[Federal Register Volume 68, Number 32 (Tuesday, February 18, 2003)]
[Notices]
[Pages 7810-7827]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-3689]


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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

[[Page 7811]]

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 24, 2003, through February 6, 2003. 
The last biweekly notice was published on February 4, 2003 (68 FR 
5668).

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 20, 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 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

[[Page 7812]]

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, 304-415-4737 or by e-mail to 
[email protected].

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

    Date of amendments request: December 13, 2002.
    Description of amendments request: The proposed amendments would 
revise Technical Specification 3.5.2, Emergency Core Cooling System--
Operating, by removing the Note that modifies the Limiting Condition 
for Operation. The proposed change would remove the requirement to have 
the charging pumps operable when thermal power is greater than 80% of 
rated thermal power (RTP). The proposed change would also remove 
Surveillance Requirement 3.5.2.4 for verifying the required charging 
pump flow rate.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The charging pumps were credited in the previous analysis to 
mitigate the consequences of a small-break loss-of-coolant accident 
(LOCA) above 80% of rated thermal power (RTP). The charging pumps 
were not considered to be an initiator of the accident. The new 
analysis for the small-break LOCA does not assume the charging pumps 
are initiators of the accident. Therefore, removing the requirement 
to maintain the charging pumps operable above 80% RTP and removing 
Surveillance Requirement 3.5.2.4 from the Technical Specification 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The consequence of a small-break LOCA is the potential for 
inadequate core cooling and decreased negative reactivity such that 
the reactor core is not protected after the design basis event. The 
previous analysis for the small-break LOCA above 80% RTP assumed 
unborated flow from a single charging pump to ensure there was 
adequate cooling flow delivered to the Reactor Coolant System. The 
revised small-break LOCA analysis was performed such that flow from 
the charging pumps was not credited. Since the charging pump flow is 
no longer credited in the small-break LOCA analysis, the proposed 
changes do not involve a significant increase in the consequences of 
a small-break LOCA.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different [kind] 
of accident from any accident previously evaluated.
    This request[ed] change does not involve a change in the 
operation of the plant and no new accident initiation mechanism is 
created by the proposed changes. Since the charging pump flow is no 
longer credited in the small-break LOCA analysis, the requirement to 
have the charging pumps operable above 80% RTP and the charging pump 
Surveillance Requirement 3.5.2.4 can be removed from the Technical 
Specification. 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. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety function of the Emergency Core Cooling System is to 
provide core cooling and negative reactivity, to ensure that the 
reactor core is protected after design basis events. For a small-
break LOCA, the previous analysis credited flow from the charging 
pumps above 80% RTP to supply supplemental cooling flow to the 
Reactor Coolant System. Credit for flow from a single charging pump 
was only taken for the water inventory.
    The revised small-break LOCA analysis was performed using the 
newest Nuclear Regulatory Commission accepted versions of the 
Westinghouse evaluation models for Combustion Engineering designed 
pressurized water reactors. The revised small-break LOCA analysis 
incorporated several changes to plant parameters used in the 
analysis, one of which was the elimination of the need to credit the 
charging pump flow above 80% RTP. Since the charging pump flow is no 
longer credited in the small-break LOCA analysis, the requirement to 
have the charging pumps operable above 80% RTP and charging pump 
Surveillance Requirement 3.5.2.4 can be removed from the Technical 
Specification.
    The proposed change to Technical Specification 3.5.2 does not 
modify any other charging pump requirements specified in the 
Technical Requirements Manual (e.g., requirements on charging pump 
availability for boration and cooldown remain in effect).
    Therefore, the safety function of the Emergency Core Cooling 
System is maintained and the margin of safety is not significantly 
reduced by the proposed changes.


[[Page 7813]]


    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Richard J. Laufer.

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 amendments request: November 12, 2002.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications, as necessary, to support an 
expansion of the core flow operating range (i.e., Maximum Extended Load 
Line Limit Analysis Plus (MELLLA+)). As part of the MELLLA+ 
implementation, Carolina Power & Light Company would implement the 
Detect and Suppress Solution-Confirmation Density (DSS-CD) approach to 
automatically detect and suppress neutronic/thermal-hydraulic 
instabilities.
    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:

    10 CFR 50.91(a) states ``At the time a licensee requests an 
amendment, it must provide to the Commission its analysis about the 
issue of no significant hazards consideration using the standards in 
Sec.  50.92.'' The following provides this analysis for the MELLLA+ 
operating range to a minimum core flow rate of 85% of rated with 
120% of the original licensed thermal power.
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The expansion of the core operating range discussed herein will 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    The probability (frequency of occurrence) of a DBA [design-basis 
accident] occurring is not affected by the operating range 
expansion, because the plant continues to comply with the regulatory 
and design basis criteria established for plant equipment (ASME 
[American Society of Mechanical Engineers] code, IEEE [Institute of 
Electrical and Electronics Engineers] standards, NEMA [National 
Electrical Manufacturers Association] standards, Regulatory Guides, 
etc.). An evaluation of the probabilistic safety assessments 
concludes that the calculated core damage frequencies do not 
significantly change due to the MELLLA+ operating range expansion. 
Scram setpoints (equipment settings that initiate automatic plant 
shutdowns) are established such that there is no significant 
increase in scram frequency due to the MELLLA+ operating range 
expansion. No new challenge to safety related equipment results from 
the MELLLA+ operating range expansion. The changes in consequences 
of hypothetical accidents, which occur from operation in the MELLLA+ 
region, are in all cases insignificant. The MELLLA+ accident 
evaluations do not exceed any NRC-approved acceptance limits. The 
spectrum of hypothetical accidents and abnormal operational 
occurrences has been investigated, and will meet the plant's 
currently licensed regulatory criteria. In the area of core design, 
for example, the fuel operating limits such as Maximum Average 
Planar Linear Heat Generation Rate (MAPLHGR) and Safety Limit 
Minimum Critical Power Ratio (SLMCPR) are met, and fuel reload 
analyses will show plant transients meet the criteria accepted by 
the NRC as specified in [GE Nuclear Energy, ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A and NEDE-
24011-P-A-US, (latest approved revision)]. Challenges to fuel (ECCS 
[emergency core cooling system] performance) are evaluated, and 
shown to still meet the criteria of 10 CFR 50.46, Appendix K, 
Regulatory Guide 1.70, and UFSAR [Updated Final Safety Analysis 
Report] Section 6.3. Challenges to the containment have been 
evaluated, and the containment and its associated cooling systems 
meet 10 CFR 50 Appendix A Criterion 38, Long Term Cooling, and 
Criterion 50, Containment. Radiological release events (accidents) 
have been evaluated, and shown to meet the regulatory limits of 10 
CFR 50.67.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The MELLLA+ operating range expansion will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Equipment that could be affected by MELLLA+ 
has been evaluated and no new operating mode, safety related 
equipment lineup, accident scenario, or equipment failure mode was 
identified. The full spectrum of accident considerations, defined in 
the UFSAR, has been evaluated, and no new or different kind of 
accident has been identified. The MELLLA+ operating range expansion 
uses fully developed technology, and applies it within the 
capabilities of existing plant equipment. The technology includes 
NRC approved codes, standards and methods applied in accordance with 
existing regulatory criteria.
    3. Will the change involve a significant reduction in a margin 
of safety?
    The MELLLA+ operating range expansion will not involve a 
significant reduction in a margin of safety. The calculated loads on 
all affected structures, systems and components have been shown to 
remain within design allowables for all design basis event 
categories. No NRC acceptance criterion is exceeded. The margins of 
safety currently included in the design of the plant are not 
affected by the MELLLA+ operating range expansion. Because the plant 
configuration and response to transients and hypothetical accidents 
do not result in exceeding the presently approved NRC acceptance 
limits, operation in the MELLLA+ region does not involve a 
significant reduction in a margin of safety.
    Conclusion: A MELLLA+ operating range expansion to a minimum 
core flow rate of 85% of rated with 120% of original licensed 
thermal power has been investigated. The BSEP [Brunswick Steam 
Electric Plant] licensing requirements have been evaluated and it 
has been demonstrated that this MELLLA+ operating range expansion 
can be accommodated:
    [sbull] Without a significant increase in the probability or 
consequences of an accident previously evaluated,
    [sbull] Without creating the possibility of a new or different 
kind of accident from any accident previously evaluated, and
    [sbull] Without exceeding any presently existing regulatory 
limits or acceptance criteria applicable to the plant, which might 
cause a reduction in a margin of safety.
    Having made negative declarations regarding the 10 CFR 50.92 
criteria, this assessment concludes that an operating range 
expansion to a minimum core flow rate of 85% of rated with 120% of 
original licensed thermal power does not involve a Significant 
Hazards Consideration.
    10 CFR 50.91(a) states ``At the time a licensee requests an 
amendment, it must provide to the Commission its analysis about the 
issue of no significant hazards consideration using the standards in 
Sec.  50.92.'' The following provides this analysis for the DSS-CD 
long-term stability solution.
    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will implement DSS-CD as the long-term 
stability solution. The DSS-CD solution is designed to identify the 
power oscillation upon inception and initiate control rod insertion 
to terminate the oscillations prior to any significant amplitude 
growth. The DSS-CD provides protection against violation of the 
Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated 
oscillations. Compliance with General Design Criteria (GDC) 10 and 
12 of 10 CFR part 50, Appendix A is accomplished via an automatic 
action. The DSS-CD introduces an enhanced detection algorithm that 
detects the inception of power oscillations and generates an earlier 
power suppression trip signal exclusively based on successive period 
confirmation recognition. The existing Option III algorithms are 
retained (with generic setpoints) to provide defense-in-depth 
protection for unanticipated reactor instability events.
    A developing instability event is suppressed by the DSS-CD 
system with substantial margin to the SLMCPR and no clad damage, 
with the event terminating in a scram and never developing into an 
accident. In addition, the DSS-CD solution defense-in-depth features 
incorporate all the

[[Page 7814]]

backup scram algorithms plus the licensed scram feature of the 
existing Option III system. The DSS-CD system does not interact with 
equipment whose failure could cause an accident. Scram setpoints in 
the DSS-CD will be established so that analytical limits are met. 
The reliability of the DSS-CD will meet or exceed that of the 
existing system. No new challenges to safety-related equipment will 
result from the DSS-CD solution. Because an instability event would 
reliably terminate in an early scram without impact on other safety 
systems, there is no significant increase in the probability of an 
accident.
    Proper operation of the DSS-CD system does not affect any 
fission product barrier or Engineered Safety Feature. Thus, the 
proposed change cannot change the consequences of any accident 
previously evaluated. As stated above, the DSS-CD solution meets the 
requirements of GDC 10 and 12 by automatically detecting and 
suppressing design basis thermal-hydraulic oscillations prior to 
exceeding the fuel SLMCPR.
    Based on the above, the operation of the DSS-CD solution within 
the framework of the Option III OPRM hardware will not increase the 
probability or consequences of an accident previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The DSS-CD solution operates within the existing Option III OPRM 
[Oscillation Power Range Monitor] hardware. No new operating mode, 
safety-related equipment lineup, accident scenario, system 
interaction, or equipment failure mode was identified. Therefore, 
the DSS-CD solution will not adversely affect plant equipment.
    Because there are no hardware design changes * * *, there is no 
change in the possibility or consequences of a failure. The worst 
case failure of the equipment is a failure to initiate mitigating 
action (i.e., scram), but no failure can cause an accident of a new 
or different kind than any previously evaluated.
    Based on the above, the proposed change to the DSS-CD solution 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    The DSS-CD solution is designed to identify the power 
oscillation upon inception and initiate control rod insertion to 
terminate the oscillations prior to any significant amplitude 
growth. The DSS-CD solution algorithm will maintain or increase the 
margin to the SLMCPR for anticipated instability events. The safety 
analyses in NEDC-33075P * * * demonstrate the margin to the SLMCPR 
for postulated bounding stability events. As a result, there is no 
impact on the MCPR [minimum critical power ratio] Safety Limit 
identified for an instability event.
    The current Option III algorithms (Period Based Detection, 
Amplitude Based, and Growth Rate) are retained (with generic 
setpoints) to provide defense-in-depth protection for unanticipated 
reactor instability events.
    Based on the above, the proposed change will not involve a 
significant reduction in the margin of safety.
    Conclusions: The DSS-CD stability solution has been 
investigated. The BSEP licensing requirements have been evaluated 
and it has been demonstrated that the DSS-CD stability solution can 
be accommodated:
    [sbull] Without a significant increase in the probability or 
consequences of an accident previously evaluated,
    [sbull] Without creating the possibility of a new or different 
kind of accident from any accident previously evaluated, and
    [sbull] Without exceeding any presently existing regulatory 
limits or acceptance criteria applicable to the plant, which might 
cause a reduction in a margin of safety.
    Having made negative declarations regarding the 10 CFR 50.92 
criteria, this assessment concludes that the DSS-CD stability 
solution does not involve a Significant Hazards Consideration.

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

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

    Date of amendment request: June 7, 2002, supplemented by letter 
dated January 8, 2003.
    Description of amendment request: The proposed amendments would 
revise the Updated Final Safety Analysis Report to eliminate credit for 
the flow path from the spent fuel pool to the high pressure injection 
pump as one source of primary system makeup following a tornado. The 
proposed amendments would also credit the Standby Shutdown Facility as 
the assured means of achieving safe shutdown for all three Oconee units 
following a tornado. By letter dated January 8, 2003, Duke Energy 
Corporation provided a revised No Significant Hazards Consideration 
(NSHC) that supercedes the NSHC that was noticed in the Federal 
Register on July 23, 2002 (67 FR 48216).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has 
made the determination that this amendment request involves a No 
Significant Hazards Consideration by applying the standards 
established by the NRC regulations in 10 CFR 50.92. This ensures 
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.
    The changes being requested in this amendment request involve 
(1) the elimination of the Spent Fuel Pool [SFP] as a suction source 
to a High Pressure Injection [HPI] pump for primary system make-up, 
and (2) to fully credit the Standby Shutdown Facility (SSF) as the 
primary assured means of achieving safe shutdown of all three units 
following a tornado. Following the modification to fully tornado 
protect the SSF, this facility becomes the station's assured flow 
path for both primary make-up and secondary decay heat removal for 
all three units.
    Although the probability of a severe tornado strike at the 
station does not change, new tornado insights gained from a review 
of the current external event risk analysis have resulted in an 
enhanced risk model that more accurately characterizes station 
tornado damage risk. The proposed changes are part of the revised 
tornado mitigation strategy that provides for an assured, 
deterministic success path rather than the current strategy that is 
based on risk insights and diversity for achieving safe shutdown. 
This effort has resulted in an overall reduction in tornado risk at 
the station and consequently, would not result in a significant 
increase in the consequences of an accident previously evaluated.
    Other than the fortification of walls of existing structures to 
harden them against tornado damage, there are no physical changes to 
the plant structures, systems, or components (SSCs), nor are there 
any changes to safety limits or set points. Also, no new 
radiological release pathways are created.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The changes being proposed in this amendment request do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The initial placement of the SFP-
HPI flow path into the LB [licensing basis] was based on 1989 risk 
analyses that showed a potential need for primary make-up due to 
inventory losses from a reactor coolant pump (RCP) seal loss-of-
cooling accident (LOCA). The upgrade of the RCP seals has 
significantly reduced the probability of a seal LOCA and 
subsequently, alleviated the initial reliance on the SFP-HPI flow 
path for primary make-up. If multi-unit primary make-up and decay 
heat removal are required following an event, the tornado protected 
SSF RBMU [sic] [(RCMU) reactor coolant makeup] or SSF ASW [auxiliary 
service water] pumps have the capabilities to perform these 
functions for all three units.
    3. Involve a significant reduction in a margin of safety.

[[Page 7815]]

    As mentioned previously, new tornado insights gained from a 
review of the current external event risk analysis have resulted in 
an enhanced risk model that more accurately characterizes station 
tornado damage risk. The proposed changes are part of the revised 
tornado mitigation strategy that provides for an assured, 
deterministic success path rather than a strategy that is based on 
risk insights and diversity for achieving safe shutdown.
    There is no safety limit, set point, or design parameter changes 
required. The integrity of the fuel cladding, reactor coolant 
system, and containment are preserved. Thus, 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: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: December 30, 2002.
    Description of amendment request: The proposed amendment revises 
two technical specifications (TSs). The first change proposes to revise 
TS 2.1.1.2, ``Minimum Critical Power Ratio Safety Limit (MCPRSL)'' to 
support operation during Cycle 17 with a mixed core. The second change 
proposes to revise the local power range monitor (LPRM) calibration 
frequency specified in the TS for the oscillation power range monitor 
(OPRM) in Surveillance Requirement (SR) 3.3.1.3.2. This change will 
correct an inconsistency between the LPRM calibration frequency 
specified in SR 3.3.1.3.2 and SR 3.3.1.1.7, ``Reactor Protection System 
(RPS) Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below. The licensee addresses each 
change separately.

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    1. The requested change to TS 2.1.1.2, MCPRSL to support the 
cycle 17 core loading does not involve any plant modifications or 
operational changes that could affect system reliability, 
performance, or possibility of operator error. The requested changes 
do not affect any postulated accident precursors, do not affect any 
accident mitigation systems, and do not introduce any new accident 
initiation mechanisms. The consequences of accidents previously 
evaluated are not changed because the number of rods that are 
protected from transition boiling is predicted to be greater than 
99.9 percent which meets the acceptance criterion in NUREG-0800, 
Section 4.4.
    2. The requested change to SR 3.3.1.3.2, OPRM/LPRM calibration 
frequency, does not involve a modification to the plant or introduce 
the probability of an operator error. The LPRMs are not the 
precursor to any accident. Making the LPRM surveillance frequency 
for the OPRM consistent with that approved for the RPS/APRM [reactor 
protection system/average power range monitor] does not change 
system reliability. The proposed LPRM surveillance frequency is 
supported by the uncertainties used to perform the MCPRSL analyses. 
Therefore, the number of rods that are calculated to experience 
transition boiling during normal operation or anticipated 
operational occurrences will not be changed and the consequences of 
these events will not be increased.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    1. The ATRIUM-10 fuel to be used in cycle 17 is compatible with 
the co-resident SVEA-96 fuel. This compatibility is demonstrated by 
application of the FRA-ANP critical power methodology to the core 
design that includes the ATRIUM-10 and SVEA-96 fuel. The proposed 
changes do not represent any new modes of operation, changes in 
setpoints or plant modifications other than those required for the 
reactor core. The change does not introduce new postulated accident 
precursors or mitigation systems. Reload design and analysis will be 
performed in accordance with approved NRC methodology.
    2. Increasing the time interval for the OPRM/LPRM surveillance 
reduces the frequency to be consistent with the LPRM surveillance 
frequency for the RPS/APRM and does not involve a modification to 
the plant, introduce a new operator error or revise setpoints.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    1. The proposed MCPRSL does not involve a significant reduction 
in the margin of safety associated with the criterion set forth in 
NUREG-0800, section 4.4. The safety limit established for the core 
ensures that the criterion for the number of fuel rods allowed to 
experience transition boiling will be maintained for normal plant 
operation and anticipated operational transients.
    The core operating limits will continue to be determined using 
methodologies that have been approved by the NRC.
    2. The proposed LPRM surveillance frequency is supported by the 
uncertainties used to perform the MCPRSL analyses. Therefore, the 
number of rods that are calculated to experience transition boiling 
during normal operation or anticipated operational occurrences will 
not be changed.
    Therefore, implementation of the change to the MCPRSL and the 
LPRM surveillance frequency 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 10, 2002, as supplemented on 
November 22, 2002, and January 28, 2003. This notice supercedes 67 FR 
68735 published on November 12, 2002, which erroneously stated that the 
October 10, 2002, application was a supplement of the licensee's 
application dated December 12, 2001. The October 10, 2002, replaced the 
December 12, 2001, application. This notice also adds supplements dated 
November 22, 2002, and January 28, 2003.
    Description of amendment request: The proposed amendment would 
change the Technical Specification Tables 3.2.A, 3.2.B, 4.2.A, and 
4.2.B. The proposed changes affect various instrument trip level 
settings and decreases the calibration frequencies for a variety of 
instruments. The proposed changes also involve clarifications to the 
Reactor Water Cleanup system trip configuration and the titles of 
certain trip systems. In addition, the proposed changes would make 
certain editorial and administrative corrections. The proposed setpoint 
changes and calibration frequencies are based on the licensee's 
evaluation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 7816]]


    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The methodology used to determine the proposed trip level 
settings and surveillance intervals ensure adequate performance of 
the affected instrumentation. In addition, the affected instruments 
are not initiators of any accident previously evaluated. Therefore, 
the proposed trip level setting and surveillance intervals will not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed changes to trip level settings and surveillance 
intervals were establish using methodologies subject to 10 CFR 
Appendix B Quality Assurance program and ensure existing 
radiological limits are met. Therefore, the proposed trip level 
settings and surveillance intervals will not involve a significant 
increase in the consequences of an accident previously evaluated.
    Other changes are editorial or administrative in nature and can 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    No new or different [kind] of accidents or malfunctions than 
those previously analyzed in Pilgrim's UFSAR [Updated Final Safety 
Analysis Report] are introduced by this proposed change because 
there are no new failure modes introduced. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Will not involve a significant reduction in the margin of 
safety.
    The proposed changes to trip level settings and surveillance 
intervals were established using approved methodologies subject to a 
10 CFR, Appendix B, Quality Assurance program and existing 
radiological limits are met. These changes do not impact Pilgrim's 
configuration or operation.
    Editorial and administrative type changes do not impact the 
operation or configuration of Pilgrim. For the above reasons the 
proposed change does not result in a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: December 4, 2002. This notice supercedes 
68 FR 2801 published on January 21, 2003, which erroneously stated that 
the December 4, 2002, application was a supplement of the licensee's 
application dated May 1, 2002. The December 4, 2002, application 
replaced the May 1, 2002, application in its entirety.
    Description of amendment request: The proposed amendment would 
extend the applicability of the current Pilgrim Nuclear Power Station 
(Pilgrim) reactor pressure vessel pressure-temperature (P-T) curves 
through the end of Operating Cycle (OC) 16. The current P-T curves were 
approved for use in License Amendment 190, dated April 13, 2001, and 
are limited to use through the end of OC 14. The proposed change would 
delete the 20 and 32 Effective Full Power Year (EFPY) curves and 
replace the wording of the title blocks to allow use through the end of 
OC 16. The proposed amendment would change Pilgrim Technical 
Specification Figures 3.6.1, 3.6.2, and 3.6.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change involves a request to extend the use of the 
current reactor pressure vessel P-T curves for two additional OCs. 
The P-T curves were generated in accordance with the fracture 
toughness requirements of 10 CFR part 50, Appendix G, and American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code 
(ASME Code), section XI, Appendix G and Regulatory Guide 1.99, 
Revision 2, Radiation Embrittlement of Reactor Vessel Materials, and 
were established in compliance with the methodology used to 
calculate and predict effects of radiation on embrittlement of 
reactor pressure vessel beltline materials. There are no physical 
changes to the plant or new modes of operation being introduced by 
the proposed change. Further, the proposed change does not involve a 
change to any activities or equipment and is not assumed in the 
safety analysis to initiate any accident sequence. The proposed 
change does not adversely affect the integrity of the reactor 
coolant pressure boundary such that its function in the containment 
of radioactive materials is affected. Additionally, the proposed 
change will not create any failure mode not bounded by previously 
evaluated accidents. Therefore, the proposed change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The current P-T curves were generated in accordance with the 
fracture toughness requirements of 10 CFR part 50, Appendix G, and 
ASME Code, section XI, Appendix G, and were approved by the U.S. 
Nuclear Regulatory Commission for use through OC 14. The proposed 
change would extend use of the P-T curves for two additional OCs. No 
new modes of operation are introduced by the proposed change. Plant 
operation in compliance with the current P-T curves ensures 
conditions in which brittle fracture of primary coolant pressure 
boundary materials is avoided. Accidents involving a breach of the 
primary coolant pressure boundary have previously been evaluated and 
no other types of accidents associated with the proposed change have 
been identified. 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?
    The proposed curves were established in compliance with the 
methodology used to calculate and predict effects of radiation on 
embrittlement of reactor pressure vessel beltline materials and are 
estimated for 48 effective full-power years. The current curves are 
approved for use through the end of OC 14 ([sim]19 EFPYs) which 
provides a conservatism factor of 1.7 between the actual EFPYs at 
the end of OC 14 and the end-of-life curve (32 EFPY). The change 
would extend the use of the proposed curves to the end of OC 16 
([sim]23 EFPYs) which provides a conservatism factor of 
approximately 2.0. The actual EFPYs at the end of OC 16 is bounded 
by the 48 EFPYs estimated for the current curves. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

[[Page 7817]]

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

    Date of amendment request: December 12, 2002.
    Description of amendment request: The proposed amendment would add 
a new Surveillance Requirement (SR) to the technical specification (TS) 
section 3.7.5, ``Auxiliary Feedwater (AF) System,'' which requires 
operation of the diesel-driven AF pump on a monthly frequency (i.e., 
once every 31 days) for greater than or equal to 15 minutes. The 
current TS SR 3.7.5.3 requires both the diesel-driven AF pump and the 
motor-driven AF pump to be operated once per quarter in accordance with 
the Inservice Testing Program; however, based on operating experience, 
Braidwood and Byron Stations conduct the diesel-driven AF pump 
surveillance on a monthly frequency to maintain a high level of 
assurance that the diesel engine would automatically start when called 
upon to perform its design basis function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change adds a new TS SR to the AF System TS section 
3.7.5. The new SR requires that the diesel-driven AF pump be 
operated for greater than or equal to 15 minutes every month. 
Operating experience has shown that conducting the diesel-driven AF 
pump surveillance on a monthly frequency maintains a high level of 
assurance that the diesel engine will automatically start when 
called upon to perform its design basis function.
    The previously analyzed events are initiated by the failure of 
plant structures, systems, or components. The AF system is not 
considered an initiator for any of these previously analyzed events. 
The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system, or component that 
initiates an analyzed event. No active or passive failure mechanisms 
that could lead to an accident are affected. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The initial conditions of design basis accident and transient 
analyses in the Byron/Braidwood Stations Updated Final Safety 
Analysis Report assume the AF system is operable. The operability of 
the AF system is assured by the proposed TS SR and is consistent 
with the initial assumptions of the accident analyses. Since 
functionality of the diesel engine can be better assured when the 
diesel-driven AF pump is operated monthly vice quarterly, Exelon is 
proposing to add a TS SR to operate the diesel-driven AF pump on a 
monthly frequency. The proposed SR will provide higher confidence 
that the diesel-driven AF pump will reliably start automatically 
during an emergency condition, consistent with the AF System design 
requirements, and continue to mitigate the consequences of the 
associated design basis accidents. Based on this evaluation, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve the use or installation of 
new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases. The current diesel-driven AF pump 
surveillance procedure is already conducted on a monthly basis and 
has been reviewed, approved and judged appropriate to provide high 
confidence that the AF diesel engine and pump will reliably start 
and operate during an emergency condition. The new SR formalizes 
this monthly surveillance practice in the TS. Based on this 
evaluation, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not alter any existing setpoints at 
which protective actions are initiated and no new setpoints or 
protective actions are introduced. The design and operation of the 
AF system remains unchanged and maintains the existing margins of 
safety. Since the increased frequency of the diesel-driven AF pump 
surveillance test maintains high assurance that the pump's diesel 
engine will successfully auto-start during an emergency, the 
proposed additional SR will provide high confidence that the AF 
system will continue to function as designed. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Based on the above, Exelon concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c).

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

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: December 23, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) section 6, Administrative 
Controls, to: (1) relocate administrative requirements discussed in 
Administrative Letter 95-06 (AL 95-06), ``Relocation of Technical 
Specification Administrative Controls Related to Quality Assurance,'' 
to the Operational Quality Assurance Program, (2) change the title of 
the senior onsite official, and (3) bring the TSs into consistency with 
changes in 10 CFR part 20.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the Seabrook Station TS do not adversely 
affect accident initiators or precursors nor alter the design 
assumptions, conditions, and configuration of the facility or the 
manner in which the plant is operated and maintained. In addition, 
the proposed changes do not affect the manner in which the plant 
responds in normal operation, transient or accident conditions nor 
do they change any of the procedures related to operation of the 
plant. The proposed changes do not alter or prevent the ability of 
structures, systems and components (SSCs) to perform their intended 
function to mitigate the consequences of an initiating event within 
the acceptance limits assumed in the Updated Final Safety Analysis 
Report (UFSAR). The proposed changes are administrative and 
editorial for the purpose of correcting or updating TS to reflect 
current NRC [Nuclear Regulatory Commission] and industry 
initiatives.
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types

[[Page 7818]]

and amounts of radioactive effluent that may be released offsite, 
nor significantly increase individual or cumulative occupational/
public radiation exposures.
    Therefore, it is concluded that these proposed revisions do not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes to the Seabrook Station TS do not change 
the operation or the design basis of any plant system or component 
during normal or accident conditions. The proposed changes do not 
include any physical changes to the plant. In addition, the proposed 
changes do not change the function or operation of plant equipment 
or introduce any new failure mechanisms. The plant equipment will 
continue to respond per the design and analyses and there will not 
be a malfunction of a new or different type introduced by the 
proposed changes.
    The proposed changes are administrative in nature and only 
correct, update and clarify the Seabrook Station Technical 
Specifications to reflect NRC guidance, i.e., AL 95-06. The proposed 
changes do not modify the facility nor do they affect the plant's 
response to normal, transient or accident conditions. The changes do 
not introduce a new mode of plant operation. The changes are an 
enhancement and do not affect plant safety. The plant's design and 
design basis are not revised and the current safety analyses remains 
in effect.
    Thus, these proposed revisions to the Seabrook Station TS do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in [a] margin of safety.
    The proposed changes are administrative changes to the Seabrook 
Station Technical Specifications. The safety margins established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety Limits as specified in the Technical 
Specifications are not revised nor is the plant design or its method 
of operation revised by the proposed changes. Thus, it is concluded 
that these proposed revisions to the Seabrook Station TS 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 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, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 17, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.9, ``Control Room Emergency 
Filtration System (CREFS),'' by deleting the one-time extension to the 
allowed outage time (AOT) for CREFS and the exception to the 
requirements of limiting condition for operation 3.0.4 and surveillance 
requirement 3.0.4 that were allowed during the AOT.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The operability of CREFS ensures that the control room will 
remain habitable for operators during and following all credible 
accident conditions. The inoperability or failure of CREFS is not an 
accident initiator or precursor. Therefore, the probability of an 
accident previously evaluated will not be significantly increased as 
a result of the proposed change. Because design limitations continue 
to be met and the integrity of the reactor coolant system pressure 
boundary is not challenged, the assumptions employed in the 
calculation of the offsite radiological doses remain valid. 
Therefore, the consequences of an accident previously evaluated will 
not be significantly increased as a result of the proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. The evaluation of the effects of the proposed changes 
indicate that all design standards and applicable safety criteria 
limits are met. These changes therefore do not cause the initiation 
of any new or different accident nor create any new failure 
mechanisms.
    Equipment important to safety will continue to operate as 
designed.
    Additionally, the changes do not result in any event previously 
deemed incredible being made credible. The changes also do not 
result in more adverse conditions or result in any increase in the 
challenges to safety systems. Therefore, operation of the Point 
Beach Nuclear Plant in accordance with the proposed amendments will 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed amendment will not otherwise affect the plant protective 
boundaries, will not cause a release of fission products to the 
public, nor will it degrade the performance of any other structures, 
systems or components (SSCs) important to safety. Therefore, 
deleting the one-time extension to the CREFS AOT will not result in 
a significant reduction in the margin of safety.

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

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

    Date of amendment request: June 28, 2002, as supplemented on 
December 18, 2002, and January 18, 2003.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by relaxing the secondary 
containment requirements and eliminating the Filtration, Ventilation, 
and Recirculation System (FRVS) charcoal filters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously analyzed?
    Response: No.
    The definition of CORE ALTERATIONS has been revised to define 
that control rod movement, provided there are no fuel assemblies in 
the associated core cell, is not a core alteration. This is 
consistent with Standard Technical Specifications (STS) NUREG-1433 
Vol.1, Rev. 2, Standard Technical Specifications, General Electric 
Plants, BWR/4 [Boiling Water Reactor, Type 4].
    The TS presently provide a period of 7 days to restore an 
inoperable FRVS ventilation unit when performing activities with the 
potential for draining the reactor vessel or discontinue such 
activities.

[[Page 7819]]

Operation of the redundant train will ensure that the remaining 
subsystem is operable, that no failures, which could prevent 
automatic actuation, have occurred and that any other failures will 
be readily detected. This is consistent with STS, NUREG-1433 Vol.1, 
Rev. 2, Standard Technical Specifications, General Electric Plants, 
BWR/4.
    The proposed changes associated with the FHA [fuel-handling 
accident] do not involve a change to structures, components, or 
systems that would affect the probability of an accident previously 
evaluated in the Hope Creek Updated Final Safety Analysis Report 
(UFSAR). The FHA for the HCGS [Hope Creek Generating Station] is 
defined as a drop of a fuel assembly over irradiated assemblies in 
the reactor core 24 hours after reactor shutdown. AST [accident 
source term] is used to evaluate the dose consequences of a 
postulated accident. The FHA has been analyzed without credit for 
Secondary Containment, Filtration Recirculation and Ventilation 
System (FRVS), and Control Room Emergency Filtration (CREF) system. 
The resultant radiological consequences are within the acceptance 
criteria set forth in 10 CFR 50.67 and Regulatory Guide 1.183. This 
amendment does not alter the methodology or equipment used directly 
in fuel handling operations. The equipment hatch, the personnel air 
locks, nor any other containment penetration, nor any component 
thereof is an accident initiator. Actual fuel handling operations 
are not affected by the proposed changes. Therefore, the probability 
of a Fuel Handling Accident is not affected with the proposed 
amendment. No other accident initiator is affected by the proposed 
changes.
    The Loss of Coolant Accident (LOCA) Dose Calculation has been 
revised to (1) eliminate credit for the FRVS recirculation charcoal 
filters, (2) reduce credited efficiency of FRVS vent charcoal 
filters, (3) reduce Engineered Safety Feature (ESF) leakage from 10 
gpm to 1 gpm and (4) reduce control room unfiltered in-leakage to 
350 cfm [cubic feet per minute]. These proposed changes do not 
eliminate any safety system. The changes are only associated with 
the credit provided by the system in reducing the radiological 
consequences and therefore, do not affect any accident initiator. 
The results of that analysis show that the Exclusion Area Boundary 
(EAB), Low Population Zone (LPZ), and Control Room (CR) doses are of 
the same order of magnitude as the previous analysis and remain 
within the acceptance criteria in 10 CFR 50.67 and Regulatory Guide 
1.183.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously analyzed.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    Response: No.
    The proposed amendment will not create the possibility for a new 
or different type of accident from any accident previously 
evaluated. Changes to the allowable activity in the primary and 
secondary systems do not result in changes to the design or 
operation of these systems. The evaluation of the effects of the 
proposed changes indicates that all design standard and applicable 
safety criteria limits are met.
    Equipment important to safety will continue to operate as 
designed. Component integrity is not challenged. The changes do not 
result in any event previously deemed incredible being made 
credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems. The 
systems affected by the changes are used to mitigate the 
consequences of an accident that has already occurred. The proposed 
TS changes and modifications do not significantly affect the 
mitigative function of these systems.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    (3) Does the change involve a significant reduction in [a] 
margin of safety?
    Response: No.
    The proposed changes revise the TS to establish operational 
conditions where specific activities represent situations during 
which significant radioactive releases can be postulated. These 
operational conditions are consistent with the design basis analysis 
and are established such that the radiological consequences are at 
or below the regulatory guidelines. Safety margins and analytical 
conservatisms are retained to ensure that the analysis adequately 
bounds all postulated event scenarios. The proposed TS continue to 
ensure that the TEDE [total effective dose equivalent] for the CR, 
the EAB, and LPZ are below the corresponding acceptance criteria 
specified in 10 CFR 50.67 and RG1.183.
    Therefore, these 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 Clifford.

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

    Date of amendment request: October 9, 2002, as supplemented 
November 22, 2002, and December 6, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 6.8.4.f, ``Primary Containment 
Leakage Rate Testing Program,'' to allow a one-time interval extension 
to the requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously analyzed?
    Response: No.
    The proposed revision to section 6.8.4.f adds a one-time 
extension to the current interval for containment integrated leak 
rate test (ILRT). The current test interval of 10 years, based upon 
past performance, would be extended on a one-time basis to 15 years 
from the last ILRT. The proposed extension to ILRT testing cannot 
increase the probability of an accident previously evaluated since 
the containment ILRT testing extension is not a modification to 
plant systems, nor a change to plant operation that could initiate 
an accident. The proposed extension to Type A testing does not 
involve a significant increase in the consequences of an accident 
since research documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' found that very few potential 
containment leakage paths fail to be identified by Type B and C 
tests. The NUREG concluded that reducing the ILRT testing frequency 
to once per twenty years would lead to an imperceptible increase in 
risk. Containment performance monitoring is performed in accordance 
with the Maintenance Rule (10 CFR 50.65) and inspections required by 
American Society of Mechanical Engineers (ASME) code are performed 
in order to identify indications of containment degradation that 
could affect leak tightness. Type B and C testing required by the 
technical specifications (TS) will identify any containment opening, 
such as valves, that would otherwise be detected by the ILRT. Reg. 
Guide 1.174 provides guidance for determining the risk impact of 
plant-specific changes to the licensing basis. It also recommends 
the use of risk analysis techniques to ensure and show that the 
proposed change is consistent with the defense-in-depth philosophy. 
The increase in large early release frequency (LERF) resulting from 
a change in the ILRT test frequency from the current once in every 
10 years to once in every 15 years is less than 1E-7 per year, 
thereby meeting Regulatory Guide 1.174 definition of a very small 
change in risk. The change in conditional containment failure 
probability (CCFP) is estimated to be 0.25% for the proposed change. 
These factors show that an ILRT test extension will not represent a 
significant increase in the consequences of an accident.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    Response: No.
    The proposed revision to section 6.8.4.f adds a one-time 
exception to the current interval for the ILRT. The current test 
interval of 10 years, based upon past performance, would be extended 
on a one-time basis to 15 years from the last Type A test. Primary 
containment is designed to contain energy and fission products 
during and after an event. The Individual Plant

[[Page 7820]]

Examination (IPE) identifies events that lead to containment 
failure. Revision to the ILRT test interval does not change this 
list of events. There are no physical changes being made to the 
plant and there are no changes to the operation of the plant that 
could introduce a new failure mode creating a new or different kind 
of accident.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    The proposed revision to section 6.8.4.f adds a one-time 
extension to the current interval for the ILRT. The current test 
interval of 10 years, based upon past performance, would be extended 
on a one-time basis to 15 years from the last ILRT. The proposed 
extension to ILRT testing interval 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 exception 
in ILRT leakage testing resulted in an imperceptible increase in 
risk to the public. NUREG-1493 found that the containment leakage 
rate contributes a very small amount to the individual risk, and 
that the decrease in Type A testing frequency would have a minimal 
affect on this risk since most potential leakage paths are detected 
by Type C testing. Type B and Type C testing will continue to be 
performed at a frequency currently required by the Technical 
Specifications (TS). The containment inspections being performed in 
accordance with ASME, section XI, and Maintenance Rule (10 CFR 
50.65) provide a high degree of assurance that the containment will 
not degrade in a manner that is only detectable by Type A testing.
    Reg. Guide 1.174 provides guidance for determining the risk 
impact of plant-specific changes to the licensing basis. It also 
recommends the use of risk analysis techniques to ensure and show 
that the proposed change is consistent with the defense-in-depth 
philosophy. The increase in large early release fraction (LERF) 
resulting from a change in the ILRT test frequency from the current 
once in every 10 years to once in every 15 years is less than 1E-7 
per year, thereby meeting Regulatory Guide 1.174 definition of a 
very small change in risk. The change in conditional containment 
failure probability (CCFP) is estimated to be 0.25% for the proposed 
change.
    Therefore, these changes do not involve a significant reduction 
in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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 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: July 25, 2002, as supplemented on 
October 21, 2002.
    Description of amendment request: The proposed change would revise 
Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, Technical 
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. The delay period would be extended 
from the current limit of up to 24 hours, to ``* * * up to 24 hours or 
up to the limit of the specified frequency, whichever is greater.'' In 
addition, the following requirement would be added to SR 4.0.3: ``A 
risk evaluation shall be performed for any surveillance delayed greater 
than 24 hours and the risk impact shall be managed.'' PSEG is also 
proposing changes to adopt a TS Bases Control Program and changes to SR 
4.0.1.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process (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). The licensee affirmed the applicability of the model NSHC 
determination for amendments concerning missed surveillances in its 
original application dated July 25, 2002. The proposed amendment would 
also make administrative changes to SRs 4.0.1 and 4.0.3 to be 
consistent with NUREG-1431, Revision 2, ``Standard Technical 
Specifications, Westinghouse Plants.'' These changes are necessary to 
make the current Salem TSs compatible with the proposed CLIIP changes 
for missed surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[Specification 4.0.3]

    The proposed change relaxes the time allowed to perform a missed 
Surveillance. The time between Surveillances is not an initiator to 
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.

[Specification 4.0.1]

    The proposed additional requirement equating failure to meet a 
surveillance with failure to meet the [limiting condition for 
operation] is consistent with current interpretation of the 
technical specifications. This change, along with relocation and 
rewording of existing requirements from Specification 4.0.3, are 
administrative in nature and do not adversely affect accident 
initiators, design functions, facility configuration or the manner 
of operation or control. The ability of structures, systems and 
components to perform their intended function remains unaffected.

[Bases Control Program]

    The proposed change to adopt a Technical Specification Bases 
Control Program is also administrative in nature and does not 
adversely affect accident initiators, design functions, facility 
configuration or the manner of operation or control. The ability of 
structures, systems or components to perform their intended function 
remains unaffected. Future changes to the TS Bases will continue to 
be administratively controlled in accordance with the requirements 
of 10 CFR 50.59.
    Therefore, these three changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    None of the three proposed changes involves 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. Thus, these changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?

[Specification 4.0.3]

    The [extended] time allowed to perform a missed Surveillance 
does not result in a significant reduction in the margin of safety. 
As supported by the historical data, the likely outcome of any 
Surveillance is verification that the LCO is met. Failure to perform 
a Surveillance within the prescribed Frequency does not cause 
equipment to become inoperable. The only effect of the additional 
time allowed to perform a missed Surveillance on the margin of 
safety is the

[[Page 7821]]

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.

[Specification 4.0.1]

    The proposed changes to TS 4.0.1, including relocation and 
rewording of existing requirements from Specification 4.0.3, are 
administrative in nature and do not reduce the level of programmatic 
or procedural controls associated with the Surveillance 
Requirements. There are no substantive differences in meaning or 
intent between the existing specifications and the corresponding STS 
requirements. Further, these changes have no impact on equipment 
design, configuration, analytical basis, setpoints or operation.

[Bases Control Program]

    The proposed change to adopt a Technical Specification Bases 
Control Program is also administrative in nature and does not reduce 
the level of programmatic or procedural controls associated with the 
Bases. There is no impact on equipment design, configuration, 
analytical basis, setpoints or operation.
    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.

    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.

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

    Date of amendment request: December 19, 2002.
    Description of amendment request: The proposed amendments would 
change the Operating Licenses and Technical Specifications associated 
with an increase in the licensed reactor power level of 1.5 percent for 
each reactor (from 2763 megawatts thermal (MWt) to 2804 MWt).
    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:

    [Southern Nuclear Company] SNC's conclusion that the proposed 
change to the Plant Hatch Unit 1 and 2 Operating Licenses and 
Technical Specifications does not involve a significant hazards 
consideration is based upon the following:
    1. The proposed amendment does not change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated[.]
    The comprehensive analytical efforts performed to support the 
proposed uprate conditions included a review and evaluation of all 
components and systems that could be affected by this change. 
Performance requirements for these systems were evaluated and found 
acceptable. Furthermore, evaluation of accident analyses confirmed 
the effects of the proposed uprate are bounded by the current dose 
analyses. The systems will function as designed. The performance 
requirements for these systems were evaluated and found acceptable.
    The primary loop components (e.g., reactor vessel, reactor 
internals, control rod drive housings, piping and supports, and 
recirculation pumps) continue to comply with their applicable 
structural limits and will continue to perform their intended design 
functions. Thus, the probability of a structural failure of these 
components is not increased as a result of this change.
    The Nuclear Steam Supply System (NSSS) systems will still 
perform their intended design functions during normal and accident 
conditions. The balance-of-plant (BOP) systems and components will 
continue to meet their applicable structural limits and perform 
their intended design functions. Thus, the probability of a 
structural failure of these components is not increased as a result 
of this change.
    The NSSS/BOP interface systems will continue to perform their 
intended design functions. The safety relief valves and containment 
isolation valves still meet design sizing requirements at the 
uprated power level.
    Because the integrity of the plant will not be affected by 
operation at the uprated condition, SNC concluded that all 
structures, systems, and components required to mitigate a transient 
remain capable of fulfilling their intended functions. The reduced 
uncertainty in the flow input to the core thermal power uncertainty 
measurement allows most of the current safety analyses to be used, 
with small changes to the core operating limits, to support 
operation at a core power of 2804 MWt. Other analyses performed at a 
nominal power level were either evaluated or reperformed for the 
1.5% increased power level. The results demonstrate that the 
applicable analysis acceptance criteria continue to be met at the 
1.5% uprate conditions. Thus, all Plant Hatch Final Safety Analysis 
Report accident analyses continue to demonstrate compliance with the 
relevant event acceptance criteria. The analyses performed to assess 
the effects of mass and energy release remain valid. The source 
terms used to assess radiological consequences were reviewed and 
determined to bound operation at the 1.5% uprated condition. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment will not change create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed change will have no adverse 
effect on any safety-related system 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 previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    Operation at the uprated power condition does not involve a 
significant reduction in a margin of safety. Analyses of the primary 
fission product barriers confirm that all relevant design criteria 
remain satisfied, both from the standpoint of the integrity of the 
primary fission product barrier and from the standpoint of 
compliance with the required acceptance criteria. As appropriate, 
all evaluations were performed using methods that were either 
reviewed and approved by the NRC, or are in compliance with 
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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

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: November 14, 2002.
    Description of amendment request: The proposed amendment would 
delete

[[Page 7822]]

the turbine missile design basis from the Updated Final Safety Analysis 
Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The turbine missile generation probability will not be 
significantly increased by elimination of the regulatory commitments 
in the UFSAR. No plant changes are proposed that would significantly 
increase the probability of turbine missile generation. Turbine 
missile generation does not pose a credible threat to safety related 
components and consequently has no potential to increase 
radiological consequences.
    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 involve no physical modification of the 
plant or different operating configurations.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Turbine missiles do not constitute a credible threat to nuclear 
safety at STP [South Texas Project]. They are not a consideration in 
any plant safety analysis. Changing the regulatory commitment with 
regard to design for turbine missiles has no effect on any margin of 
safety.
    Based upon the analysis provided herein, the proposed amendments 
do not involve a significant hazards consideration.

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

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

    Date of amendment request: January 14, 2003 (TS 02-08).
    Description of amendment request: The proposed amendment would 
revise applicability requirements for TS 3.3.9.4, ``Containment 
Building Penetrations.'' This revision will modify the current 
applicability requirement associated with movement of ``irradiated 
fuel'' by adding a new applicability statement for the containment 
building equipment door.
    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 revises the applicability of the containment 
building penetration function and associated action. This change 
does not alter the function of the penetrations but does revise when 
the feature is required to be available for the mitigation of 
postulated accidents. These penetrations only function to minimize 
the release of radioactive material for accident mitigation and are 
not considered to be a source of any postulated accident. The 
analysis verifies that a fuel handling accident (FHA) occurring at 
least 100 hours after being critical in a reactor core will not 
result in dose consequences above the regulatory limits without the 
containment closure function provided by the CBED [containment 
building equipment door]. The applicability and action for the CBED 
will not be changed when movement of recently irradiated fuel is in 
progress and this function ensures acceptable dose consequences. 
Therefore, the proposed change will not increase the probability of 
an accident because the penetration function has not been altered 
and this function is not a potential source for accidents. 
Additionally, the proposed change will not significantly increase 
the consequences of an accident because the analysis has verified 
that dose consequences will be maintained less than the required 
regulatory limits.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change only modifies when containment building 
penetrations need to be available for accident mitigation and does 
not alter their function, design, or operation. These penetrations 
only serve to minimize the release of radioactive material in the 
event of postulated accidents and do not have the potential to 
create an accident. Since the function of the penetrations is not 
being changed and they do not have an accident generation potential, 
the possibility of a new or different kind of accident is not 
created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change will not alter the function, design, or 
operation of the containment building penetrations for postulated 
accidents that require this feature for the mitigation of the event. 
The analysis has determined that the CBED availability can be 
limited to those activities that involve the movement of irradiated 
fuel that has been in a critical reactor core within the previous 
100 hours. Therefore, not requiring the CBED to be available 100 
hours or longer afterwards will not impact plant safety or result in 
dose consequences above established regulatory limits. The proposed 
change will not alter any setpoints or other functions that serve to 
maintain the safety limits. Therefore, the proposed change will not 
involve a significant reduction in a margin of safety.

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

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of amendment request: January 14, 2003.
    Description of amendment request: The proposed amendment will 
revise the Yankee Rowe Nuclear Power Station License and Technical 
Specifications to delete operational and administrative requirements 
that would no longer be required once the spent nuclear fuel has been 
transferred from the spent fuel pool to the Independent Spent Fuel 
Storage Installation (ISFSI).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed changes reflect the complete transfer of all 
spent nuclear fuel from the Spent Fuel Pit to the Independent Spent 
Fuel Storage Installation (ISFSI). Design basis accidents related to 
the Spent Fuel Pit are discussed in the YNPS FSAR. These postulated 
accidents are predicated on spent nuclear fuel being stored in the 
Spent Fuel Pit. With the removal of the spent fuel from the Spent 
Fuel Pit, there are no remaining important to safety systems 
required to be monitored and there are no remaining credible 
accidents that require that actions of a Certified Fuel Handler or 
non-Certified Fuel Handler to prevent occurrence or mitigate the 
consequences.
    The YNPS FSAR provides a discussion of radiological events 
postulated to occur as a result of decommissioning with the bounding 
consequence resulting from a materials handling event. The proposed 
changes do not have an adverse impact on decommissioning activities 
or any of their postulated consequences.

[[Page 7823]]

    The proposed change to the Design Features section of the 
Technical Specifications clarifies that the spent fuel is being 
stored in dry casks within an ISFSI. The probability or consequences 
of accidents at the ISFSI are evaluated in the dry cask vendor's 
FSAR and are independent of the accidents evaluated in the YNPS 
FSAR.
    Based on the above, the proposed changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes reflect the reduced operational risks 
as a result of the spent nuclear fuel being transferred to dry casks 
within an ISFSI. The proposed changes do not modify any physical 
systems, or components. The plant conditions for which the YNPS FSAR 
design basis accidents relating to spent fuel have been evaluated 
are no longer applicable. The aforementioned proposed changes do not 
affect any of the parameters or conditions that could contribute to 
the initiation of an accident. Design basis accidents associated 
with the dry cask storage of spent fuel are already considered in 
the dry cask system's Final Safety Analysis Report. No new accident 
scenarios are created as a result of deleting non-applicable 
operational and administrative requirements. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    No. As described above, the proposed changes reflect the reduced 
operational risks as a result of the spent nuclear fuel being 
transferred to dry casks within an ISFSI. The design basis and 
accident assumptions within the YNPS FSAR and the Defueled Technical 
Specifications relating to spent fuel are no longer applicable. The 
proposed changes do not affect remaining plant operations, systems, 
or components supporting decommissioning activities. In addition, 
the proposed changes do not result in a change in initial 
conditions, system response time, or in any other parameter 
affecting the course of a decommissioning activity accident 
analysis. Therefore, the proposed changes will not involve a 
significant reduction in the margin of safety.
    Based on the considerations noted above, it is concluded that 
the proposed changes will not endanger the public health and safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Scott W. Moore.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: January 16, 2003.
    Brief description of amendment request: The proposed amendment 
would revise the applicable Technical Specifications requirements for 
rod position monitoring during the current operating cycle (Cycle 22) 
to allow the use of an alternate method of determining rod position. 
This would be effective until repair of the indication system can be 
completed during the next shutdown of sufficient duration.
    Date of publication of individual notice in Federal Register: 
January 24, 2003 (68 FR 3566).
    Expiration date of individual notice: February 7, 2003, for 
comments; February 24, 2003, for hearings.

Florida Power and Light Company, Docket No. 50-251, Turkey Point Plant, 
Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: November 26, 2002.
    Brief description of amendments: The proposed license amendments 
would revise Technical Specifications (TSs) to increase the total spent 
fuel wet storage capacity by adding a spent fuel storage rack in the 
cask area in each unit's spent fuel pool. Also, it would revise the 
location called out in the Design Features sections 5.6.1.1a and b of 
the TSs referring to Updated Final Safety Analysis Report Appendix 14D, 
rather than referring to Westinghouse Report WCAP-14416-P.
    Date of publication of individual notice in the Federal Register: 
January 28, 2003 (68 FR 4246).
    Expiration date of individual notice: February 27, 2003.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 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].

[[Page 7824]]

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

    Date of amendment request: May 29, 2001, and its supplements dated 
August 29, 2001, and September 24, 2002.
    Brief description of amendment: The amendment revises paragraph 
2.C.(5), ``Physical Protection,'' of Facility Operating License No. 
DPR-61 to reference the Defueled Physical Security Plan that includes 
the security plan for the Independent Spent Fuel Storage Installation.
    Date of issuance: January 30, 2003.
    Effective date: January 30, 2003, and shall be implemented within 
30 days from the date of issuance and prior to the transfer of spent 
nuclear fuel to the Independent Spent Fuel Storage Installation.
    Amendment No.: 199.
    Facility Operating License No. DPR-61: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44163). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 30, 2003.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: August 8, 2002, as supplemented 
October 23, 2002.
    Brief description of amendment: The amendment authorized changes to 
the Updated Final Safety Analysis Report (USFAR) for Fermi 2 by 
allowing implementation of the Boiling Water Reactor Vessel and 
Internals Project reactor pressure vessel Integrated Surveillance 
Program as the basis for demonstrating the compliance with the 
requirements of Appendix H, ``Reactor Vessel Material Surveillance 
Program Requirements,'' to 10 CFR part 50.
    Date of issuance: January 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 152.
    Facility Operating License No. NPF-43: Amendment authorizes changes 
to the USFAR.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56320). The October 23, 2002, supplemental letter provided 
additional clarifying information that did not change the original no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the original notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
January 30, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 18, 2002, as supplemented 
August 7, and October 9 and October 30, 2002, and January 15, 2003.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.7.15 in response to Boraflex degradation to 
provide revised spent fuel pool (SFP) storage criteria, and revised 
fuel enrichment and burnup requirements that take credit for soluble 
boron. TS 4.3.1 is revised to increase the required soluble boron 
credit from a concentration of 730 parts per million (ppm) to 850 ppm 
to ensure acceptable levels of subcriticality in the SFPs. Associated 
changes to the TS Bases are also included.
    Date of issuance: February 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 210 & 191.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42820). The supplements dated August 7, and October 9 and October 30, 
2002, and January 15, 2003, provided clarifying information that did 
not change the scope of the April 18, 2002, application nor the initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 4, 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 deletes TS 5.5.3, 
``Post Accident Sampling System (PASS),'' and thereby eliminates the 
requirements to have and maintain the PASS at Columbia Generating 
Station. The amendment also addresses related changes to TS 5.5.2, 
``Primary Coolant Sources Outside Containment,'' and License Condition 
2.C.(13), ``Post Accident Sampling.''
    Date of issuance: January 27, 2003.
    Effective date: January 27, 2003, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 184.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78518). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 27, 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 letters 
dated July 9, August 2, September 16, and November 7 and 22, 2002.
    Brief description of amendment: This amendment increases the 
licensed power level by approximately 1.7 percent from 3,039 megawatts 
thermal (MWt) to 3,091 MWt. These changes result from increased 
feedwater flow measurement accuracy to be achieved by utilizing high 
accuracy ultrasonic flow measurement instrumentation.
    Date of issuance: January 31, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 129.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40022). The July 9, August 2, September 16, and November 7 and 22, 
2002, supplemental letters 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 31, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 24, 2001, as supplemented 
on May 22, 2002.
    Brief description of amendment: The amendment revised information 
in the

[[Page 7825]]

Final Safety Analysis Report regarding the protection of the component 
cooling water (CCW) system from natural phenomena. The change addresses 
the fact that a portion of one safety-related loop of the CCW system is 
routed through the fuel storage building, where the structure was not 
designed to protect the CCW piping from the effects of natural 
phenomena.
    Date of issuance: January 27, 2003.
    Effective date: January 27, 2003.
    Amendment No.: 214.
    Facility Operating License No. DPR-64: Amendment revised the Final 
Safety Analysis Report.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50466). The May 22, 2002, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated January 27, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 26, 2002, as revised by 
letters dated October 9 and 30, 2002.
    Brief description of amendment: The amendment revises the 
definition of Operable in Technical Specification (TS) 1.0.K with 
respect to support system requirements for alternating current power 
sources. Conforming changes are also made to a specific support system 
TS in Sections 3/4.5, ``Core and Containment Cooling Systems'', 3/4.7, 
``Station Containment Systems'', and 3/4.10, ``Auxiliary Electrical 
Power Systems,'' and associated Bases.
    Date of Issuance: February 4, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 213.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: (67 FR 78519). The 
Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated February 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 19, 2002, as 
supplemented December 26, 2002.
    Brief description of amendments: The amendments add a new 
analytical method to Technical Specifications (TS) section 5.6.5, 
``Core Operating Limits Report.'' The change supports the core design 
efforts used for the Unit 2 refueling outage which began on January 21, 
2003.
    Date of issuance: February 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 159 & 145.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63694). The December 26, 2002, supplemental letter provided clarifying 
information that was within the scope of the initial notice and did not 
change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 4, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: May 31, 2002, as supplemented 
by letter dated October 16, 2002.
    Brief description of amendments: These amendments revised Technical 
Specifications (TSs) 3.8.2.1, ``DC Sources--Operating,'' and 3.8.2.2, 
``DC Sources--Shutdown''; and added the new Specification 6.8.4.i, 
``Battery Monitoring and Maintenance Program.'' The changes also 
included the relocation of the following TS items to a licensee-
controlled program: (1) A number of surveillance requirements that 
require the performance of preventive maintenance, and (2) certain 
battery and battery cell parameter values that are periodically 
verified to monitor early indications of DC subsystem degradation.
    Date of issuance: January 29, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 164 and 126.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58643). The supplement dated October 16, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 29, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: February 28, 2001, as 
supplemented by letter dated June 13, 2002.
    Brief description of amendments: Revise the Technical 
Specifications to eliminate the requirement for at least one person 
qualified to stand watch to be present in the control room when nuclear 
fuel is stored in the spent fuel pool.
    Date of issuance: January 31, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: 183 and 170.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34283).
    The June 13, 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 amendments is 
contained in a Safety Evaluation dated January 31, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania

    Date of application for amendment: May 31, 2002, as supplemented 
July 19, and September 3, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.1.1.4, upper limit for the moderator temperature 
coefficient (MTC), from 0 x 10-4 change in reactivity per 
degree Fahrenheit ([Delta]k/k/[deg]F) to +0.2 x 10-4 
[Delta]k/k/[deg]F for power

[[Page 7826]]

levels up to 70 percent of rated thermal power (RTP), and ramping 
linearly to 0 x 10-4 [Delta]k/k/[deg]F from 70 percent to 
100 percent RTP. The change is needed to address future core designs 
with higher energy requirements, associated with plant operation at 
higher capacity factors.
    Date of issuance: February 6, 2003.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 251.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58644). The July 19, and September 3, 2002, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated February 6, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 27, 2002, as supplemented 
on October 7, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specifications section 3.6.3, ``Emergency Power Sources,'' to extend 
the current allowable outage time for an inoperable diesel generator 
from 7 days to 14 days, and section 3.4.4, ``Emergency Ventilation 
System,'' and section 3.4.5, ``Control Room Air Treatment System,'' to 
reflect the change to section 3.6.3.
    Date of issuance: February 3, 2003.
    Effective date: February 3, 2003.
    Amendment No.: 179.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21290). The October 7, 2002, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 3, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 25, 2002. The application 
was initially submitted to the Nuclear Regulatory Commission with an 
incorrect date of April 25, 2001. The Nuclear Management Company, LLC, 
subsequently submitted a letter dated May 30, 2002, correcting the date 
of the application as April 25, 2002.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 3.7/4.7, ``Containment Systems,'' to allow the use 
of 10 CFR part 50, Appendix J, Option B, for Types B and C containment 
leak rate testing and adds a new TS section 6.8.M, ``Programs and 
Manuals--Primary Containment Leakage Rate Testing Program.''
    Date of issuance: February 4, 2003.
    Effective date: As of the date of issuance and to be implemented 
within 75 days.
    Amendment No.: 132.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56325).
    The May 30, 2002, letter corrected the date of the application and 
did not change the NRC staff's initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 4, 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: July 25, 2002, as supplemented 
by letter dated October 23, 2002.
    Brief description of amendments: These amendments revise the 
Susquehanna Steam Electric Station Final Safety Analysis Report (SSES 
FSAR) by replacing the current plant-specific reactor pressure vessel 
material surveillance program with the Boiling Water Reactor Integrated 
Surveillance Program.
    Date of issuance: February 6, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 208 and 182.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the SSES FSAR.
    Date of initial notice in Federal Register: September 3, 2002 (67 
FR 56328). The October 23, 2002, supplemental letter provided 
additional information that clarified the application, but 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 amendments is contained 
in a Safety Evaluation dated February 6, 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: October 23, 2002.
    Brief description of amendment: The amendment updates the reference 
to 10 CFR 20.203 with the corresponding reference to 10 CFR 20.1601. 
Hope Creek Generating Station Technical Specification (TS) 6.12, ``High 
Radiation Area,'' is revised to be consistent with the Standard TSs, 
General Electric Plants (NUREG-1433, Rev. 2).
    Date of issuance: January 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 142.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75884).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 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: August 21, 2002.
    Brief description of amendments: The amendments revise the 
technical specifications (TSs) to replace reference to specific valves 
for preventing uncontrolled boron dilution. The revised TSs incorporate 
a general statement for preventing uncontrolled boron dilution, 
consistent with the improved standard TSs.
    Date of issuance: January 27, 2003.
    Effective date: January 27, 2003.
    Amendment Nos.: Unit 1-149; Unit 2-137.

[[Page 7827]]

    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61686).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 27, 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 23, 2002.
    Brief description of amendments: The amendments relocated the 
shutdown margin limits to the Core Operating Limits Report and modified 
certain boration requirements consistent with NUREG-1431. The 
amendments also correct some typographical errors in the Technical 
Specification pages.
    Date of issuance: February 4, 2003.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-150; Unit 2-138.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42830).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 4, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 6, 2002.
    Brief description of amendments: The amendments revised the Browns 
Ferry Nuclear Plant, Units 2 and 3, Updated Final Safety Analysis 
Report (UFSAR) to modify the basis for TVA's compliance with the 
requirements of Appendix H to title 10 of the Code of Federal 
Regulations part 50, ``Reactor Vessel Material Surveillance Program 
Requirements.''
    Date of issuance: January 28, 2003.
    Effective date: As of the date of issuance, to be incorporated into 
the UFSAR at the time of its next update.
    Amendment Nos.: 279 & 238.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the UFSAR.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70770). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 28, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 3, 2002, as 
supplemented October 17, 2002, and January 29, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the 
delay period, before entering a Limiting Condition for Operation, 
following a missed surveillance. Thise changes to SR 4.0.3 will allow 
an extension of up to 24 hours or the limit of the surveillance 
frequency, whichever is greater. The amendments also include editorial 
changes to make the revised TS consistent with the Standard TS for 
Westinghouse plants. In addition, the amendments include the adoption 
of the TS Bases Control Program listed in NUREG-1431, Revision 2.
    Date of issuance: February 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 280 and 271.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68745). The January 29, 2003, supplemental letter provided 
clarifying information that was within the scope of the initial notice 
and did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 5, 2003.
    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 10th day of February, 2003.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-3689 Filed 2-13-03; 8:45 am]
BILLING CODE 7590-01-P