal pressure to accept rod pressures that may exceed nominal 
RCS operating pressure. The use of this revised criterion continues 
to ensure that the consequences of an accident remain within 
acceptable limits. Since the core operating limits will continue to 
be established by an NRC-approved methodology and the results will 
be verified to meet the established acceptance criteria of 10 CFR 
50.46, the change will provide adequate core protection. Thus, the 
proposed amendment 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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 20, 2002.
    Description of amendment request: The proposed amendment makes 
several administrative changes to the Waterford Steam Electric Station, 
Unit 3, Technical Specifications (TSs) to revise, delete, correct, or 
clarify certain titles, page numbers, and heading information. The 
proposed amendment also revises personnel and committee titles that 
have been changed, revises administrative reporting requirements to 
conform to 10 CFR 50.4, and deletes redundant or unnecessary 
requirements from TSs 5.4, 6.6, and 6.7.
    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 changes are primarily to correct titles, page 
numbering errors, and otherwise make the TS index pages consistent 
with other NRC [U. S. Nuclear Regulatory Commission] approved pages. 
These changes are all of an administrative nature and have no effect 
on any plant equipment or structures. Therefore, these changes do 
not increase the probability or consequences of an accident 
previously evaluated.
    The proposed amendment also deletes TS 5.4.1 and 5.4.2. Values 
for RCS [Reactor Coolant System] design pressure, temperature, and 
volume are contained in the Final Safety Analysis Report. Any 
changes to these are controlled by 10 CFR 50.59. Therefore, removing 
the section from the TS will not increase the probability or 
consequences of previously evaluated accidents.
    The proposed amendment also deletes TS 6.6 and 6.7, and revises 
TS 6.9.1 and TS 6.9.2 to administratively conform reporting 
requirements to those in 10 CFR [part] 50. Therefore, removing these 
sections from the TS will not increase the probability or 
consequences of previously evaluated accidents.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
involve a physical alteration of the plant. No new or different 
equipment or modes of operation are being introduced by this 
proposed change. Thus, the changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

[[Page 5674]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes are primarily administrative in nature 
and can not affect any safety barriers. The proposed change to TS 
5.4 only deletes unnecessary information. 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: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: December 6, 2002.
    Description of amendment request: The proposed amendment would 
increase the surveillance interval of the Local Power Range Monitor 
(LPRM) calibrations from 1000 megawatt-days/ton to 2000 megawatt-days/
ton.
    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:

    Operation of the JAF [James A. FitzPatrick] plant in accordance 
with the proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:
    1. Involve an increase in the probability or consequences of an 
accident previously evaluated. The revised surveillance interval 
continues to ensure that the LPRM signal is adequately calibrated. 
The proposed change results in no change in radiological 
consequences of the design basis LOCA [loss-of-coolant accident] as 
currently analyzed for JAF. This change will not alter the basic 
operation of process variables, structures, systems, or components 
as described in the JAF UFSAR [Updated Final Safety Analysis 
Report], and no new equipment is introduced by the change in LPRM 
surveillance interval. The performance of the APRM [Average Power 
Range Monitor] and RBM [Rod Block Monitor] systems are not 
significantly affected by the proposed LPRM surveillance interval 
increase. Therefore, the probability of accidents previously 
evaluated is unchanged.
    The consequences of an accident can be affected by the thermal 
limits existing at the time of the postulated accident, but LPRM 
chamber exposure has no significant effect on the calculated thermal 
limits because LPRM accuracy does not significantly deviate with 
exposure. For the extended calibration interval, the total nodal 
power uncertainty remains less than the uncertainty assumed in the 
thermal analysis basis safety limit, maintaining the accuracy of the 
thermal limit calculation. Therefore, the thermal limit calculation 
is not significantly affected by LPRM calibration frequency, and the 
consequences of an accident previously evaluated are unchanged.
    The change does not affect the initiation of any event, nor does 
it negatively impact the mitigation of any event. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated. The proposed change will not 
physically alter the plant or its mode of operation. The performance 
of the APRM and RBM systems are not significantly affected by the 
proposed LPRM surveillance interval increase. As such, no new or 
different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing are consistent with current safety 
analysis assumptions. Therefore, the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety. The 
proposed change has no impact on equipment design or fundamental 
operation, and there are no changes being made to safety limits or 
safety system allowable values that would adversely affect plant 
safety as a result of the proposed change. The performance of the 
APRM and RBM systems are not significantly affected by the proposed 
LPRM surveillance interval increase. The margin of safety can be 
affected by the thermal limits existing prior to an accident; 
however, uncertainties associated with LPRM chamber exposure have no 
significant effect on the calculated thermal limits. The thermal 
limit calculation is not significantly affected because LPRM 
sensitivity with exposure is well defined. LPRM accuracy remains 
within the total nodal power uncertainty assumed in the thermal 
analysis basis, thus maintaining thermal limits and the safety 
margin.
    Since the proposed change does not affect safety analysis 
assumptions or initial conditions, the margin of safety in the 
safety analyses are maintained. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Richard J. Laufer.

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: January 9, 2003.
    Description of amendment request: The proposed Technical 
Specification (TS) amendment request changes the definition of a Logic 
System Functional Test, deletes the definition of a Simulated Automatic 
Actuation, clarifies Surveillance Requirement 4.5.G.1.a regarding 
simulated automatic actuation testing, and revises associated TS 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves surveillance requirements and 
definitions of surveillance tests. As such, the proposed change does 
not involve any plant physical changes, change any Technical 
Specification instrumentation setpoints, or introduce any new mode 
of plant operation. The proposed change to surveillance requirements 
and definitions does not result in any significant change in the 
availability of logic systems or safety-related systems themselves. 
Protective functions will be maintained. The proposed change does 
not degrade plant design, operation, or the performance of any 
safety system assumed to function in the accident analysis.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility for a new or different kind of 
accident from any previously evaluated.
    The proposed change does not: introduce any new accident 
initiators or failure mechanisms because the changes do not 
introduce any new modes of plant operation, make any physical 
changes (no new or different type of equipment will be installed); 
or change any Technical Specification instrumentation setpoints or 
methods of plant operation. The proposed changes will not 
substantially impose new requirements or eliminate any existing 
requirements.
    Therefore, the changes to the surveillance requirements and 
testing definitions that encompass this proposed change do not

[[Page 5675]]

create the possibility of a new or different kind of accident than 
those previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. There is no change or impact on any safety 
analysis assumptions. The proposed change does not involve any 
increase in calculated off-site dose consequences. Operability of 
protective instrumentation and the associated systems is unaffected, 
and performance of equipment will not be significantly affected. 
Since the proposed change is consistent with the BWR/4 Standard 
Technical Specifications, NUREG-1433, Revision 2, approved by the 
NRC [Nuclear Regulatory Commission] staff, revising the Technical 
Specifications in a manner which clarifies and reflects the approved 
level of detail ensures that safety margins are acceptable. 
Therefore, there is no significant reduction in the margin of safety 
as a result of this Technical Specification change.

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

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: March 14, 2002.
    Description of amendment request: The proposed license amendment 
request (LAR) will allow exercising and testing the Inclined Fuel 
Transfer System (IFTS) prior to the beginning of the refueling outage, 
thus increasing system reliability and refuel outage efficiency. The 
proposed LAR does not provide for the movement of fuel. The proposed 
LAR supplements Amendment No. 100 by including a time limit on the 
removal of the IFTS blind flange, providing a requirement to install 
the upper pool IFTS gate prior to IFTS blind flange removal, and 
limiting the unbolted configuration on the IFTS blind flange when it is 
rotated.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change permits removal of the Inclined Fuel 
Transfer System (IFTS) blind flange for a maximum duration of 60 
days per cycle when primary Containment operability is required in 
MODES 1 (Power Operation), 2 (Startup), or 3 (Hot Shutdown). The 
proposed change also limits the duration the IFTS blind flange may 
be unbolted when in MODES 1, 2, or 3. The proposed change does not 
involve modifications to plant systems or design parameters that 
could contribute to the initiation of any accidents previously 
evaluated.
    Regarding the probability and consequences of design basis and 
beyond design basis accidents, a comprehensive technical evaluation 
was completed in accordance with Regulatory Guide (RG) 1.174, ``An 
Approach for Using Probabilistic Risk Assessment In Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis'' and RG 
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision 
Making: Technical Specifications.'' This evaluation determined that 
the proposed change is technically justified and the associated risk 
is insignificant.
    The proposed change permits alteration of the containment 
boundary for the IFTS penetration. Regarding the consequences of 
accidents, the proposed change has been determined via a 
probabilistic risk assessment to be acceptable regarding its overall 
impact to the plant's risk, consistent with the Nuclear Regulatory 
Commission's Safety Goal Policy Statement. The resulting pressures 
and temperatures from a design basis Loss Of Coolant Accident (LOCA) 
are considered the primary challenge to the integrity of the 
containment. Pursuant to Amendment 100, the existing Technical 
Specifications require maintaining an adequate water seal to prevent 
leakage from the bottom of the IFTS transfer tube and isolating the 
drain piping. This water seal is adequate to mitigate the effects of 
the design basis peak post-accident pressures and temperatures. The 
proposed change requires the installation of the upper IFTS pool 
gate to provide protection of the Suppression Pool Make Up system 
water inventory. A time limit for IFTS blind flange removal of 60 
days per cycle and a 20 hour limit for the unbolted configuration of 
the IFTS flange have been established as conservative measures to 
limit the associated risk to the containment boundary for all 
accident conditions. The proposed change has been found to be 
acceptable regarding flooding and seismic design issues.
    Therefore, the function of the containment to provide an 
adequate boundary in the event of a design basis LOCA is not 
compromised with the proposed change and the proposed change does 
not result in a significant increase in the probability of the 
consequences of previously evaluated accidents.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed change consists of the removal of the IFTS blind 
flange when in MODES 1, 2, or 3. The IFTS blind flange is a passive 
component that is not part of the primary reactor coolant pressure 
boundary and is not involved in the operation or shutdown of the 
reactor. Being passive, its presence or absence does not affect any 
of the parameters or conditions that could contribute to the 
initiation of any incidents or accidents that are created from a 
loss of coolant or positive reactivity incident. Re-aligning the 
boundary of the primary containment to include portions of the IFTS 
is passive in nature and therefore has no influence on the 
possibility of creating a new or different kind of accident. 
Furthermore, operation of the IFTS is unrelated to the operation of 
the reactor and there is no mishap in the process that can lead or 
contribute to the possibility of losing any coolant in the reactor 
or introducing the chance for positive or negative reactivity or 
other accidents different from and not bounded by those previously 
evaluated.
    Therefore, the proposed change does not result in creating the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change involves the re-alignment of the primary 
containment boundary by removing the IFTS blind flange, which is a 
passive component. The margin of safety that has the potential of 
being impacted by the proposed change involves the dose consequences 
of postulated accidents, which are directly related to potential 
leakage through the primary containment boundary. The potential 
leakage pathways due to the proposed change have been reviewed, and 
leakage can only occur from the administratively controlled IFTS 
transfer tube drain piping. Pursuant to Amendment 100, an individual 
is currently designated to provide timely isolation of this drain 
piping when this proposed change is in effect. The conservatively 
calculated dose, which might be received by the designated 
individual while isolating the drain piping, is well within the 
guidelines of General Design Criterion 19. Furthermore, the drain 
piping isolation valve is included in the Primary Containment 
Leakage Rate Testing Program to ensure that leakage from the piping 
and components located outboard of the blind flange will be 
maintained consistent with the leakage rate assumptions of the 
accident analysis. It has been determined that the proposed change 
would not have a substantial impact on the ultimate pressure 
capacity of the containment as it relates to the Large Early Release 
Frequency (LERF) nor would it have a substantial impact on LERF from 
seismic events. Therefore, the dose consequences of an event would 
be unchanged, and the associated margin of safety would also be 
unchanged.
    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

[[Page 5676]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: March 14, 2002.
    Description of amendment request: The amendment request proposes a 
one-time exception to the requirement in Nuclear Energy Institute (NEI) 
94-01 to perform an integrated leak rate test (ILRT) at a frequency of 
10 years. The exception is to allow ILRT testing within 15 years from 
the last ILRT, completed July 1, 1994. The proposed amendment is 
considered risk-informed, therefore Regulatory Guide 1.174, ``An 
approach for Using Probabilistic Risk Assessment in Risk-Informed 
Decisions on Plant-Specific Changes to the Licensing Basis,'' has been 
followed, while using the methodology of Electric Power Research 
Institute (EPRI) report, ``Risk Impact Assessment of Revised 
Containment Leak Rate Testing Intervals,'' (EPRI TR-104285).
    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. This proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurence, nor is it an 
activity or modification that could lead to equipment failure or 
accident initiation.
    The proposed extension to Type A testing does not result in a 
significant increase in the consequences of an accident as 
documented in NUREG-1493. The NUREG notes that very few potential 
containment leakage paths are not identified by Type B and C tests. 
It concludes that reducing Type A (ILRT) testing frequency to once 
per twenty years leads to an imperceptible increase in risk.
    Other testing and inspections provide a high degree of assurance 
that the containment will not degrade in a manner detectable only by 
Type A testing. The last three Type A tests performed at PPNP 
identified containment leakage within the acceptable criteria, 
indicating a very leak-tight containment. Inspections required by 
the ASME Code are performed in order to identify indications of 
containment degradation that could affect leak-tightness. 
Containment pressure is monitored each shift during plant operation 
and would identify containment vessel shell leakage into the annulus 
by a decrease in containment pressure. Type B and C testing, 
required by Technical Specifications, identifies any containment 
leakage from designed penetrations, such as from valves, that would 
otherwise be detected by a Type A test. These factors establish that 
an extension to the PPNP Type A test interval will not represent a 
significant increase in the consequences of an accident.
    Thus, the proposed amendment does not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    2. This proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revision to the Technical Specifications adds a 
one-time extension to the current interval for Type A testing for 
PPNP. The current test interval of ten years, based on past 
performance, would be extended on a one-time basis to fifteen years 
from the last Type A test. The proposed extension to Type A testing 
does not create the possibility of a new or different type of 
accident since there are no physical changes to the plant or changes 
to the operation of the plant that could introduce a new failure.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. This proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed revision to the PPNP Technical Specifications adds 
a one-time extension to the current interval for Type A testing. The 
current test interval of ten years, based on past performance, would 
be extended on a one-time basis to fifteen years from the last Type 
A test. The proposed extension to Type A testing will not 
significantly reduce the margin of safety. The NUREG-1493 generic 
study of the effects of extending containment leakage testing found 
that a 20-year interval in Type A testing resulted in an 
imperceptible increase in risk to the public. NUREG-1493 found that, 
generically, the design containment leakage rate contributes only 
about 0.1 percent of the overall risk and that decreasing the Type A 
testing frequency would have a minimal effect on this risk since 95% 
of the Type A detectable leakage paths would already be detected by 
Type B and C testing. Furthermore, for PPNP, monitoring containment 
vessel pressure each shift during operation further reduces the risk 
of any containment leakage path going undetected. The PPNP test and 
inspection performance has satisfactorily demonstrated that the 
containment remains very leak tight. The proposed change has no 
effect on Core Damage Frequency (CDF). The change in Large Early 
Release Frequency (LERF) was computed and found to be a ``very 
small'' change in accordance with the guidelines of Regulatory Guide 
1.174. The computed change in Conditional Containment Failure 
Probability (CCFP) and offsite dose have also been evaluated and are 
considered to be insignificant.
    Therefore, the change does not involve a significant reduction 
in a margin of safety.
    Based on the above considerations, it is concluded that a 
significant hazard would not be introduced as a result of this 
proposed change.

    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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 11, 2002.
    Description of amendment request: The proposed amendment would 
revise Crystal River Unit 3 Improved Technical Specifications (ITS) 
3.3.15 ``Reactor Building Purge Isolation-High Radiation;'' ITS Bases 
3.7.15 ``Spent Fuel Assembly Storage;'' ITS 3.9.3 ``Containment 
Penetrations;'' and ITS 3.9.6 ``Refueling Canal Water Level'' to 
account for handling irradiated fuel within containment that has not 
occupied part of a critical reactor core within the previous 72 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Crystal River Unit 3 (CR-3) proposes to revise Improved 
Technical Specifications (ITS) 3.3.15, 3.9.3, 3.9.6, and Bases 
3.7.15.
    Florida Power Corporation (FPC) has determined that this license 
amendment request does not involve a significant hazards 
consideration as defined in 10 CFR 50.92 based on the following:
    (1) Does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change does not increase the probability of a fuel 
handling accident in that the proposed change deals with the results 
of such an accident, not the cause of such an accident. The proposed 
change does not increase the consequences of an accident previously 
evaluated in that the CR-3 Alternate Source Term (AST) has been

[[Page 5677]]

approved by the NRC, and this proposed change implements that NRC 
approval. The AST for the Fuel Handling Accident (FHA) takes no 
credit for containment isolation nor for a filtered release.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes to the ITS do not affect nor create a 
different type of fuel handling accident. The fuel handling accident 
analyses assume that all of the iodine and noble gases that become 
airborne, escape, and reach the exclusion area boundary and low 
population zone with no credit taken for filtration, containment of 
the source term, or for decay or deposition in the containment. The 
proposed changes do not involve the addition or modification of 
equipment nor do they alter the design of plant systems. The revised 
operations are consistent with the fuel handling accident analyses. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Does not involve a significant reduction in margin of 
safety.
    The calculated doses to both the public and control room 
operators are well within the limits given in 10 CFR 50.67. The 
proposed changes do not alter the bases for assurance that safety-
related activities are performed correctly or the basis for any ITS 
that is related to the establishment of or maintenance of a safety 
margin.
    The systems that have been included in the proposed change will 
have administrative controls in place to assure that the systems are 
available and can be promptly returned to operation to further 
reduce dose consequences. These administrative controls will include 
a single normal or contingency method to promptly close the 
equipment hatch opening. This prompt method need not completely 
block the hatch opening nor be capable of resisting pressure, but is 
to enable the ventilation systems to draw the release from the 
postulated FHA in the proper direction such that it can be 
monitored. Therefore, operations of the facility in accordance with 
the proposed amendment would not involve a significant reduction in 
margin of safety.
    Based on the above, FPC concludes that the proposed license 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and accordingly, a finding 
of ``no significant hazards consideration'' is justified.

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

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

    Date of amendment request: December 19, 2002.
    Description of amendment request: The proposed amendment would 
revise Crystal River Unit 3 Improved Technical Specification 2.1.1, 
``Reactor Core Safety Limits.'' The proposed change will permit the use 
of the BHTP correlation, which is needed to utilize the Framatome ANP 
high thermal performance (HTP) spacer grid design.
    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:

    FPC [Florida Power Corporation] has evaluated the proposed 
License Amendment Request (LAR), which consists of the identified 
Improved Technical Specification (ITS) change, against the criteria 
of 10 CFR 50.92(c). The ITS change allows the use of the BHTP 
Correlation for departure from nucleate boiling (DNB) calculations 
of reload cores containing the Mark-B/HTP fuel design.
    FPC has concluded that this proposed LAR does not involve a 
significant hazards consideration. The following is a discussion of 
how each of the criteria is satisfied.
    (1) [Does not] [i]nvolve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed safety limit value ensures that fuel integrity will 
be maintained during normal operations and anticipated operational 
occurrences (AOOs), and that the design requirements will continue 
to be met. The proposed methodology for the BHTP departure from 
nucleate boiling (DNB) correlation will be generically reviewed and 
approved by the NRC prior to its use by Crystal River Unit 3 (CR-3) 
in mixed core reload analyses. The core operating limits will be 
developed in accordance with the new methodology and any limitations 
established by the NRC in its safety evaluation of the new 
methodology. The proposed safety limit value does not affect the 
performance of any equipment used to mitigate the consequences of an 
analyzed accident. There is no impact on the source term or pathways 
assumed in accidents previously evaluated. No analysis assumptions 
are violated and there are no adverse effects on the factors that 
contribute to offsite or onsite dose as the result of an accident. 
Therefore, the safety limit value for the BHTP correlation will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) [Does not] [c]reate the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed safety limit value does not change the methods 
governing normal plant operation, nor are the methods utilized to 
respond to plant transients altered. The BHTP correlation is not an 
accident/event initiator. No new initiating events or transients 
result from the use of the BHTP correlation and the related safety 
limit changes. Therefore, the safety limit value for the BHTP 
correlation will not involve the possibility of a new or different 
kind of accident from any previously evaluated.
    (3) [Does not] [i]nvolve a significant reduction in a margin of 
safety.
    The proposed safety limit value has been established in 
accordance with the methodology for the BHTP correlation, to ensure 
that the applicable margin of safety is maintained (i.e., there is 
at least 95% probability at a 95% confidence level that the hot fuel 
rod in the core does not experience departure from nucleate boiling 
(DNB)). The proposed methodology for the BHTP DNB correlation will 
be generically reviewed and approved by the NRC prior to its use by 
CR-3. The other reactor core safety limits will continue to be met 
by analyzing the reload for the mixed core using NRC approved 
methods, and incorporation of resultant operating limits into the 
Core Operating Limits Report (COLR). Therefore, the safety limit 
value for the BHTP correlation will not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 20, 2002.
    Description of amendment request: This proposed amendment provides 
editorial and administrative changes to the Technical Specifications. 
The changes correct typographical, spelling, numbering syntax, page 
break, and font consistency errors as well as removing blank pages and 
associated references. There are no substantive changes made in the 
proposed amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change

[[Page 5678]]

involve a significant increase in the probability or consequences of 
an accident previously evaluated?
    No. The proposed amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because the proposed amendments are purely 
administrative or editorial in nature. These amendments make no 
substantive Technical Specification changes and do not affect any 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant; and they do not affect Technical 
Specifications that preserve safety analysis assumptions. Therefore, 
the proposed changes do not affect the probability or consequences 
of accidents previously analyzed.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    No. The use of the administratively changed Technical 
Specifications does not create the possibility of a new or different 
kind of accident from any previously evaluated, since the proposed 
amendments will not change the physical plant or the modes of plant 
operation defined in the facility operating license. No new failure 
mode is introduced due to the administrative changes and 
clarifications, since the proposed changes do not involve the 
addition or modification of equipment, nor do they alter the design 
or operation of affected plant systems, structures, or components.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    No. The operating limits and functional capabilities of the 
affected systems, structures, and components are unchanged by the 
proposed amendments. The changed Technical Specifications, which 
correct administrative and editorial errors, and clarify existing 
Technical Specification requirements, do not reduce any of the 
margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: December 31, 2002.
    Description of amendment request: The proposed amendment would 
change the reactor vessel material surveillance program to incorporate 
the Boiling Water Reactor Vessel and Internals Project (BWRVIP) 
Integrated Surveillance Program (ISP) into the licensing basis.
    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.
    Pressure-temperature (P/T) limits (CNS [Cooper Nuclear Station] 
Technical Specifications Figures 3.4.9-1, 2, and 3) are imposed on 
the reactor coolant system to ensure that adequate safety margins 
against non-ductile or brittle fracture exist during normal 
operation, anticipated operational occurrences, and system 
hydrostatic tests. The P/T limits are based on the nil-ductility 
reference temperature, RTNDT, as described in ASME 
Section XI, Appendix G. Changes in the fracture toughness properties 
of RPV [reactor pressure vessel] beltline materials, resulting from 
the neutron irradiation and the thermal environment, are monitored 
by a surveillance program in compliance with the requirements of 10 
CFR 50 [title 10 of the Code of Federal Regulations part 50] 
Appendix H. The effect of neutron fluence on the shift in the 
RTNDT of RPV materials is predicted by methods given in 
RG [Regulatory Guide] 1.99, Revision 2.
    This change is not related to any accidents previously 
evaluated. Rather, the reactor vessel surveillance program, 
corresponding material evaluations, and adjustment of a plant's P/T 
limits, as necessary, protect against the possibility of reactor 
vessel brittle fracture. Monitoring, evaluation, and adjustment of 
CNS P/T limits to ensure adequate margin exists to brittle fracture 
will continue. This change only replaces a plant-specific monitoring 
and evaluation program with an integrated industry program, the 
BWRVIP ISP. The NRC has reviewed this program and approved it for 
implementation in a Safety Evaluation, dated February 1, 2002.
    CNS's current P/T limits were established based on adjusted 
reference temperatures developed in accordance with the procedures 
described in RG 1.99, Revision 2. Calculation of adjusted reference 
temperature by these procedures includes a margin term to ensure 
conservative, upper-bound values are used for the calculation of the 
P/T limits. This change does not affect the existing P/T limits in 
the CNS Technical Specifications Figures 3.4.9-1, 2, and 3. This 
change will not affect any plant safety limits or limiting 
conditions of operation. The proposed change will not affect reactor 
pressure vessel performance as no physical changes are involved 
aside from changes related to surveillance capsule withdrawal, and 
CNS vessel P/T limits will remain conservative in accordance with RG 
1.99, Revision 2 criteria. The proposed change will not cause the 
reactor pressure vessel or interfacing systems to be operated 
outside of their design or testing limits. Also, the proposed change 
will not alter any assumptions previously made in evaluating the 
radiological consequences of accidents. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the CNS license basis to reflect 
participation in the BWRVIP ISP. Participation in the BWRVIP ISP 
will continue to ensure that the CNS reactor vessel materials are 
monitored and evaluated as necessary to protect against brittle 
fracture. This proposed change does not involve a modification of 
the design of plant structures, systems, or components. The proposed 
change will not impact the manner in which the plant is operated as 
plant operating and testing procedures will not be affected by the 
change. The proposed change will not degrade the reliability of 
structures, systems, or components important to safety as equipment 
protection features will not be deleted or modified, equipment 
redundancy or independence will not be reduced, supporting system 
performance will not be downgraded, the frequency of operation of 
equipment will not be increased, and increased or more severe 
testing of equipment will not be imposed. No new accident types or 
failure modes will be introduced as a result of the proposed change. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from that previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Conformance with 10 CFR [part] 50 Appendix G defines the 
accepted safety margin for Reactor Coolant Pressure Boundary 
fracture toughness. The P/T limits are not derived from Design Basis 
Accident (DBA) analyses. They are prescribed during normal operation 
to avoid encountering pressure, temperature, and temperature rate of 
change conditions that might cause undetected flaws to propagate and 
cause nonductile failure of the reactor pressure vessel, a condition 
that is unanalyzed. Since the P/T limits are not derived from any 
DBA, there are no acceptance limits related to the P/T limits. 
Rather the P/T limits are acceptance limits themselves since they 
preclude operation in an unanalyzed condition.
    This proposed change will not alter the required margins as 
defined in 10 CFR [part] 50, Appendix G. This proposed change will 
not affect any safety limits, limiting safety system settings, or 
limiting conditions of operation. The proposed change does not 
represent a change in initial conditions, or in a system response 
time, or in any other parameter affecting the course of an accident 
analysis supporting the Bases of any Technical Specification. The 
proposed

[[Page 5679]]

change does not involve revision of the P/T limits. Rather, this 
change involves a revision to the surveillance capsule withdrawal 
schedule, a revision to the reactor vessel fluence calculational 
methodology to achieve consistency within the BWRVIP ISP, and 
participation in future BWRVIP ISP developments. The current P/T 
limits were established based on adjusted reference temperatures for 
vessel beltline materials calculated in accordance with RG 1.99, 
Revision 2 which will continue to conform to 10 CFR [part] 50 
Appendix G. Therefore, the proposed change does not involve a 
significant reduction in any safety margins.
    In summary, it is concluded that this License Amendment Request 
does not involve significant hazards consideration results. NPPD has 
researched the existing regulatory precedent and has identified five 
BWR licensees with similar License Amendment Requests currently 
under NRC staff review:

[sbull] Browns Ferry Units 2 and 3--Submittal date November 6, 2002.
[sbull] Monticello Generating Station--Submittal date September 19, 
2002.
[sbull] River Bend--Submittal date August 15, 2002.
[sbull] Fermi Unit 2--Submittal date August 8, 2002.
[sbull] Susquehanna Units 1 and 2--Submittal date July 25, 2002.

    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.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: January 13, 2003.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) operating license and 
Technical Specifications to increase the licensed rated power by 1.4 
percent to 1673 megawatts thermal (MWt) using measurement uncertainty 
recapture.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the Kewaunee Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    There are no changes as a result of the measurement uncertainty 
recapture (MUR) power uprate to the design or operation of the plant 
that could affect system, component, or accident mitigative 
functions. All systems and components will function as designed and 
the applicable performance requirements have been evaluated and 
found to be acceptable.
    The reduction in power measurement uncertainty allows for some 
of the safety analyses to continue to be used without modification. 
This is because the safety analyses were performed or evaluated at 
either 102 percent of 1650 MWt or higher. Analyses at these power 
levels support a core power level of 1673 MWt with a measurement 
uncertainty of 0.6 percent. Radiological consequences of USAR 
[updated safety analysis report] chapter 14 accidents were assessed 
previously using the alternate source term (AST) methodology 
(reference 7.1, TAC [technical assignment control] No. MB4596). 
These analyses were performed at 102 percent of 1650 MWt and 
continue to be bounding. The USAR chapter 14 analyses and accident 
analyses submitted to the NRC [Nuclear Regulatory Commission] with 
the fuel transition (reference 7.3, TAC No. MB5718) continue to 
demonstrate compliance with the relevant accident analyses 
acceptance criteria. Therefore, there is no significant increase in 
the consequences of any accident previously evaluated.
    The primary loop components (reactor vessel, reactor internals, 
control rod drive mechanisms, loop piping and supports, reactor 
coolant pumps, steam generators, and pressurizer) were evaluated at 
an uprated core power level of 1772 MWt and continue to comply with 
their applicable structural limits. These analyses also demonstrate 
the components will continue to perform their intended design 
functions. Changing the applicability of the heatup and cooldown 
curves is based on uprated fluence values. This does not have a 
significant effect on the reactor vessel integrity. Thus, there is 
no significant increase in the probability of a structural failure 
of the primary loop components.
    All of the NSSS [Nuclear Steam Supply System] systems will 
continue to perform their intended design functions during normal 
and accident conditions. The auxiliary systems and components 
continue to comply with the applicable structural limits and will 
continue to perform their intended functions. The NSSS/BOP [balance 
of plant] interface systems were evaluated at 1772 MWt and will 
continue to perform their intended design functions. Plant 
electrical equipment was also evaluated and will continue to perform 
their intended functions. Therefore, there is no significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the Kewaunee Nuclear Power Plant in accordance 
with the proposed amendments does not result in a new or different 
kind of accident from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. All 
systems, structures, and components previously required for the 
mitigation of an event remain capable of fulfilling their intended 
design function at the uprated power level. The proposed change has 
no adverse effects on any safety-related systems or component and 
does not challenge the performance or integrity of any safety-
related system. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the Kewaunee Nuclear Power Plant in accordance 
with the proposed amendments does not result in a significant 
reduction in a margin of safety.
    Operation at the 1673 MWt core power does not involve a 
significant reduction in the margin of safety. The current accident 
analyses have been previously performed with a two percent power 
measurement uncertainty or at uprated core powers that exceed the 
MUR uprated core power. System and component analyses have been 
completed at a core power in excess of the MUR uprated core power. 
Analyses of the primary fission product barriers at uprated core 
powers have concluded that all relevant design basis criteria remain 
satisfied in regard to integrity and compliance with the regulatory 
acceptance criteria. As appropriate, all evaluations have been 
either reviewed and approved by the NRC, are in the process of being 
approved by the NRC, or are in compliance with applicable regulatory 
review guidance and standards. 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: John H. O'Neill, Jr., Esq., Shaw Pittman, 
Potts & Trowbridge, 2300 N. Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: September 12, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.5.2, ``ECCS [Emergency Core 
Cooling System]--Operating,'' and TS 3.5.3, ``ECCS-Shutdown,'' to add a 
surveillance requirement to verify every 31 days that the ECCS piping 
is full of water; consistent with NUREG-1431, Standard Technical 
Specifications, Westinghouse Plants, Revision 2.

[[Page 5680]]

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

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    This license amendment request proposes to add a surveillance 
requirement to verify the ECCS is full of water every 31 days while 
operating in Modes 1, 2, 3 and 4.
    This proposed change does not cause an increase in the 
probabilities of any accidents previously evaluated, because the 
change will not cause an increase in the probability of any 
initiating events for accidents previously evaluated. In particular, 
the change affects the ECCS, which serves to mitigate rather than 
initiate accidents.
    The consequences of the accidents previously evaluated in the 
PBNP [Point Beach Nuclear Plant] Final Safety Analysis Report (FSAR) 
are determined by the results of analyses that are based on initial 
conditions of the plant, the type of accident, transient response of 
the plant, and the operation and failure of equipment and systems. 
The change proposed in this license amendment request provides an 
appropriate surveillance requirement for the ECCS, and thus does not 
increase the probability of failure of this equipment or its ability 
to operate as required for the accidents previously evaluated in the 
PBNP FSAR.
    Therefore, the consequences of an accident previously evaluated 
in the PBNP FSAR will not be significantly increased as a result of 
the proposed change, because the factors that are used to determine 
the consequences of accidents are not being changed.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    Equipment important to safety will continue to operate as 
designed. The proposed change does not result in any event 
previously deemed incredible being made credible. The change does 
not result in more adverse conditions or result in any increase in 
the challenges to safety systems. Therefore, operation of the Point 
Beach Nuclear Plant in accordance with the proposed amendment will 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed amendment will not otherwise affect the plant protective 
boundaries and will not cause a release of fission products to the 
public. Venting the piping associated with a train of ECCS will 
render that ECCS train inoperable while it is being vented. 
Performance of this surveillance will therefore affect the 
availability of the associated ECCS train, but performance of the 
surveillance requirement at the specified frequency is consistent 
with the requirements of NUREG-1431, Standard Technical 
Specifications for Westinghouse Plants, Revision 2. Additionally, 
verifying the ECCS piping is full of water ensures that the system 
will perform properly, injecting its full capacity into the RCS 
[reactor coolant system], upon demand. Therefore, adopting a 
surveillance requirement to verify the ECCS piping is full of water, 
will not result in more than a minimal reduction in the margin of 
safety.

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

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

    Date of amendment request: September 26, 2002.
    Description of amendment request: The proposed change would revise 
the Steam Generator low-low level trip setpoint and allowable values 
provided in the Salem Nuclear Generating Station, Unit Nos. 1 and 2, 
Technical Specifications Table 2.2-1, ``Reactor Trip System 
Instrumentation Trip Setpoints,'' and Table 3.3-4, ``Engineered Safety 
Feature Actuation System Instrumentation Trip Setpoints.'' The changes 
are necessary based on PSEG Nuclear's evaluation of a loss of feedwater 
transient at Diablo Canyon. During the event, Diablo Canyon personnel 
observed a flow induced pressure drop in the steam generator mid-deck 
area. The proposed change accounts for a level measurement bias 
resulting from the pressure drop that was not considered in the 
previous Westinghouse analysis. This bias has the effect of providing 
nonconservative level readings and setpoints.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to Tables 2.2-1 and 3.3-4 changes both the 
allowable trip setpoint and allowable value for the Steam Generator 
Water Level-Low-Low from =9.0% to =14.0% and 
from =8.0% to =13% respectively. The Steam 
Generator Water Level Low-Low trip provides core protection by 
preventing operation with the steam generator water level below the 
minimum volume required for adequate heat removal capacity. The 
signal is used as a primary protection signal for the design basis 
loss of normal feedwater, loss of offsite power and feedwater line 
break safety analysis. The specified setpoint provides allowance 
that there will be sufficient water inventory in the steam 
generators at the time of trip to allow for starting delays of the 
auxiliary feedwater system. The change in the setpoint and allowable 
value allows the trip to function as originally designed accounting 
for the differential pressure created by steam flow past the mid-
deck plate in the moisture separator section of the steam generator.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Steam Generator Water Level-Low-Low 
trip setpoint and allowable values allow the trip to function as 
originally designed. They do not alter the plant configuration in 
any way, and do not replace or modify existing plant equipment, or 
affect any plant operations. No additional failure mechanisms are 
introduced as a result of the changes to the setpoints and allowable 
values.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment would not involve a significant 
reduction in the margin of safety.
    The proposed changes to the allowable trip setpoint and 
allowable value for the Steam Generator Water Level-Low-Low trip 
maintains core protection by preventing operation with the steam 
generator water level below the minimum volume required for adequate 
heat removal capacity.
    Therefore, it is concluded that the proposed changes to the 
steam generator low low level trip setpoint and allowable value[s] 
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

[[Page 5681]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 23, 2002.
    Description of amendment request: The proposed change would revise 
the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, 
Technical Specification (TS) 6.12, ``High Radiation Area'' to be 
consistent with the Standard TSs for Westinghouse Plants (NUREG-1431, 
Revision 2) by updating the current reference to title 10 of the Code 
of Federal Regulations (10 CFR), section 20.203 with the corresponding 
reference to 10 CFR 20.1601.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not affect accident initiators or 
precursors and do not alter the design assumptions, conditions, 
configuration of the facility, or manner in which the plant is 
operated. The proposed changes do not alter or prevent the ability 
of structures, systems, or components to perform their intended 
safety function to mitigate the consequences of an initiating event 
within the acceptance limits assumed in the UFSAR. The proposed 
changes are administrative in nature. Technical Specification (TS) 
6.12 will be updated to include the new 10 CFR 20 (effective 06/20/
91) requirements. The proposed changes do not alter the conditions 
or assumptions in any of the previous accident analyses, and as a 
result, the radiological consequences associated with these analyses 
remain unchanged.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated.
    The proposed changes are administrative in nature and the 
relocated procedural details do not change the level of programmatic 
controls and procedural details. Accordingly, the proposed changes 
do not create any new failure modes or limiting single failures 
associated with a plant structure, system, or component important to 
safety. Also, there will be no change in the types or increase in 
the amounts of any effluents released offsite.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment would not involve a significant 
reduction in the margin of safety.
    The proposed changes do not impact equipment design or 
operation, nor do the changes affect any TS safety limits or safety 
system settings that could adversely affect plant safety. The 
proposed changes are administrative in nature. Technical 
Specification (TS) 6.12 will be updated to include the new 10CFR20 
requirements (effective 06/20/91) and are in conformance with NUREG-
1431. Furthermore, the proposed changes do not result in a change in 
the types or an increase in the amounts of any effluents released 
offsite.
    Therefore, it is concluded that the proposed changes do not 
involve a significant reduction in a margin of safety.

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

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, 
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: October 4, 2002 (TS 02-07).
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.8.4.h, ``Containment Leakage Rate 
Testing Program,'' to allow a one-time, 5-year extension to the current 
10-year test interval for the performance-based leakage rate test 
program for 10 CFR 50, Appendix J, Type A tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change for extending Type A test frequency does not 
significantly increase the probability of an accident previously 
evaluated since the change is not a modification to plant systems, 
nor a change to plant operation that could initiate an accident. TVA 
performed an evaluation of the risk significance for the proposed 
increase to the SQN Units 1 and 2 Type A test frequency. The results 
of the TVA risk evaluation indicates that the increase in Large 
Early Release Frequency (LERF) remains below the level of risk 
significance defined in NRC Regulatory Guide (RG) 1.174, ``An 
Approach for Using Risk Assessment In Risk-Informed Decisions On 
Plant-Specific Changes to the Licensing Basis.'' TVA's evaluation 
indicates that the increase in frequency for all releases (small, 
large, early and late) and the increase in radiation dose to the 
population is also non-risk significant. The proposed test interval 
extension does not involve a significant increase in the 
consequences of an accident. Research documented in NUREG-1493 
determined that generically, very few potential containment leakage 
paths fail to be identified by Type A tests. An analysis of 144 Type 
A test results, including 23 failures, found that no failures were 
due to containment liner breach. The NUREG concluded that reducing 
the Type A test frequency to once per 20 years would lead to an 
imperceptible increase in risk. Furthermore, the NUREG concluded 
that Type B and C testing provides assurance that containment 
leakage from penetration leak paths (i.e., valves, flanges, 
containment air-locks) identify any leakage that would otherwise be 
detected by the Type A tests. In addition to the NUREG conclusions, 
TVA's American Society of Mechanical Engineers (ASME) IWE program 
performs containment inspections in order to detect evidence of 
degradation that may affect either the containment structural 
integrity or leak tightness. In addition to the IWE examinations, 
TVA will perform additional nondestructive examinations of the steel 
containment vessel in the ice condenser region (inaccessible areas) 
at various elevations. These additional non-destructive examinations 
will provide added assurance of containment integrity during the 5-
year extended interval. Accordingly, TVA's proposed extension of the 
Type A test interval does not increase 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 to extend the Type A test interval does not 
create the possibility of a new or different type of accident 
because there are no physical changes made to the plant or plant 
equipment governing normal plant operation. There are no changes to 
the operation of the plant that would introduce a new failure mode 
creating the possibility of a new or different kind of accident. TVA 
will perform additional non-destructive examinations of the steel 
containment vessel in the ice condenser region (inaccessible areas) 
at various elevations. These additional non-destructive examinations 
will provide added assurance of containment integrity during the 5 
year extended interval.

[[Page 5682]]

    3. Does the proposed change not involve a significant reduction 
in a margin of safety?
    The proposed change to extend the Type A test interval will not 
significantly reduce the margin of safety. A generic study 
documented in NUREG-1493 indicates that extending the Type A leak 
test interval to 20 years would result in an imperceptible increase 
in risk to the public. The NUREG also found that, generically, the 
containment leakage rate contributes a very small amount to the 
individual risk and that the decrease in the Type A test frequency 
would have a minimal affect on risk because most potential leakage 
paths are detected by Type C testing. Previous Type A leakage tests 
conducted on SQN Units 1 and 2 indicate that leakage from 
containment have been less than the 10 CFR 50, Appendix J leakage 
limit of 1.0 La. A review of the previous Type A test 
results indicate a stable trend with a 10 percent margin below the 
1.0 La leakage limit. Accordingly, these test results, in 
conjunction with the research findings from NUREG-1493, provide 
assurance that the proposed extension to the Type A test interval 
would not significantly reduce the margin of safety. Based on the 
above, TVA concludes that the proposed amendment presents no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: November 15, 2002 (TS 02-06).
    Brief description of amendments: The proposed amendments would 
revise the Technical Specification (TS) 3.7.1.3, ``Condensate Storage 
Water,'' Limiting Condition for Operation by increasing the required 
minimum amount of stored water from 190,000 gallons to 240,000 gallons. 
This change is being made to support the replacement steam generator 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), 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 does not change the physical design and 
construction of the condensate storage tank (CST). The purpose of 
the increased water volume is to ensure that the required volume of 
water, preserved by the technical specification (TS), is sufficient 
to meet Sequoyah Nuclear Plant (SQN) Licensing and Design Basis 
after installation of the replacement steam generators. The change 
in the administratively controlled inventory of the CST will not 
increase the probability of an accident. Therefore, the proposed 
change does not involve a significant increase in the probability of 
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?
    This change increases the minimum required volume of water in 
the CST, thus ensuring that the auxiliary feedwater (AFW) system can 
perform its required safety function, using a preferred water source 
for plant transient mitigation. The maximum and normal water levels 
in the CST are not being changed. Additionally, increasing the 
minimum water volume requirement will not initiate any accident. 
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?
    This change does not reduce any margin associated with the CST 
inventory available to AFW. The requirement for sufficient CST 
volume to maintain hot standby and subsequent cooldown to hot 
shutdown continues to be met by the minimum volume increase. 
Additionally, the essential raw cooling water (ERCW) system still 
provides the long-term supply of safety grade cooling water to the 
AFW in the event that all inventory of the CST is lost. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, 
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: November 15, 2002 (TS 02-01).
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to revise the trip setpoint 
column of the Reactor Protection System and Engineered Safety Features 
(ESF) instrumentation tables to utilize a nominal setpoint value and 
revise the associated Bases discussions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revisions for the nominal trip setpoint 
representation are administrative changes that will not impact the 
application of the reactor trip or ESF actuation system 
instrumentation requirements. This is based on the setpoint 
requirements being applied without change, as well as the Avs 
[allowable values], in accordance with the setpoint methodology. The 
removal of the inequalities associated with the trip setpoint values 
will be more appropriate for the use of nominal setpoint values but 
will not differ in application from the setpoint methodology 
utilized by TVA. The revision of the radiation monitoring 
instrumentation table to use an Av will continue to provide 
appropriate operability limits. Deletion of the nominal terminology 
associated with overtemperature delta temperature average 
temperature at rated thermal power (T') and reactor coolant system 
power operated relief valve (PORV) lift settings provides a better 
representation of the limits associated with these values. In 
addition, this change will not alter plant equipment or operating 
practices. Therefore, the implementation of these changes will not 
increase the probability or consequences of an accident.
    The revision of the reactor coolant pump (RCP) underfrequency 
trip setpoint and the Avs for the RCP underfrequeny and undervoltage 
and the containment purge radiation high has been evaluated and the 
results are documented in approved calculations. These calculations 
verify that the revised values are acceptable in accordance with 
appropriate calculation methodologies and that they will continue to 
support the accident analysis. This is based on margin being 
available in the accuracy determinations that could be used without 
impacting the intended functions of this instrumentation and 
maintains the established safety limits. These revisions will not 
require changes to the instrumentation settings currently being used 
or the methods for maintaining them. The offsite dose potential will 
not be impacted because this instrumentation will continue to 
adequately provide the designed safety functions to limit the 
release of radioactivity. Therefore, the proposed revision of these 
values will not significantly increase the probability or 
consequences of an accident.
    The relocation and enhancement of current radiation monitoring 
and loss of voltage

[[Page 5683]]

functions to new LCOs [limiting condition for operations] does not 
alter the intended functions of these systems or physically alter 
these systems. While some requirements have change[d] from current 
limitations, these changes have provided more appropriate criteria 
to ensure that the accident mitigation functions are maintained 
properly and are available. Changes to Avs have been evaluated in 
accordance with TVA setpoint methodology and have been verified to 
acceptably protect the associated safety limits. Format changes 
provide a clearer representation of the requirements and provide 
more consistency with the standard TSs in NUREG-1431. These changes 
continue to support or improve the required safety functions and 
therefore, will not increase the possibility or consequence of an 
accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The revision of the nominal trip setpoint representation and 
elimination of the nominal nomenclature, as well as the revised 
setpoint value and Avs, and the relocated LCOs will not alter the 
plant configuration or functions. The revised setpoint and the 
proposed operability limits will continue to provide acceptable 
initiation of safety functions for the mitigation of postulated 
accidents as required by the design basis. The primary function of 
the reactor protection system, the ESF actuation system, and the new 
actuation function LCOs is to initiate accident mitigation 
functions. These functions are not considered to be initiators of 
postulated accidents. The PORVs provide accident mitigation 
functions and could be the source of a loss of coolant accident. 
However, a clarification of how to apply the actuation setpoints 
without a change to the setpoints will not impact accident 
generation. The proposed changes do not create the possibility of a 
new or different kind of accident because the design functions are 
not altered and the proposed values meet the accident analysis 
requirements for accident mitigation.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The setpoint and Av revisions proposed in this request were 
evaluated and found to be acceptable based on operating margin 
available in the accuracy determinations. The reassignment of this 
excess margin to the setpoint and Av will not impact the safety 
limits required for the associated functions. The nominal trip 
setpoint representation change and the elimination of inappropriate 
nominal indications does not alter the TS functions or their 
application and will not require changes to design settings. The 
relocated requirements to new LCOs provide appropriate limits and 
enhancements to the actuation functions. Plant systems will continue 
to be actuated for those plant conditions that require the 
initiation of accident mitigation functions. The margin of safety is 
not significantly reduced because the proposed changes to the Av and 
setpoint representations will not change design functions and the 
initiation of accident mitigation functions for appropriate plant 
conditions will not be adversely impacted.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

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 delete 
the monthly analog rod position test for the control rod bottom 
bistables currently required by Technical Specification (TS) Table 4.1-
1, Item 9.
    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 deletes the monthly analog rod position test 
that verifies the operation of the rod bottom bistables. However, 
the TSs still require bistable action to be functionally verified to 
ensure operability on an 18-month frequency as part of the overall 
analog rod position indication system calibration. Furthermore, the 
TS-required monthly rod bottom bistable action test was being 
performed to address instrument drift in the rod bottom setpoint, 
which will essentially be eliminated by the design of new digital-
based IRPI [Individual Rod Position Indication] electronics being 
installed. Consequently, elimination of the monthly rod bottom 
bistable action test will not result in the failure of any plant 
structures, systems, or components and does not have a detrimental 
impact on the integrity of any plant structure, system, or component 
that initiates an analyzed event. The proposed change will not alter 
the operation of or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident. As a 
result, the probability of any accident previously evaluated is not 
significantly increased.
    Consequences of analyzed events are the result of the plant 
being operated within assumed parameters at the onset of any event, 
and the successful functioning of at least one train or division of 
the equipment credited with mitigating the event. These changes do 
not impact the capability of the credited equipment to perform, nor 
is there any change in the likelihood that credited equipment will 
fail to perform. Deletion of the monthly rod bottom bistable action 
test does not affect the ability of the control rods to perform 
their function. Surveillance tests to verify the operability of the 
IRPI System are still being performed. Furthermore, the Rod Position 
Demand Counter System provides redundant control rod position 
indication. As a result, the consequences of any accident previously 
evaluated are not significantly affected by the proposed change.
    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 deletes the monthly surveillance of rod 
bottom bistable action in the Individual Rod Position Indication 
system. This change does not alter the methods governing normal 
plant operation. The IRPI provides indication of rod position, is 
one of two independent systems that are provided to detect a rod 
drop and is the backup to detection by rapid reduction of ex-core 
neutron flux. The dropping of a rod assembly can occur when the rod 
drive mechanism is de-energized from the Rod Control System. This 
accident has been evaluated in the UFSAR and in all cases the DNB 
design bases is met by demonstration that the DNBR is greater than 
the limiting value. Thus, this change deleting the monthly analog 
rod position test does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The digital-based IRPI system continues to meet the design 
function of providing reliable control rod position indication. The 
proposed change and associated replacements with digital-based IRPI 
system electronics provides enhanced testing through the automatic 
self-testing diagnostic features. Consequently, the overall ability 
to detect failures is not degraded. Therefore, the change deleting 
the monthly analog rod position test 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.

[[Page 5684]]

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 amendments request: December 19, 2002.
    Description of amendments request: The proposed changes would 
revise the Technical Specifications (TS) to Facility Operating License 
Nos. DRP-32 and DRP-37 for Surry Power Station, Units 1 and 2, 
respectively, to reflect changes in regulations, correct typographical 
and editorial errors made in previous TS revisions, and to revise TS 
cross-references to Updated Final Safety Analysis Report sections.
    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:

    Dominion has reviewed the requirements of 10 CFR 50.92 as they 
relate to the proposed administrative change to the Surry Power 
Station Units 1 and 2 Technical Specifications (TS) and Bases. The 
proposed change to the Surry TS makes administrative revisions to 
reflect changes in regulations, corrects editorial and typographical 
errors from previous TS revisions, and revises TS cross-references 
to Updated Final Safety Analysis Report (UFSAR) sections. Due to the 
strictly administrative nature of the proposed TS change, we have 
determined that a significant hazards consideration does not exist. 
The basis for this determination is provided as follows:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change is administrative in nature and as such does 
not impact the condition or performance of any plant structure, 
system or component. The proposed administrative change does not 
affect the initiators of any previously analyzed event nor the 
assumed mitigation of accident or transient events. As a result, the 
proposed change to the Surry Technical Specifications does not 
involve any increase in the probability [nor] the consequences of 
any accident or malfunction of equipment important to safety 
previously evaluated since neither accident probabilities or 
consequences are being affected by this proposed administrative 
change.
    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 is administrative in nature, and therefore 
does not involve any changes in station operation or physical 
modifications to the plant. In addition, no changes are being made 
in the methods used to respond to plant transients that have been 
previously analyzed. No changes are being made to plant parameters 
within which the plant is normally operated or in the setpoints, 
which initiate protective or mitigative actions and no new failure 
modes are being introduced. Therefore, the proposed administrative 
change to the Surry Technical Specifications does not create the 
possibility of a new or different kind of accident or malfunction of 
equipment important to safety from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed change is administrative in nature, and does not 
impact station operation or any plant structure, system or component 
that is relied upon for accident mitigation. Furthermore, the margin 
of safety assumed in the plant safety analysis is not affected in 
any way by the administrative ``cleanup'' of the Surry Technical 
Specifications. Therefore, the proposed administrative change to the 
Surry Technical Specifications does not involve a reduction in a 
margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has 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 PDR, located at One White Flint North, Public File 
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

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

    Date of application for amendment: December 19, 2001, as 
supplemented July 30, 2002, and November 14, 2002.
    Brief description of amendment: The amendment includes a revision 
of the Technical Specification (TS) Limiting Conditions for Operation 
3.4, ``Decay Heat Removal Capability,'' conforming changes to TS Table 
3.5-2, ``Accident Monitoring Instruments,'' and TS 4.9.1.2, ``Decay 
Heat Removal--Periodic Testing,'' and numerous editorial changes.
    Date of issuance: January 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 242.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12598).
    The supplements dated July 30, and November 14, 2002, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 5685]]

Safety Evaluation dated January 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 2, 2002.
    Brief description of amendments: These amendments change the 
administrative controls in Technical Specification 5.7, ``High 
Radiation Area.''
    Date of issuance: January 13, 2003.
    Effective date: January 13, 2003.
    Amendment Nos.: 225 and 252.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50950).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 13, 2003.
    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: May 14, 2002, as supplemented by letter 
dated December 17, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification Table 3.3.8.1-1, ``Loss of Power Instrumentation,'' by 
changing the degraded voltage--voltage basis and loss-of-coolant 
accident time delay allowable values to reflect the results of new 
calculations performed in association with a design basis 
reconstitution.
    Date of issuance: January 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
no later than November 30, 2003.
    Amendment No.: 128.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42823).
    The December 17, 2002, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2003.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington Date of application for amendment: October 
22, 2002.

    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to change TS section 5.0, ``Administrative 
Controls,'' and adopt Technical Specification Task Force (TSTF) -258, 
Revision 4. The change revises: (1) Section 5.2.2, ``Unit Staff,'' to 
delete details of staffing requirements and delete requirements for the 
Shift Technical Advisor (STA) as a separate position while retaining 
the function, (2) section 5.5.4, ``Radioactive Effluent Controls 
Program,'' to be consistent with the intent of 10 CFR part 20, (3) 
section 5.6.4, ``Monthly Operating Reports,'' to delete periodic 
reporting requirements for main steam safety/relief valve challenges to 
be consistent with Generic Letter 97-02, ``Revised Contents of the 
Monthly Operating Report,'' and (4) section 5.7, ``High Radiation 
Area,'' in accordance with 10 CFR 20.1601(c). TS section 5.3.2 is added 
to incorporate regulatory definitions for the senior reactor operator 
(SRO) and reactor operator (RO) positions.
    Date of issuance: January 9, 2003.
    Effective date: January 9, 2003, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 182.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75870).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 9, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 28, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) sections 3.7, ``Auxiliary Electrical Systems,'' and 
4.6, ``Emergency Power System Periodic Tests,'' to relocate the 
requirements for the gas turbine generators to the Updated Final Safety 
Analysis Report (UFSAR) and the plans, programs and procedures that 
document and control the credited functions of these systems, 
structures, and components. The amendments also deleted TS 3.7.B.2.b. 
to remove the option that allows power operation for up to 72 hours 
with a gas turbine as the only available 13.8 kilovolt power source.
    Date of issuance: January 17, 2003.
    Effective date: This license amendment is effective as of the date 
of its issuance and shall be implemented within 60 days and only after 
incorporation of the required changes into the UFSAR and completion of 
the necessary implementation and procedural changes.
    Amendment No.: 236.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications and Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: May 14, 2002 (67 FR 
34484).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 5, 2002.
    Brief description of amendment: The amendment relocates Technical 
Specification (TS) 3/4.6.I to the Pilgrim Nuclear Power Station Updated 
Final Safety Analysis Report. The affected TS contains snubber 
operability and surveillance requirements. The associated Bases section 
will also be relocated.
    Date of issuance: January 14, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 195.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68735).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 14, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 27, 2002.
    Brief description of amendments: The amendments revise Appendix B,

[[Page 5686]]

``Environmental Protection Plan (Non-Radiological),'' of the licenses 
to remove a parenthetical reference to a superseded section of 10 CFR 
part 51.
    Date of issuance: January 21, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 211 & 205.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised Appendix B of the licenses.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66009).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 21, 2003.
    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: October 8, 2002.
    Brief description of amendment: The amendment relocates Technical 
Specification 2.13, ``Nuclear Detector Cooling System,'' and its 
associated Bases to the Fort Calhoun Station Updated Safety Analysis 
Report.
    Date of issuance: January 16, 2003.
    Effective date: January 16, 2003, and shall be implemented within 
120 days of the date of issuance. Implementation includes the 
incorporation of changes to the Fort Calhoun Station Updated Safety 
Analysis Report as described in the licensee's application dated 
October 8, 2002.
    Amendment No.: 214.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68741).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2003.
    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 22, 2002, as supplemented by letter 
dated October 8, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.19, Surveillance Requirement (SR) 3.0.4 and adds 
TS 5.20 and SR 3.0.5 to extend the delay period before entering a 
limiting condition for operation following a missed surveillance.
    Date of issuance: January 16, 2003.
    Effective date: January 16, 2003, and shall be implemented with 120 
days from the date of issuance, including the incorporation of changes 
to the technical specification Bases as described in the licensee's 
application dated July 22, 2002, as supplemented by letter dated 
October 8, 2002.
    Amendment No.: 215.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56326).
    The supplemental letter of October 8, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 23, 2002.
    Brief description of amendments: These amendments revised Technical 
Specification (TS) section 2.6.2.4, ``Residual Heat Removal [RHR] 
Suppression Pool Cooling,'' to adopt TS Task Force (TF) change 230, 
Revision 1 (TSTF-230, Revision 1). This change to Required Action B of 
Limiting Condition for Operation 3.6.2.3 allows two RHR suppression 
pool cooling subsystems to be inoperable for up to 8 hours.
    Date of issuance: January 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 207, 181.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66012). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 17, 2002, as 
supplemented December 31, 2002.
    Brief description of amendment: The amendment provides a one-time 
change to Technical Specification (TS) 4.8.1.1.2.h.14 to allow the 
testing of Hope Creek's emergency diesel generator (EDG) lockout relays 
to be performed at power until startup from its eleventh refueling 
outage (spring 2003). The current TS surveillance requirement only 
allows the EDG lockout relays to be tested during shutdown conditions. 
PSEG requested that the TS change be issued on an exigent basis in 
accordance with title 10 of the Code of Federal Regulations (10 CFR), 
part 50, section 50.91(a)(6). Approval and implementation of the TS 
change allows the testing that has been completed to be used to comply 
with TS 4.8.1.1.2.h.14.
    Date of issuance: January 10, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Public Comments Requested as to Proposed No Significant Hazards 
Consideration: Yes (67 FR 79163) December 27, 2002. That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by January 27, 2003, but indicated that if the Commission makes 
a final no significant hazards consideration determination, any such 
hearing would take place after the issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, final determination of no significant hazards 
consideration, and state consultation are contained in a safety 
evaluation dated January 10, 2003.
    NRC Section Chief: James W. Clifford.

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: February 18, 2002, as supplemented in 
letter dated July 23, 2002.
    Brief description of amendments: The proposed amendments revised

[[Page 5687]]

Technical Specification 3/4.6.1.7, ``Containment Ventilation System,'' 
to extend the intervals between operability tests of the normal and 
supplementary containment purge valves.
    Date of issuance: January 7, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-147; Unit 2-135.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12608). The July 23, 2002, supplemental letter provided clarifying 
information that was within the scope of the original Federal Register 
notice (67 FR 12608) and did not change the initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 7, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 22, 2002.
    Brief description of amendments: The proposed amendments revised 
Technical Specification 3/4.3.5, allowing the automatic operation of 
the atmospheric steam relief valves during Mode 2 to maintain secondary 
side pressure at or below an indicated steam generator pressure of 1225 
psig during startup and shutdown of the reactors.
    Date of issuance: January 13, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-148; Unit 2-136.
    Facility Operating License Nos. NPF-76 and NPF-80: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45571).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 13, 2003.
    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 28th day of January 2003.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-2415 Filed 2-3-03; 8:45 am]
BILLING CODE 7590-01-P