[Federal Register Volume 68, Number 4 (Tuesday, January 7, 2003)]
[Notices]
[Pages 798-816]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-156]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, December 13, through December 26, 2002. The 
last biweekly notice was published on December 24, 2002 (67 FR 78515).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. The filing of requests for a 
hearing and petitions for leave to intervene is discussed below.
    By February 6, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714,\1\ which is 
available at the Commission's PDR, located at One White Flint North, 
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

[[Page 799]]

notice of a hearing or an appropriate order.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2) 
regarding petitions to intervene and contentions. For the complete, 
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April 
29, 2002.
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

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 amendment request: November 13, 2002, as supplemented 
November 20, 2002
    Description of amendment request: The proposed amendments delete 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post-Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the TS for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means or is of little use in the assessment and 
mitigation of accident conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post-Accident Sampling System 
(PASS).'' The NRC staff issued a notice of opportunity for comment in 
the Federal Register on December 27, 2001 (66 FR 66949), on possible 
amendments

[[Page 800]]

concerning TSTF-413, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on March 20, 
2002 (67 FR 13027). The licensee affirmed the applicability of the 
following NSHC determination in its application dated November 13, 
2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: 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. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Located in Mecklenburg County, North Carolina

    Date of amendment request: December 2, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for Administrative Controls in 
Section 5.0 concerning Responsibility, Unit Staff, Unit Staff 
Qualifications, and Controls of the High Radiation Area.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    As required by 10 CFR 50.91(a)(1), this analysis is provided to 
demonstrate that the proposed license amendment does not involve a 
significant hazard.
    Conformance of the proposed amendment to the standards for a 
determination of no significant hazards, as defined in 10 CFR 50.92, 
is shown in the following:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. Implementation of this amendment would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Approval of this amendment will have 
no effect on accident probabilities or consequences since the 
changes are purely administrative in nature.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. Implementation of this amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No physical 
changes are being made to the plant. Therefore, the introduction of 
any new accident scenarios does not exist. The amendment does not 
impact any plant systems that are accident initiators nor does it 
adversely impact any accident mitigating system. This amendment is 
purely administrative in nature.
    (3) Does the proposed change involve a significant reduction in 
margin of safety?
    No. Implementation of this amendment will not involve a 
significant reduction in a margin of safety. Margin of safety is 
related to the confidence in the ability of the fission product 
barriers to perform their design functions during and following an 
accident situation. These barriers include the fuel cladding, the 
reactor coolant system, and the containment system. The performance 
of these fission product barriers will not be impacted by 
implementation of this amendment. System[s] and components are not 
affected and therefore are capable of performing as designed. This 
amendment is purely administrative nature, it will have no effect on 
any safety margins.
    Conclusion.
    Based on the preceding analysis, it is concluded that the 
proposed license amendment does not involve a Significant

[[Page 801]]

Hazards Consideration Finding as defined in 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: December 12, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for TS Table 3.3.2-1 Footnote 
(c) to correct an editorial error, TS 3.4.3 is revised to update the 
Reactor Coolant System Pressure-Temperature limits for use up to 34 
Effective Full Power Years (EFPY) and TS 3.4.12 is revised to update 
the Low Temperature Over-Pressure limits for use up to 34 EFPY. 
Associated changes are also proposed for the TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Duke has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendments by focusing 
on the three standards set forth in 10 CFR 50.92, ``issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the reactor coolant system (RCS) 
pressure and temperature (P-T) limits and low temperature 
overpressure protection (LTOP) limits are developed utilizing the 
methodology of American Society of Mechanical Engineers (ASME) 
Section XI, Appendix G, in conjunction with the methodology of ASME 
Code Case N-641. Usage of these methodologies provides compliance 
with the underlying intent of 10 CFR [Part] 50 Appendix G and 
provides operational limits established to prevent non-ductile 
failure of the reactor vessel. The Loss of Coolant Accident analysis 
and other accident analyses in the Updated Final Safety Analysis 
Report (UFSAR) do not assume failure of the reactor vessel. The P-T 
and LTOP limits are not initiators or contributors to accident 
analyses addressed in the UFSAR. The proposed changes do not alter 
any assumption previously made in the radiological consequence 
evaluations nor affect the mitigation of the radiological 
consequences of an accident previously evaluated. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The changes to RCS P-T limits and LTOP limits are proposed to 
prevent non-ductile failure of the reactor vessel. The proposed 
changes do not modify the RCS pressure boundary, nor make any 
physical changes to the facility. The proposed changes do not 
introduce any new mode of system operation or failure mechanism. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are developed utilizing the methodology of 
ASME Section XI, Appendix G, in conjunction with the methodology of 
ASME Code Case N-461. Usage of these methodologies provides 
compliance with the underlying intent of 10 CFR [Part] 50 Appendix G 
and provides operational limits established to prevent non-ductile 
failure of the reactor vessel. This Code case constitutes relaxation 
from the current requirements of 10 CFR [Part] 50 Appendix G. The 
alternate methodology allowed by the Code case is based on industry 
experience gained since the inception of the 10 CFR [Part] 50 
Appendix G requirements and replaces some requirements that have now 
been determined to be excessively conservative. The more appropriate 
assumptions and provisions allowed by the Code case maintain a 
margin of safety that is consistent with the intent of 10 CFR [Part] 
50 Appendix G. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    Based on the above, Duke concludes that the proposed amendments 
present no significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: December 12, 2002.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating License and Technical Specifications 
(TSs) to increase the licensed core thermal power level to 3114.4 
megawatts (MWt), which is a 1.4% increase above the currently 
authorized power level of 3071.4 MWt. The proposed power uprate 
involves the improvement in the core power uncertainty allowance 
originally required for the emergency core cooling system (ECCS) 
evaluations performed in accordance with Appendix K, ``ECCS Evaluation 
Models,'' to Part 50 of Title 10 of the Code of Federal Regulations. In 
addition, changes would be made in TS Sections 1.1, 2.1, 2.3, 3.1, 3.4, 
6.9, and the applicable TS Bases would be revised to account for the 
change in power level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed 1.4% increase in maximum core thermal power is 
based on the use of instrumentation that supports a reduction in the 
measurement uncertainty value assumed in certain safety analyses. 
The affected analyses now use an uncertainty value of 2% which was 
required by 10 CFR [Part] 50 Appendix K at the time that the plant 
was originally licensed. At that time, measurement of feedwater 
flowrate in the plant secondary side used differential pressure-type 
flow venturis. The plant secondary side thermal calorimetric is used 
to determine reactor thermal power. A June 2000 revision to 10 CFR 
[Part] 50 Appendix K permitted the use of lower uncertainty values 
in the affected analyses, if the reduced value can be justified. 
Entergy Nuclear Operations (ENO) has implemented the use of Caldon, 
Inc. Leading Edge Flowmeter (LEFM) technology to measure feedwater 
flowrate. The LEFM measures fluid velocity by measuring the transit 
time of ultrasonic pulses introduced into the fluid stream. The LEFM 
Check System implemented at Indian Point 2 has a demonstrated 
measurement accuracy of 0.6%. Based on this measurement accuracy, 
the licensed thermal power can be increased 1.4% by reducing the 
assumed uncertainty used in safety analyses

[[Page 802]]

with respect to core thermal power from 2.0% to 0.6%. This results 
in a net increase in licensed reactor core thermal power; from 
3071.4 MWt to 3114.4 MWt. The LEFM and the flow venturi 
instrumentation are used to collect data and there is no automatic 
initiation function performed by this instrumentation. Use of the 
LEFM instrumentation is therefore not an accident initiator and does 
not increase the probability of occurrence of an existing analyzed 
accident. Also, the LEFM instrumentation and the venturi 
instrumentation do not mitigate accidents so that the consequences 
of previously analyzed accidents are not increased.
    Analyses and evaluations associated with the proposed change to 
core thermal power have demonstrated that applicable acceptance 
criteria for plant systems, components, and analyses (including the 
Final Safety Analysis Report [FSAR] Chapter 14 safety analyses) will 
continue to be met for the proposed 1.4% increase in licensed core 
thermal power for Indian Point 2. The subject increase in core 
thermal power will not result in conditions that could adversely 
affect the integrity (material, design, and construction standards) 
or the operational performance of any potentially affected system, 
component or analysis. Therefore, the probability of an accident 
previously evaluated is not affected by this change. The subject 
increase in core thermal power will not adversely affect the ability 
of any safety-related system to meet its intended safety function. 
Further, the radiological dose evaluations in support of this power 
uprate effort show that the current FSAR Chapter 14 radiological 
analyses are unaffected, and that the current dose analyses of 
record bound plant operation with the subject increase in licensed 
core thermal power level.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed license amendment increases the maximum allowed 
core thermal power through the use of feedwater flow instrumentation 
that supports a reduction in the measurement uncertainty assumed in 
certain safety analyses. The LEFM Check System instrumentation has 
greater measurement accuracy than the differential pressure-type 
flow venturi instrumentation that was originally used so that the 
measurement uncertainty assumed in certain analyses can be 
correspondingly reduced. Both the venturi and LEFM flow 
instrumentation provide data that is used by plant operators to 
monitor the thermal output of the plant. The instrumentation does 
not perform an automatic actuation function and there are no output 
signals to plant safety systems or control systems. Therefore, 
instrumentation malfunction or failure does not introduce new 
accident scenarios or equipment failure mechanisms. Operation, 
maintenance, or failure of either instrumentation system does not 
have an adverse effect on safety-related systems or any structures, 
systems, and components required for transient or accident 
mitigation.
    Operating the plant at a new maximum core thermal power of 
3114.4 MWt, which is 1.4% greater than the current maximum of 3071.4 
MWt, is bounded by existing or updated analyses which demonstrate 
that established limits and acceptance criteria continue to be met. 
Operating at the new power level does not create new or different 
accident initiators and existing credible malfunctions are bounded 
by existing or updated analyses or evaluations.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The evaluations and analyses associated with the proposed 
increase in maximum core thermal power demonstrate that applicable 
acceptance criteria will continue to be met. The existing licensed 
maximum core thermal power level incorporates a 2% measurement 
uncertainty for the analysis of loss-of-coolant-accidents as 
originally required by Appendix K of 10 CFR [Part] 50. The 
regulations have subsequently been revised to allow the option of 
justifying smaller measurement uncertainties by using more accurate 
instrumentation to calculate reactor thermal power. Certain analyses 
that already assume a bounding core power level because of the 2% 
measurement uncertainty are not changed as a result of the proposed 
increase in core thermal power. Use of the LEFM instrumentation with 
improved measurement accuracy supports the use of a smaller 
measurement uncertainty assumption in the safety analyses. Other 
analyses were updated or evaluations were performed to demonstrate 
that nuclear steam supply and balance-of-plant systems and 
components will continue to perform, under normal and credible 
transient conditions, within established limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: November 21, 2002.
    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Limerick Generating Station 
(LGS), Unit 2, Technical Specifications (TSs) contained in Appendix A 
to the Operating License. This proposed change will revise the TS 
section on safety limits to incorporate revised safety limit minimum 
critical power ratios (SLMCPRs) due to the cycle-specific analysis 
performed by Global Nuclear Fuel for LGS, Unit 2, Cycle 8.
    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 derivation of the cycle specific Safety Limit Minimum 
Critical Power Ratios (SLMCPRs) for incorporation into the Technical 
Specifications (TS), and their use to determine cycle specific 
thermal limits, has been performed using the methodology discussed 
in ``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US, 
June, 2000, which incorporates Amendment 25. Amendment 25 was 
approved by the NRC [Nuclear Regulatory Commission] in a March 11, 
1999 safety evaluation report.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling. The GE-14 fuel is in compliance with 
Amendment 22 to ``General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing 
acceptance criteria. The probability of fuel damage will not be 
increased as a result of this change. Therefore, the proposed TS 
change does not involve a significant increase in the probability or 
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 SLMCPR is a TS numerical value, calculated to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core if the limit is not violated. The new SLMCPRs are calculated 
using NRC approved methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. Additionally, the GE-14 fuel is in 
compliance with Amendment 22 to ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and 
U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which provides the 
fuel licensing acceptance criteria. The SLMCPR is

[[Page 803]]

not an accident initiator, and its revision will 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.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of the proposed change to 
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs 
are calculated using methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which 
incorporates Amendment 25. The SLMCPRs ensure that greater than 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated when all uncertainties are considered, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed TS change will not involve a significant reduction in the 
margin of safety previously approved by the NRC.

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

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

    Date of amendment request: November 26, 2002.
    Description of amendment request: The proposed amendments revise 
Technical Specification (TS) 3.1.3.1, Control Rod Operability,'' by 
adding required actions for scram discharge volume (SDV) vent and drain 
valves to align with those in NUREG-1433, ``Standard Technical 
Specification, General Electric Plants, BWR/4,'' Revision 2. 
Additionally, modifications are proposed to change TS 3.6.3, ``Primary 
Containment Isolation Valves,'' to clarify the relationship between TS 
3.1.3.1 and TS 3.6.3 regarding SDV vent and drain valve.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The scram discharge volume (SDV) and control rod drive (CRD) 
system, including the associated SDV vent and drain isolation 
valves, are not initiators to any accident sequence analyzed in the 
Updated Final Safety Analysis Report (UFSAR). Operation in 
accordance with the proposed Technical Specification (TS) ensures 
that the SDV and control rods are capable of performing their 
function as described in the UFSAR; therefore, the mitigative 
functions supported by the SDV and control rods will continue to 
provide the protection assumed by the analysis. The addition of 
specific TS actions to be taken for inoperable SDV vent or drain 
isolation valves will not challenge the ability of the SDV and 
control rods to perform their design function. Appropriate 
monitoring and maintenance, consistent with industry standards, will 
continue to be performed. In addition, the CRD system including the 
SDV isolation valves is within the scope of 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' which will ensure the control of maintenance 
activities associated with the CRD system and SDV isolation valves.
    Under the proposed TS changes, the SDV vent and drain lines may 
be unisolated under administrative control. This allows any 
accumulated water in the line to be drained, to preclude a reactor 
scram on SDV high level. This is acceptable since the administrative 
controls ensure the valve can be closed quickly, by a dedicated 
operator, if a scram occurs with the valve open. The 8-hour 
allowable outage time to isolate the line is based on the low 
probability of a scram occurring while the line is not isolated and 
unlikelihood of significant CRD seal leakage.
    The proposed changes do not involve any physical change to 
structures, systems, or components (SSCs) and do not alter the 
method of operation or control of SSCs. The current assumptions in 
the safety analysis regarding accident initiators and mitigation of 
accidents are unaffected by these proposed changes. No additional 
failure modes or mechanisms are being introduced and the likelihood 
of previously analyzed failures remains unchanged.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by these proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase because of these 
proposed changes.
    Based on the above discussion, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner. There are no 
setpoints, at which protective or mitigative actions are initiated, 
affected by these proposed changes. These proposed changes will not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. Any 
alteration in procedures will continue to ensure that the plant 
remains within analyzed limits, and no change is required to the 
procedures relied upon to respond to an off-normal event as 
described in the UFSAR. As such, no new failure modes are being 
introduced. The changes do not alter assumptions made in the safety 
analysis and licensing basis.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes are acceptable because the 
operability of the SDV and SDV isolation valves is unaffected, there 
is no detrimental impact on any equipment design parameter, and the 
plant will still be required to operate within assumed conditions. 
Operation in accordance with the proposed TS ensures that the SDV 
and control rods are capable of performing their functions as 
described in the UFSAR. Therefore, the support of the SDV and 
control rods in the plant response to analyzed events will continue 
to provide the margins of safety assumed by the analysis. The 
additions to TS for inoperable SDV vent and drain isolation valves 
will not challenge the ability of the SDV or control rods to perform 
their design function. Appropriate monitoring and maintenance, 
consistent with industry standards, will continue to be performed. 
In addition, CRD system, including the SDV vent and drain isolation 
valves, are within the scope of 10 CFR 50.65, ``Requirements for 
monitoring the effectiveness of maintenance at nuclear power 
plants,'' which will ensure the control of maintenance activities 
associated with the CRD system. This provides sufficient management 
control of the requirements that assure the control rods and CRD 
system are maintained in a highly reliable condition. Although there 
is an increase in allowable outage time, this increase was evaluated 
and determined not to be a significant reduction in a margin of 
safety.
    The proposed TS Actions for inoperable SDV vent and drain 
isolation valves are reasonable and consistent with approved 
standards, guidance and regulations.
    Based on the above discussion, the proposed TS 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.

[[Page 804]]

    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Andersen.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: June 4, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) Surveillance Requirement (SR) 
4.0.3 to extend the delay period, before entering a Limiting Condition 
for Operation, following a missed surveillance. The delay period would 
be extended from ``* * * up to 24 hours to permit completion of the 
surveillance when the allowable (equipment inoperability) outage time 
limits of the ACTION requirements are less than 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.'' 
The proposed amendment is consistent with TS Task Force traveler TSTF-
358, which has been approved by the Nuclear Regulatory Commission 
(NRC). The TS Bases will be revised under the licensee's existing TS 
Bases control program to be consistent with the bases for TSTF-358.
    Basis for proposed no significant hazards consideration 
determination: The NRC staff issued a notice of opportunity for comment 
in the Federal Register on June 14, 2001 (66 FR 32400), on possible 
amendments concerning missed surveillances, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on September 28, 2001 (66 FR 49714). The licensee reviewed the 
model NSHC presented in the Federal Register and concluded that it is 
applicable to Davis-Besse. The model NSHC determination was 
incorporated by reference into its application dated June 4, 2002, to 
satisfy the requirements of 10 CFR 50.91(a), and is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function. Therefore, this change does 
not involve a significant reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 9, 2002.
    Description of amendment request: The proposed amendment utilizes 
the Alternate Source Term radiological calculations to update the 
design basis analysis in the Updated Safety Analysis Report for the 
Fuel Handling Accident. Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors,'' was utilized in the development of the 
proposed amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. This proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment involves implementation of the 
Alternative Source Term for the Fuel Handling Accident at the Perry 
Nuclear Power Plant (PNPP). There are no physical design 
modifications to the plant associated with the proposed amendment. 
The revised calculations do not impact the initiators of a Fuel 
Handling Accident in any way. They also do not impact the initiators 
for any other design basis events. Therefore, because design basis 
accident initiators are not being altered by adoption of the 
Alternative Source Term analyses, the

[[Page 805]]

probability of an accident previously evaluated is not affected.
    With respect to consequences, the only previously evaluated 
accident that could be affected is the Fuel Handling Accident. The 
Alternative Source Term is an input to calculations used to evaluate 
the consequences of an accident, and does not by itself affect the 
plant response, or the actual pathway of the radiation released from 
the fuel. It does however, better represent the physical 
characteristics of the release, so that appropriate mitigation 
techniques may be applied. For the Fuel Handling Accident, the AST 
analyses demonstrate acceptable doses, within regulatory limits, 
after 24 hours of radiological decay, without credit for 
Containment/Fuel Handling Building integrity, filtration system 
operability, or Control Room automatic isolation. Therefore, the 
consequences of an accident previously evaluated are not 
significantly increased.
    Based on the above conclusions, this proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. This proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes). Also, no changes are proposed 
to the methods governing plant/system operation during handling of 
recently irradiated fuel, so no new initiators or precursors of a 
new or different kind of accident are created. New equipment or 
personnel failure modes that might initiate a new type of accident 
are not created as a result of the proposed amendment.
    Thus, this amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. This proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment is associated with the implementation of 
a new licensing basis for PNPP Fuel Handling Accidents. Approval of 
the change from the original source term to a new source term taken 
from Regulatory Guide 1.183 is being requested. The results of the 
accident analyses, revised in support of the proposed license 
amendment, are subject to revised acceptance criteria. The analyses 
have been performed using conservative methodologies, as specified 
in Regulatory Guide 1.183. Safety margins have been evaluated and 
analytical conservatism has been utilized to ensure that the 
analyses adequately bound the postulated limiting event scenario. 
The dose consequences of the limiting Fuel Handling Accident remains 
within the acceptance criteria presented in 10 CFR 50.67, ``Accident 
Source Term,'' and Regulatory Guide 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries, as well as the 
Control Room, are within corresponding regulatory limits. For the 
Fuel Handling Accident, Regulatory Guide 1.183 conservatively sets 
the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) 
limits below the 10 CFR 50.67 limit, and sets the Control Room limit 
consistent with 10 CFR 50.67.
    Since the proposed amendment continues to ensure the doses at 
the EAB, LPZ and Control Room are within corresponding regulatory 
limits, the proposed license amendment does not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 23, 2002.
    Description of amendment request: The proposed amendment would 
revise Crystal River Unit 3 Improved Technical Specifications (ITS) 
4.2.1, ``Fuel Assemblies,'' and ITS 4.2.2, ``Control Rods,'' to permit 
the use of Framatome ANP M5 advanced alloy for fuel rod cladding and 
fuel assembly structural components.
    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:

    Florida Power Corporation (FPC) has evaluated the proposed 
License Amendment Request (LAR), which consists of the identified 
Technical Specification changes and exemption requests, against the 
criteria of 10 CFR 50.92(c). The Technical Specification changes are 
categorized as follows:
    1. Modification of Section 4.2.1, DESIGN FEATURES, Fuel 
Assemblies, and to include the M5 advanced alloy for fuel rod 
cladding and fuel assembly structural material[.]
    2. Removal of design information such as maximum fuel 
enrichment, nominal active fuel length, maximum individual rod 
weight, and details of Control Rod content. Adopting the wording 
from the Standard ITS.
    3. Addition to ITS 4.2.1 of the following sentence: ``A limited 
number of lead test assemblies that have not completed 
representative testing may be placed in nonlimiting core regions.'' 
Crystal River Unit 3 does not intend to load lead test assemblies in 
the upcoming fuel cycle (Cycle 14). This sentence is being added for 
consistency with NUREG 1430, Revision 2.
    FPC has concluded that this proposed LAR does not involve a 
significant hazards consideration. The following is a discussion of 
how each of the criteria is satisfied.
    (1) [Does not] [i]nvolve a significant increase in the 
probability or consequences of an accident previously evaluated.
    M5 advanced alloy: Topical reports BAW-10227P-A, ``Evaluation of 
Advanced Cladding and Structural Material (M5) in PWR [Pressurized 
Water Reactor] Reactor Fuel,'' February 2000 and BAW-10179P-A, 
Revision 4, ``Safety Criteria and Methodology for Acceptable Cycle 
Reload Analyses,'' March 2001 provide the licensing basis for the 
Framatome ANP (FRA-ANP) advanced cladding and structural material, 
designated M5. The M5 material can be used for fuel rod cladding, as 
well as for fuel assembly spacer grids, fuel rod end plugs, and fuel 
assembly guide and instrument tubes. By letter dated August 2, 2001 
(Reference 4), the NRC approved BAW-10179P-A, Revision 4, for 
referencing in license applications. BAW-10179P-A, Revision 4 
incorporates BAW-10227P-A. The M5 material was shown in these 
documents to have equivalent or superior properties to the current 
Zircaloy-4 material. The cladding itself is not an accident 
initiator and does not affect accident probability. The M5 cladding 
has been shown to meet all 10 CFR 50.46 design criteria and, 
therefore, will not increase the consequences of an accident.
    Removal of design parameters of maximum fuel enrichment, active 
fuel length, rod weight and Control Rod content: This change moves 
design features from Improved Technical Specifications (ITS) to the 
Final Safety Analysis Report (FSAR) and other design documents and 
analyses. The Framatome ANP enhanced fuel design will involve 
increased rod weight and active fuel length. The approved Framatome 
ANP topical report, BAW-10179P-A, ``Safety Criteria and Methodology 
for Acceptable Cycle Reload Analyses,'' will continue to be used to 
ensure that the required safety limits for the fuel are satisfied. 
Therefore, the relocation of design information does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Addition of a limited number of lead test assemblies: This 
change is administrative in nature and is proposed for consistency 
with the ITS standard. Crystal River Unit 3 does not intend to load 
lead test assemblies in the upcoming fuel cycle. When lead test 
assemblies are to be loaded, the approved Framatome ANP topical 
report BAW-10179P-A will be used to ensure that all applicable 
limits of the safety analysis are met and that the lead test 
assemblies are placed in nonlimiting core locations. Applicable 
mixed core penalties and core operating limits will be developed and 
applied. Therefore, use of lead test assemblies will not involve a 
significant

[[Page 806]]

increase in the probability or consequences of an accident 
previously evaluated.
    (2) [Does not] [c]reate the possibility of a new or different 
kind of accident from any accident previously evaluated.
    M5 advanced alloy: Topical report BAW-10227P-A demonstrated that 
the material properties of the M5 alloy are not significantly 
different from those of Zircaloy-4. Therefore, M5 fuel rod cladding 
and fuel assembly structural components will perform similarly to 
those fabricated from Zircaloy-4, thus precluding the possibility of 
the fuel becoming an accident initiator and causing a new or 
different type of accident.
    Removal of design parameters of maximum fuel enrichment, active 
fuel length, rod weight and Control Rod content: This change moves 
design features from ITS to the FSAR and other design documents and 
analyses or adds consistency with the standard ITS. The location of 
this information does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The approved FRA-ANP topical report, BAW-10179P-A will continue to 
be used to ensure that the required safety limits are satisfied. 
Therefore, these changes do not involve the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Addition of a limited number of lead test assemblies: This 
change is administrative in nature and it is proposed for 
consistency with the ITS standard. Crystal River Unit 3 does not 
intend to load lead test assemblies in the upcoming fuel cycle. When 
lead test assemblies are to be loaded, they will be designed and 
manufactured to ensure compatibility with the co-resident fuel 
assemblies, core internal structures, and fuel handling and storage 
equipment. The approved Framatome ANP topical report BAW-10179P-A 
will be used to ensure that the lead test assemblies meet all 
applicable limits of the safety analysis and that the lead test 
assemblies are placed in non-limiting core locations. Applicable 
mixed core penalties and core operating limits will be developed and 
applied. Therefore, use of lead test assemblies will not involve the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) [Does not] [i]nvolve a significant reduction in a margin of 
safety.
    M5 advanced alloy: The proposed changes will not involve a 
significant reduction in the margin of safety because it has been 
demonstrated that the material properties of the M5 alloy are not 
significantly different from those of Zircaloy-4. The M5 alloy is 
expected to perform similarly or better [than] Zircaloy-4 for all 
normal operating and accident scenarios, including both non-LOCA 
[loss-of-coolant accident] and LOCA scenarios. For LOCA scenarios, 
where the slight differences in M5 material properties relative to 
Zircaloy-4 could have some impact on the overall accident scenario, 
plant-specific LOCA analyses will be performed prior to the use of 
fuel assemblies with fuel rods or fuel assembly components 
containing M5. These LOCA analyses, required by ITS 5.6.2.18, ``Core 
Operating Limits Report (COLR),'' will demonstrate that all 
applicable margins of safety will be maintained by the use of the M5 
alloy.
    Removal of design parameters of maximum fuel enrichment, active 
fuel length, rod weight and Control Rod content: Approved 
methodologies will be used in the cycle-specific safety analysis to 
evaluate the use of the M5 advanced alloy, and account for various 
assembly differences (various rod weights and active fuel lengths). 
The location of the design information does not affect the margin of 
safety.
    Addition of a limited number of lead test assemblies: This 
change is administrative in nature and is proposed for consistency 
with the ITS standard. Crystal River Unit 3 does not intend to load 
lead test assemblies in the upcoming fuel cycle. When lead test 
assemblies are to be loaded, the approved Framatome ANP topical 
report BAW-10179P-A will be used to ensure that all applicable 
limits of the safety analysis are met and that the lead test 
assemblies are placed in nonlimiting core locations. Applicable 
mixed core penalties and core operating limits will be developed and 
applied. There will be no significant reduction in the margin of 
safety when a limited number of lead test assemblies are utilized.

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

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: November 25, 2002.
    Description of amendment request: The proposed license amendment 
would modify plant Technical Specifications (TSs) and the associated 
spent fuel pool (SFP) criticality analyses to eliminate credit for the 
BoraflexTM neutron absorber in SFP fuel storage racks and 
credit specific rules to control fuel assembly positioning in the SFP 
racks. TS 3.9.11 is revised to add a Limiting Condition for Operation 
for the SFP soluble boron concentration and require periodic 
surveillance of this parameter. This submittal provides justification 
for removing the description of the poison material in the spent fuel 
racks from Section 5 of the Unit 1 TSs, that was requested to be added 
by the licensee's cask pit spent fuel storage rack submittal dated 
October 23, 2002. In addition, a new SFP dilution analysis was 
performed that supports the criticality analysis requirement for a 
minimum soluble boron concentration.
    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 operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    No. The proposed amendment to eliminate reliance on 
BoraflexTM and to credit SFP soluble boron for reactivity 
control in the spent fuel pool storage racks was evaluated for 
impact on the following previously evaluated events:
    [sbull] A fuel handling accident (FHA)
    [sbull] A fuel mispositioning event
    [sbull] A cask drop accident
    [sbull] A loss of spent fuel pool cooling
    The proposed amendment does not modify the facility. A new 
criticality analysis credits existing soluble boron in the SFP water 
and specific fuel positioning rules for reactivity control, without 
requiring any physical changes to the fuel storage racks. The 
amendment does not change any rack module location or any module's 
designation as Region 1 or Region 2 storage. There is no significant 
increase in the probability of a fuel handling accident in the SFP 
that is caused by crediting soluble boron and new fuel positioning 
rules, rather than BoraflexTM, for reactivity control. 
The probability of a fuel handling accident is a function of the 
equipment design and procedures used when handling irradiated fuel. 
Neither of these features is affected when soluble boron, instead of 
BoraflexTM, is credited for reactivity control in the 
SFP.
    There is no increase in the probability of an accidental fuel 
assembly mispositioning when crediting the presence of soluble boron 
in fuel pool water for reactivity control. Fuel assembly selection 
and manipulation will continue to be controlled by approved fuel 
handling procedures; these procedures require the identification of 
a verified target location prior to grappling the assembly. Fuel 
placement will be in accordance with the revised TS.
    There is no increase in the consequences of either an FHA or an 
accidental mispositioning of a fuel assembly into the SFP racks. 
Consequences of a FHA are not increased because the proposed 
amendment does not change the fuel fission product inventory, local 
meteorological conditions, or the fission product partition factor 
provided by fuel pool water. The consequences of an accidental 
misload are not increased because the criticality analysis 
demonstrates that the fuel array will remain sub-critical, even if 
the pool contains a boron concentration below the minimum level 
required by Technical Specifications. The TS will ensure that an 
adequate SFP soluble boron concentration is maintained for all 
conditions.
    The proposed fuel positioning rules do not cause the total 
radionuclide inventory present in the spent fuel pool to increase, 
or

[[Page 807]]

alter the type or mass of casks that may be placed in the fuel pool, 
or alter any facet of operation of the spent fuel cask crane. No 
characteristics of the existing spent fuel cask drop analysis for 
Unit 1 are affected by the proposed fuel positioning rules or by 
credit for soluble boron. Therefore, there is no increase in either 
the probability or the consequences of a cask drop accident caused 
by this change.
    The proposed change does not increase either the probability or 
the consequences of a loss of normal SFP cooling. The proposed fuel 
positioning rules do not require any interaction with the fuel pool 
cooling system. Credit for a portion of the existing soluble boron 
concentration does not change its interaction with the fuel pool 
cooling system. The ability to detect and mitigate a loss of SFP 
cooling event is unchanged, and the revised criticality analysis 
considered the effects of boiling in the SFP and found them 
acceptable.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    No. The proposed change does not modify the physical plant, 
nuclear fuel, or the design function and operation of the spent fuel 
pool storage racks at St. Lucie Unit 1. A TS controlled minimum 
concentration of soluble boron has always been required in the St. 
Lucie Unit 1 spent fuel pool; as such, the possibility of an 
inadvertent fuel pool dilution event has always existed. However, 
the spent fuel pool dilution analysis that accompanies this 
submittal demonstrates that no credible dilution event could 
increase fuel pool reactivity such that the effective neutron 
multiplication factor (keff) exceeds 0.95. Therefore, 
implementation of credit for soluble boron to control reactivity in 
the SFP will not create the possibility of a new or different type 
of criticality accident.
    The limiting fuel assembly mispositioning event does not 
represent a new or different type of accident. The mispositioning of 
a fuel assembly within the fuel storage racks has always been 
possible. The locations of SFP rack modules and the specific modules 
assigned to each storage region remain unchanged; analysis results 
show that the storage racks remain subcritical, with substantial 
margin, following a worst case fuel misloading event. Therefore, a 
fuel assembly misload event that involves new fuel storage 
arrangements required by the criticality analysis does not result in 
a new or different type of criticality accident.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    (3) Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    No. The revised fuel positioning requirements proposed by this 
license amendment provide sufficient safety margin to ensure that 
the spent fuel pool storage racks will always remain subcritical. To 
comply with the requirements of 10 CFR 50.68 when crediting soluble 
boron, the current TS reactivity limit for the fuel storage racks 
(i.e., keff less than or equal to 0.95 when flooded with 
unborated water) will be replaced with two separate limits 
(keff less than 1.0 when flooded with unborated water, 
and keff less than or equal to 0.95 when flooded with 
water containing 500 ppm boron).
    The proposed amendment maintains the 0.95 reactivity limit by a 
combination of restrictions on fuel characteristics and fuel 
positioning, storage cell geometry and by crediting a portion of the 
soluble boron in the SFP, rather than by crediting Boraflex.
    The proposed license amendment does not reduce the margin of 
safety provided by the soluble boron normally present in fuel pool 
water; the TS minimum permissible boron concentration is not 
decreased. The TS minimum required value of 1720 ppm is 
substantially greater than the 500 ppm value required by the updated 
criticality analysis to assure keff remains = 0.95 for 
non-accident conditions; it is also substantially greater than the 
soluble boron concentration necessary to compensate at a 95% 
probability, with a 95 percent confidence for the limiting 
postulated reactivity anomaly in the fuel pool storage racks.
    No credible dilution of the fuel pool can result in an SFP 
soluble boron concentration less than the minimum value required by 
the criticality analysis. Therefore, an inadvertent dilution event 
can not challenge safety margins.
    Based on these evaluations and the supporting analyses, 
operating the facility with the proposed amendment does not involve 
in a significant reduction in any margin of safety.

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

GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation 
(SNEC), Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF), 
Bedford County, Pennsylvania

    Date of amendment request: April 22, 2002, as supplemented on 
December 5, 2002.
    Description of amendment request: The proposed amendment would 
allow removal of the upper half of the SNEF containment vessel and make 
a change to the organization to add the position of Vice-President GPU 
Nuclear Oversight to reflect the merger of GPU Inc. and FirstEnergy 
Corp.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    GPU Nuclear has determined that Technical Specification Change 
Request No. 62 involves no significant hazard consideration as 
defined in 10 CFR 50.92.
    1. The proposed changes to the SNEC Technical Specifications do 
not involve a significant increase in the probability of occurrence 
or consequences of an accident or malfunction of equipment important 
to safety previously analyzed in the safety analysis report.
    As described in the change to delete Technical Specification 
1.1.2, radiation levels inside the Containment Vessel will be below 
that necessary to maintain the Containment Vessel as an Exclusion 
Area. Further as required by modified Technical Specification 2.1.1 
ventilation controls will be established to monitor and control any 
potential releases of airborne radioactivity during activities 
involving removal of the upper dome. Finally an analysis has been 
performed to determine the dose to a maximally exposed individual 
due to an accidental release while cutting the Containment Vessel. 
In developing a source term for the event it was assumed that 
following the concrete removal process the interior surfaces of the 
upper Containment Vessel dome was homogeneously coated with concrete 
dust. NUREG 1507 ``Minimum Detectable Concentrations with Typical 
Radiation Survey Instruments for Various Contaminants and Field 
Conditions'' describes an experiment to determine the attenuation 
effects due to dusty conditions. The maximum dust loading presented 
was 9.99 mg/cm2 for soil. This value was converted to 
concrete dust by comparing the relative densities of the material 
(1.5 g/cm3 for soil and 2.3 g/cm3 for 
concrete) or 15.3 mg/cm2. This amount of dust coating the 
internal surfaces of the Containment Vessel dome (9.05E6 
cm2) results in 299 pounds of dust being left in the 
Containment Vessel.
    Table 1 provides the mix of isotopes remaining at the SNEC 
Facility based on the most recent survey results and isotope decay. 
During the removal operation a resuspension factor of 1.9E-2/m (as 
described in NUREG/CR 0130 ``Technology, Safety and Costs of 
Decommissioning a Reference Pressurized Water Reactor Power 
Station'', Volume 2, page J-27) was selected to represent the amount 
of concrete dust going airborne. This parameter is about one order 
of magnitude larger than that used in any other accident analyses 
described in the NUREG. This entire volume of dust was assumed to be 
released, unfiltered, directly to the environment.
    An accident dispersion factor (c/Q) of 3.41E-3 sec/
m3, was also selected as it is the highest, thus most 
conservative, value used in the SNEC Facility Offsite Dose 
Calculation Manual (ODCM). Additionally composite dose conversion 
factors were selected from

[[Page 808]]

Table 5-1 of EPA 400-R-92-001 ``Manual of Protective Action Guides 
and Protective Guides for Nuclear Incidents'' (US EPA, May 1992).
    Based on the above a calculated dose of 3.23E-4 mrem to the 
maximally exposed individual represents a conservative estimate for 
an accidental release. For comparison Section 3.1 of the SNEC 
Facility USAR estimated the dose from an unfiltered release due to a 
material handling event of 1.5 mrem to the maximally exposed 
individual.
    Thus this proposed change does not involve a significant 
increase in the probability of occurrence or consequences of an 
accident or malfunction of equipment important to safety previously 
analyzed in the SNEC Facility USAR.
    For the portions of the amendment that would make a change to 
the organization to add the position of Vice-President GPU Nuclear 
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp, 
these changes are administrative in nature. As such they have no 
effect on the probability of occurrence or consequences of an 
accident or malfunction of equipment important to safety.
    2. The proposed changes to the SNEC Technical Specifications 
will not create the possibility for an accident or malfunction of a 
different type than any previously evaluated in the safety analysis 
report.
    As described in the response to item 1 above, the limiting 
accidental release during segmentation of the Containment Vessel 
dome involves the direct release of radioactive material to the 
environment. This event is similar to both a material handling event 
as described in Section 3.1 of the SNEC Facility USAR, and loss of 
engineering controls during segmentation as described in Section 3.4 
of the SNEC Facility USAR. Thus the possibility of a new accident is 
not created.
    For the portions of the amendment that would make a change to 
the organization to add the position of Vice-President GPU Nuclear 
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp, 
these changes are administrative in nature. As such they have no 
effect on the possibility of an accident or malfunction of a 
different type.
    3. The changes will not involve a significant reduction in the 
margin of safety as defined in the basis for any technical 
specification for SNEC. The SNEC Facility Technical Specifications 
do not contain a defined margin of safety. However the implied 
margin of safety is to protect members of the public from exposure 
to radioactive material.
    At the point in time that these Technical Specifications would 
take affect general radiation levels in the SNEC Facility 
Containment Vessel would be such that the Containment Vessel could 
be opened for unrestricted use as defined in 10 CFR 20.1301. 
Additionally the dose to a maximally exposed individual from an 
accidental release during removal of the Containment Vessel dome is 
several orders of magnitude below that from the limiting accidents 
defined in the SNEC Facility USAR. Thus the margin of safety is not 
reduced.
    For the portions of the amendment that would make a change to 
the organization to add the position of Vice-President GPU Nuclear 
Oversight to reflect the merger of GPU Inc. and FirstEnergy Corp, 
these changes are administrative in nature. As such they have no 
effect on the margin of safety as defined in the basis for any 
technical specification for SNEC.

    The NRC staff has reviewed the analysis of the licensees 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 the Licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Program Director: William D. Beckner.

                                              Table 1.--Maximum Exposed Individual Dose From Cutting the CV
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                 CV concrete
                                activity (Ci)     Fraction      CV wall area       CV air     Instantaneous
           Isotope             per table 4.13   remaining as    concetration    concetration   release rate  Concentration     DCF \7\      Offsite dose
                                 SNEC char.      dust (uCi)      (uCi/m) \2\    (uCi/m) \3\   (uCi/sec) \4\   (uCi/cm) \3\                     (mrem)
                                   report
--------------------------------------------------------------------------------------------------------------------------------------------------------
Am-241.......................  8.24e-05......  4.68e-03......  5.17e-06......  9.83e-08.....  2.93e-04.....  9.99e-13.....  1.47e+05.....  1.47e-04
Co-60........................  4.60e-02......  2.61e+00......  2.89e-03......  5.49e-05.....  1.63e-01.....  5.57e-10.....  7.50e+01.....  4.18e-05
Cs-137.......................  2.38e-01......  1.35e+01......  1.49e-02......  2.84e-04.....  8.46e-01.....  2.88e-09.....  1.14e+01.....  3.28e-05
C-14.........................  5.74e-03......  3.26e-01......  3.60e-04......  6.84e-06.....  2.04e-02.....  6.96e-11.....  6.94e-01.....  4.83e-08
Eu-152.......................  1.42e-03......  8.07e-02......  8.91e-05......  1.69e-06.....  5.05e-03.....  1.72e-11.....  7.50e+01.....  1.29e-06
H-3..........................  1.29e-01......  7.33e+00......  8.10e-03......  1.54e-04.....  4.58e-01.....  1.56e-09.....  2.14e-02.....  3.34e-08
Ni-63........................  3.93e-02......  2.23e+00......  2.47e-03......  4.69e-05.....  1.40e-01.....  4.76e-10.....  2.11e+00.....  1.01e-06
Pu-239.......................  5.24e-05......  2.98e-03......  3.29e-06......  6.25e-08.....  1.86e-04.....  6.35e-13.....  1.44e+05.....  9.17e-05
Pu-241.......................  1.84e-04......  1.05e-02......  1.15e-05......  2.19e-07.....  6.54e-04.....  2.23e-12.....  2.75e+03.....  6.13e-06
Sr-90........................  1.59e-04......  9.03e-03......  9.98e-06......  1.90e-07.....  5.65e-04.....  1.93e-12.....  4.44e+02.....  8.56e-07
                              -----------------
    Total....................  4.60e-01......  2.61e+01......  ..............  .............  1.63e+00.....  .............  .............  2.70e+05
--------------------------------------------------------------------------------------------------------------------------------------------------------
\1\ Fraction remaining determined by: (299 lbs dust/5.26E6 lbs total concrete in CV) x 1E6 uCi/Ci x CV concrete activity.
\2\ Area concentration determined by dividing dust fraction remaining by 9.05E2 m\2\ (surface of CV shell being removed).
\3\ Air concentration determined by multiplying CV wall area activity by 1.9E-2/m (NUREG 0130 resuspension factor for dust sweeping).
\4\ Calculated by multiplying CV air specific activity by CV volume (2.98E3 m\3\) instantaneously released in one second.
\5\ Maximum atmospheric dispersion factor (X/Q) is 3.41E-3 sec/m\3\ at the site boundary (200 meters) and in Sector N per SNEC ODCM Revision 5.
\6\ Calculated by multiplying X/Q x activity released in uCi/sec x 1e-6 m\3\/cm\3\.
\7\ Per EPA 400-R-92-001, Table 5-1.

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

    Date of amendment request: July 10, 2002.
    Description of amendment request: This proposed amendment would 
replace the fire protection (FP) requirements contained in Facility 
Operating License (FOL) Section 2.C.(4) with the standard fire 
protection FOL condition recommended by Generic Letter 86-10, Section 
F, adapted to Cooper Nuclear Station (CNS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change would revise the CNS Operating License 
condition concerning

[[Page 809]]

the FP program and its change process. It does not alter the FP 
requirements in the FHA [fire hazard analysis] or in the USAR 
[updated safety analysis report] including the assumptions 
underlying them. Neither does it alter SSCs [structures, systems or 
components] relied on by analyses to mitigate accidents or special 
events. Since it does not change any of the FP requirements or 
analyses, this proposed amendment does not introduce a new initiator 
for any of the accidents analyzed in the CNS USAR or considered 
therein. Because it does not specifically change any FP requirements 
or mitigating SSCs, this proposed amendment does not introduce a new 
mechanism for degrading the mitigating features considered for the 
accidents analyzed. By introducing no new accident initiators and no 
new mechanisms for degradation of mitigating features, no 
significant increase in the probability or consequences of an 
accident previously evaluated is involved in the proposed change. 
Therefore, the proposed change does not result in a significant 
increase in radiological doses for any Design Basis Accident and 
does not result in a significant increase in the types or amounts of 
any effluents that may be released off-site.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed amendment does not physically change the fit, form, 
or function of any SSC credited in the accident analyses or in the 
FHA, Technical Requirements Manual (TRM), or the USAR. The proposed 
change does not alter assumptions or requirements used in the FHA, 
TRM, or USAR, nor does it affect the CNS Fire Protection program. It 
does not, therefore, alter the FP program or affect the plant's 
ability to achieve and maintain safe shutdown in the event of a 
fire, and it does not result in a reduction in the level of fire 
protection of the facility. Because it does not change FP 
requirements, the FP program or fire-mitigating SSCs, this proposed 
change does not create the possibility of a new or different kind of 
accident from those previously evaluated for CNS.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed amendment does not alter the design features of the 
approved FP plan. The proposed amendment does not alter 
administrative controls in the CNS Fire Protection program necessary 
to ensure required performance of physical barriers during 
anticipated operational occurrences and postulated accidents. The 
proposed change does not alter the NRC approved Fire Protection 
program as described in FP SER [safety evaluation report] dated May 
23, 1979, SER Supplement 1 dated November 21, 1980, SER dated 
September 21, 1983, SER dated April 16, 1984, SER dated August 21, 
1985, SER dated April 10, 1986, SER dated November 7, 1988, SER 
dated August 15, 1995. It does not affect the USAR, the TRM, the FHA 
or the commitments contained therein. It does not physically change 
the fit, form, or function of any SSC credited in the accident 
analyses or in these documents. Because it does not change the 
requirements, plan or mitigating SSCs, this proposed change does not 
involve a significant reduction in a margin of safety.
    In summary, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident or creates the possibility of a new or different kind of 
accident or involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 22, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS), Section 4.6, ``Periodic Testing of Emergency Power System.'' This 
proposed amendment would allow KNPP to inspect the diesel generators 
(DGs) at least once per refueling frequency either while the plant is 
operating or during a refueling outage. Current TS requires an 
inspection during the refueling outage without exception. In addition, 
the proposed amendment would allow KNPP to make administrative changes 
to TS Section 4.6. The proposed change provides operational flexiblity 
in the schedule of maintenance activities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The DGs are accident mitigating equipment, not accident 
initiating equipment. Consequently, there will be no impact on any 
accident probabilities by the approval of the requested amendment.
    The proposed change does not affect the performance of any 
equipment used to mitigate the consequences of an analyzed accident. 
Consequently, no analysis assumptions are violated and there are no 
adverse effects on the factors that contribute to off-site or on-
site dose as the result of an accident.
    The format, typographical, grammatical, and standardized naming 
convention changes in addition to the WORD conversion are 
administrative in nature and therefore have no impact on accident 
initiators or plant equipment.
    Based on the above, the proposed administrative changes and 
permitting DG inspections to be performed during plant operation 
does not involve a significant increase in the probabilities or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident mechanisms would be created as a result of NRC 
approval of this amendment request since no changes are being made 
to the plant that would introduce any new accident mechanisms. 
Equipment would be operated in the same configurations with the 
exception of the mode in which the inspection is credited. The 
inspection will be performed within the current approved Technical 
Specification limiting condition for operation (LCO). This amendment 
request does not impact any plant systems that are accident 
initiators or adversely impact any accident mitigating systems.
    The proposed administrative changes do not involve any 
modifications to the physical plant or operations. Administrative 
changes do not contribute to accident initiators nor do they produce 
a new accident scenario. Based on the above, implementation of the 
proposed change would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
fuel cladding, the reactor coolant system, and the containment 
system. The proposed change to the inspection timing for the DGs do 
not affect the operability requirements for the DGs, as verification 
of such operability will continue to be performed as required. 
Continued verification of operability supports the capability of the 
DGs to perform their required function of providing emergency power 
to plant equipment that supports the fission product barriers. 
Consequently, the performance of these fission product barriers will 
not be impacted by implementation of this license amendment request 
and therefore does not involve a significant reduction in the margin 
of safety.
    The administrative changes do not affect plant equipment or 
operation. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

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

[[Page 810]]

    Attorney for licensee: John H. O'Neill, Jr., Esq., Shaw Pittman, 
Potts & Trowbridge, 2300 N. Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment requests: August 27, 2002.
    Description of amendment requests: The proposed license amendments 
would revise the term ``minimum measured flow per loop'' to ``measured 
loop flow'' in the allowable value and nominal trip setpoint for the 
Reactor Coolant Flow-Low reactor trip function contained in Table 
3.3.1-1, ``Reactor Trip System Instrumentation,'' of Technical 
Specification (TS) 3.3.1. In addition, the proposed amendments would 
allow for an alternate method for the measurement of reactor coolant 
system (RCS) total volumetric flow rate through measurement of the 
elbow tap differential pressures on the RCS primary cold legs. The use 
of elbow tap differential pressures normalized to Diablo Canyon Power 
Plant Cycle 1 and 2 precision flow calorimetrics would improve the 
accuracy of the RCS flow measurement through reduction of the effect of 
hot leg temperature streaming that is present in the current flow 
calorimetric method.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises the Technical Specification (TS) 
3.3.1 Table 3.3.1-1 term ``minimum measured flow per loop'' to 
``measured loop flow'' in the allowable value and nominal trip 
setpoint for the Reactor Coolant Flow-Low reactor trip function and 
allows an alternate method for the measurement of reactor coolant 
system (RCS) total flow to meet TS surveillance requirement (SR) SR 
3.4.1.4 through measurement of the elbow tap differential pressures 
on the RCS primary cold legs.
    The change will not increase the probability of an accident 
previously evaluated because adequate RCS flow will still be 
assured. The Reactor Coolant Flow-Low reactor trip function 
allowable value and nominal trip setpoint are accident mitigation 
functions and are not an accident initiator. The elbow tap method to 
measure RCS flow and the change to the flow definition associated 
with the Reactor Coolant Flow-Low reactor trip function do not 
involve a plant modification.
    For the elbow tap method to measure RCS flow, sufficient margin 
exists to account for all reasonable instrument uncertainties and 
therefore the RCS flow will continue to be maintained at a value 
which is bounded by the design basis accident initial conditions. 
The change to the flow definition associated with the Reactor 
Coolant Flow-Low reactor trip function allowable value and nominal 
trip setpoint does not change a design basis accident initial 
condition or the conditions at the time of reactor trip during a 
design basis accident and therefore has no adverse effect on the 
design basis accidents which credit the Reactor Coolant Flow-Low 
reactor trip setpoint.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to the flow definition associated with the 
Reactor Coolant Flow-Low reactor trip function allowable value and 
nominal trip setpoint and the proposed elbow tap method to measure 
RCS flow will not create the possibility of a new or different type 
of accident from any previously evaluated. There are no physical 
changes being made to the plant and there are no changes in 
operation of the plant that could introduce a new failure mode, 
creating an accident which has not been evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to the flow definition associated with the 
Reactor Coolant Flow-Low reactor trip function allowable value and 
nominal trip setpoint and the proposed elbow tap method to measure 
RCS flow will not reduce the margin of safety. For the proposed 
elbow tap flow method, sufficient margin exists to account for all 
reasonable instrument uncertainties and thus the RCS flow will 
continue to be maintained at a value which is bounded by the design 
basis accident initial conditions, and no adverse effect on the 
plant response to design basis accidents is created. The change in 
the flow definition associated with the Reactor Coolant Flow-Low 
reactor trip function allowable value and nominal trip setpoint does 
not change a design basis accident initial condition or the 
conditions at the time of reactor trip during a design basis 
accident, and therefore has no effect on the plant response to 
design basis accidents which credit the Reactor Coolant Flow-Low 
reactor trip setpoint. Since the change does not affect the response 
to design basis accidents, it does not result in a decrease in 
departure from nucleate boiling margin or reactor coolant system 
peak pressure margin for the design basis accidents.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: November 1, 2002.
    Description of amendment requests: The proposed license amendments 
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System 
(RTS) Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation'' as follows: (1) Revise both 
the RTS and ESFAS instrumentation TS and TS Bases to change or clarify 
the allowances for bypassing and tripping tested channels with other 
channels inoperable; (2) remove Surveillance Requirement 3.3.1.10 from 
Function 16.b, ``Turbine Stop Valve Closure;'' (3) correct the nominal 
trip setpoint value for Function 16.b, ``Turbine Stop Valve Closure;'' 
(4) correct the allowable value for the Function 18.f, ``Turbine 
Impulse Chamber Pressure, P-13;'' and (5) remove and relocate the 
nonsafety-related turbine trip function from Function 5 of Table 3.3.2-
1, ``Turbine Trip and Feedwater Isolation.'' This function will be 
relocated to other owner-controlled documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes in the required action statements in the 
Limiting Conditions for Operation (LCOs) for the allowable 
surveillance testing configurations for both the reactor trip system 
(RTS) and engineered safety feature actuation system (ESFAS) 
instruments will not change the probability or consequences of an 
accident previously evaluated.
    The proposed surveillance testing configuration changes only 
clarify available surveillance testing configurations and

[[Page 811]]

limitations on those configurations. The changes do not modify how 
the RTS and ESFAS functions respond to any accident condition. These 
surveillance testing configurations provide greater flexibility to 
prevent inadvertent actuation of these functions that could be a 
precursor for an accident.
    Previous Diablo Canyon Power Plant (DCPP) submittals have been 
approved providing for the capability of surveillance testing in 
trip and/or in bypass. Surveillance testing in bypass is considered 
the preferred method for most Eagle 21 instruments. However, where 
testing by tripping a single channel without causing a function 
actuation is acceptable, that capability was also maintained.
    Although some of the changes may appear to add new allowable 
surveillance testing configurations, all of the proposed 
configurations are based on the application of the intent behind the 
existing Technical Specification (TS) wording. The limitations on 
surveillance testing configurations provided by the proposed changes 
are to ensure that there are no spurious actuations and that during 
testing a valid signal will cause the associated functions to 
actuate as designed. None of these configurations place the 
associated function in a logic that has not been previously 
evaluated and approved.
    The proposed elimination of the channel calibration for the 
turbine stop valve position switches will not change the probability 
or consequences of an accident previously evaluated since these 
switches are not subject to drift. These limit switches are 
installed with fixed limit setpoints that actuate based on valve 
position and they are not calibrated in the field. As a result, a 
channel calibration being performed on these switches provides no 
useful purpose other than to verify function similar to the 
remaining trip actuation device operational test (TADOT). As a 
result, performing only the TADOT provides all necessary assurances 
of operability.
    The correction of the turbine stop valve closure nominal trip 
setpoint is administrative in nature and will not change the 
probability or consequences of an accident previously evaluated. 
This was an oversight in the Improved Technical Specification (ITS) 
review and conversion process. The proposed change only returns the 
setpoint to the previously evaluated value.
    The proposed change to the allowable value for Function 18.f, 
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in 
nature and will not change the probability or consequences of an 
accident previously evaluated. The P-13 intended trip setpoint has 
always been maintained at 10 percent and remains unchanged. This 
modification is performed to provide consistency with current 
methodology and NUREG-1431, and does not affect the operation of the 
protective function.
    The proposed removal and relocation of the turbine trip function 
from ESFAS Function 5 will not change the probability or 
consequences of an accident previously evaluated. The turbine trip 
function is nonsafety-related and is not credited in any design 
bases accident scenario. The proposed change only clarifies 
importance of the two trip functions. The proposed changes in this 
LAR [License Amendment Request] do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes in the required action statements in the 
LCOs for the allowable surveillance testing configurations for both 
the RTS and ESFAS instruments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed changes only clarify previously available 
surveillance testing configurations and limitations on those 
configurations. These clarifications ensure maximum surveillance 
testing flexibility to prevent inadvertent actuation of these 
functions that could be a precursor for an accident. The changes do 
not modify any equipment, hardware or how the RTS and ESFAS 
functions respond to any accident condition.
    The proposed elimination of the channel calibration for the 
turbine stop valve position switches will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. This change does not modify any equipment, hardware or 
functions. The switches are installed with fixed limit setpoints 
that actuate based on valve position. The switches are not subject 
to drift and are not calibrated in the field. As a result, a channel 
calibration being performed on these switches provides no useful 
purpose other than to verify function similar to the required TADOT. 
As a result, performing only the TADOT provides equivalent 
assurances of operability.
    The correction of the turbine stop valve closure nominal trip 
setpoint in Function 16.b, ``Turbine Stop Valve Closure,'' is 
administrative in nature and will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. This was an oversight in the ITS review and conversion 
process. The proposed change does not modify any hardware or 
equipment, and only returns the setpoint to the previously evaluated 
value.
    The proposed change to the allowable value for Function 18.f, 
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in 
nature and will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The P-13 
intended (nominal) trip setpoint has always been maintained at 10 
percent and remains unchanged. This change does not modify any 
equipment or hardware. This modification is performed to provide 
consistency with current methodology and NUREG-1431, and does not 
affect the operation of the protective function.
    The proposed removal and relocation of the turbine trip function 
from ESFAS Function 5 will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The turbine trip function is nonsafety-related and is not credited 
in any design bases accident scenario. The proposed change only 
clarifies importance of the two trip functions.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes in the required action statements in the 
LCOs for the allowable surveillance testing configurations for both 
the RTS and ESFAS instruments will not involve a significant 
reduction in a margin of safety. The proposed changes only clarify 
previously available surveillance testing configurations and 
limitations on those configurations. These clarifications ensure 
maximum surveillance testing flexibility to prevent inadvertent 
actuation of these functions that could be a precursor for an 
accident. The changes do not modify any equipment, hardware or how 
the RTS and ESFAS functions respond to any accident condition.
    The proposed elimination of the channel calibration for the 
turbine stop valve position switches will not involve a significant 
reduction in a margin of safety. This change does not modify any 
equipment, hardware or functions. The switches are installed with 
fixed limit setpoints that actuate based on valve position. The 
switches are not subject to drift and are not calibrated in the 
field. As a result, a channel calibration being performed on these 
switches provides no useful purpose other than to verify function 
similar to the required TADOT. As a result, performing only the 
TADOT provides equivalent assurances of operability.
    The correction of the turbine stop valve closure nominal trip 
setpoint in Function 16.b, is administrative in nature and will not 
involve a significant reduction in a margin of safety. This was an 
oversight in the ITS review and conversion process. The proposed 
change does not modify any hardware or equipment, and only returns 
the setpoint to the previously evaluated value.
    The proposed change to the allowable value for Function 18.f, 
``Turbine Impulse Chamber Pressure, P-13,'' is administrative in 
nature and will not involve a significant reduction in a margin of 
safety. The P-13 intended (nominal) trip setpoint has always been 
maintained at 10 percent and remains unchallenged. This change does 
not modify any equipment or hardware. This modification is performed 
to provide consistency with current methodology and NUREG-1431, and 
does not affect the operation of the protective function.
    The proposed removal and relocation of the turbine trip function 
from ESFAS Function 5 does not involve a significant reduction in a 
margin of safety. The turbine trip function is nonsafety-related and 
is not credited in any design bases accident scenario. The proposed 
change only clarifies importance of the two trip functions.
    None of the proposed changes affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 812]]

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

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

    Date of amendments request: December 9, 2002.
    Description of amendments request: The proposed amendments would 
revise Technical Specification 3.7.5, ``Auxiliary Feedwater System,'' 
Surveillance Requirement (SR) 3.7.5.2 for San Onofre Nuclear Generating 
Station, Units 2 and 3. Specifically, the proposed change would change 
wording of the Frequency of SR 3.7.5.2 from ``31 days on a Staggered 
Test Basis'' to ``In accordance with the Inservice Testing Program.'' 
Such inservice tests confirm component operability, trend performance, 
and detect incipient failures by indicating abnormal performance. This 
change is requested to implement recommendations from the Standard 
Technical Specifications for Combustion Engineering Plants, NUREG-1432, 
Revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    In June 2001, the Nuclear Regulatory Commission (NRC) issued 
NUREG 1432, Revision 2, ``Standard Technical Specifications 
Combustion Engineering Plants.'' For Technical Specification 3.7.5, 
``Auxiliary Feedwater (AFW) System,'' Surveillance Requirement (SR) 
3.7.5.2 requires verification that each AFW pump's developed head at 
the flow test point is greater than or equal to the required 
developed head which ensures that AFW pump performance has not 
degraded during the cycle. This test confirms one point on the pump 
design curve and is indicative of overall performance. This proposed 
change will revise San Onofre Nuclear Generating Station (SONGS) 
Surveillance Frequency to be consistent with NUREG 1432, Revision 2. 
This change in and of itself will have no effect on the probability 
or consequences of an accident previously evaluated.
    Once this change to the Technical Specification is approved, 
changes to the Surveillance Frequency of the AFW pumps would be 
controlled in accordance with the Risk-Informed Inservice Testing 
Program.
    Therefore, the proposed change does not involve a significant 
increase in the probability of consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment will not change the design, configuration 
or method of operation of the plant. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment will change the SR 3.7.5.2 Frequency from 
``31 days on a Staggered Test Basis'' to ``In accordance with the 
Inservice Testing Program.'' The proposed change does not change the 
operation or surveillance requirements. It does not change the 
design function of any of AFW system components. Therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.
    Based on the above, Southern California Edison concludes that 
the proposed amendment present no significant hazards consideration 
under the standards set forth in 10 CFR 50.92(c), and, accordingly a 
finding of ``no significant hazards consideration'' is justified.

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

Southern 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 2, 2002.
    Description of amendment request: The proposed amendments change 
Technical Specification Surveillance Requirement 3.6.4.1.2 to require 
that only one access door in each access opening of the secondary 
containment be verified closed every 31 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. [Does] the proposed change [* * *] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated[?]
    The proposed change to Surveillance Requirement SR 3.6.4.1.2 
would require that only one of the two secondary containment access 
doors be verified closed; presently, both doors are required to be 
verified closed. This change is administrative in nature in that it 
does not involve, require, or result from any physical change to me 
secondary containment boundary or access door configuration. The 
change to Surveillance Requirement SR 3.6.4.1.2 is consistent with 
TSTF Standard Technical Specification Change Traveler TSTF-18, 
Revision 1, and Surveillance Requirement SR 3.6.4.1.3 of Revision 2 
of Volume 1 of NUREG-1433. As indicated in the ``Justification'' 
portion of Standard Technical Specification Change Traveler TSTF-18, 
Revision 1, verifying one of the two access doors is closed is 
sufficient to ensure that the infiltration of outside air does not 
prevent the establishment and preservation of the required negative 
pressure within the secondary containment. Indeed, neither the 
requirements regarding minimum negative pressure and maximum 
infiltration and drawdown time nor the actions required to be taken 
should these requirements not be met will be altered by me proposed 
Licensing amendment.
    Because the physical characteristics and performance 
requirements of the secondary containment will not be altered and 
the change to Surveillance Requirement SR 3.6.4.1.2 is consistent 
with the current revision of NUREG-1433, the proposed Licensing 
amendment can not involve a significant increase in the probability 
or consequences of any accident previously evaluated.
    2. [Does] the proposed change [* * *] create the possibility of 
a new or different kind of accident from any previously evaluated[?]
     For the reasons previously discussed, neither the secondary 
containment boundary nor the access door configuration will be 
altered by or because of the proposed change to the surveillance 
requirement. Likewise, the requirements defining and governing 
secondary containment operability and functionality, that is, 
Standby Gas Treatment system flow rate and secondary containment 
negative pressure and drawdown limits, will not be changed. The 
secondary containment, including its access openings, will remain 
physically unaltered; will function as presently described in the 
Updated Final Safety Analysis Report [(UFSAR)]; and will be subject 
to the same structural and functional requirements. Under these 
circumstances, this change can not, and 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 
decrease in the margin of safety[?]

[[Page 813]]

    The requirements defining and governing secondary containment 
operability and functionality, that is, Standby Gas Treatment system 
flow rate and secondary containment negative pressure and drawdown 
limits, will not be changed. The secondary containment, including 
its access openings will function as presently described in the [* * 
*] UFSAR and will be subject to the same structural and functional 
requirements. Therefore, this change can not, and does not, reduce 
any margin of safety associated with the secondary containment 
function.

    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.

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

    Date of amendment request: December 4, 2002.
    Brief description of amendments: The proposed amendments revise 
several of the Required Actions in the Technical Specifications (TS) 
that require suspension of operations involving positive reactivity 
additions or suspension of operations involving reactor coolant system 
(RCS) boron concentration reductions. In addition, the proposed 
amendments revise several Limiting Conditions for Operation (LCO) Notes 
that preclude reductions in RCS boron concentration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The RTS [Reactor Trip System] instrumentation 
and reactivity control systems will be unaffected. Protection 
systems will continue to function in a manner consistent with the 
plant design basis. All design, material, and construction standards 
that were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Final Safety Analysis Report] are not 
adversely affected because the changes to the Required Actions and 
LCO Notes assure the limits on SDM [Shutdown Margin] and refueling 
boron concentration continue to be met, consistent with the analysis 
assumptions and initial conditions included within the safety 
analysis and licensing basis. The activities covered by this 
amendment application are routine operating evolutions. The proposed 
changes do not reduce the capability of reborating the RCS.
    The proposed changes will not involve a significant increase in 
the probability of any event initiators. The initiating event for an 
inadvertent boron dilution event, as discussed in FSAR Section 
15.4.6, is a failure in the reactor makeup control system (RMCS) or 
operator error such that inventory makeup with the incorrect boron 
concentration enters the RCS by way of the CVCS [Chemical and Volume 
Control System]. Since the RMCS design is unchanged, there will be 
no initiating event frequency increase associated with equipment 
failures. However, there could be an increased exposure time per 
operating cycle to potential operator errors during TS Conditions 
that, heretofore, prohibited positive reactivity additions. As such, 
the RTS Instrumentation and RCS Loops TS Bases changes from TSTF 
[Technical Specification Task Force]-286, Revision 2, have been 
augmented to preclude the introduction of reactor makeup water into 
the RCS via the CVCS when one source range neutron flux channel is 
inoperable or when no RCS loop is in operation. The equipment and 
processes used to implement RCS boration or dilution evolutions are 
unchanged and the equipment and processes are commonly used 
throughout the applicable MODES under consideration. There will be 
no degradation in the performance of, or an increase in the number 
of challenges imposed on, safety-related equipment assumed to 
function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation 
performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the radiological consequence evaluations 
in the FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating limits. The proposed changes 
merely permit the conduct of normal operating evolutions when 
additional controls over core reactivity are imposed by the 
Technical Specifications. The proposed changes do not introduce any 
new equipment into the plant or alter the manner in which existing 
equipment will be operated. The changes to operating procedures are 
minor, with clarifications provided that required limits must 
continue to be met. No performance requirements or response time 
limits will be affected. These changes are consistent with 
assumptions made in the safety analysis and licensing basis 
regarding limits on SDM and refueling boron concentration.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.
    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the limits on SDM or refueling 
boron concentration. The nominal trip setpoints specified in the 
Technical Specifications Bases and the safety analysis limits 
assumed in the transient and accident analyses are unchanged. None 
of the acceptance criteria for any accident analysis is changed. 
There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (FDH), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

[[Page 814]]

requirements of the Atomic Energy Act of 1954, as amended (the Act), 
and the Commission's rules and regulations. The Commission has made 
appropriate findings as required by the Act and the Commission's rules 
and regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].

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

    Date of amendment request: September 16, 2002.
    Brief description of amendment: The amendment revises a license 
condition by deleting the requirement to include check valve MVD-V5008 
in the facility check valve program.
    Date of issuance: December 13, 2002.
    Effective date: December 13, 2002.
    Amendment No.: 251.
    Facility Operating License No. DPR-62: Amendment revises Appendix 
B, ``Additional Conditions.''
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68731).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 13, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 31, 2002, as 
supplemented on September 18, 2002.
    Brief description of amendments: The amendments change the method 
of verifying boron concentration of each safety injection tank. Rather 
than taking a sample of each tank every 31 days, the revised technical 
specification surveillance requirement requires leakage into the tanks 
to be monitored every 12 hours and a sample to be taken every 6 months.
    Date of issuance: December 19, 2002.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 255 and 232.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2002. The 
September 18, 2002, letter provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of these amendments 
is contained in a Safety Evaluation dated December 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 10, 2001, as supplemented by 
letters dated June 19 and November 8, 2002. The supplemental 
information provided clarification that did not change the scope or the 
initial no significant hazards consideration determination.
    Brief description of amendment: The amendment revises TS 3/4.9.7 
and the corresponding Bases to address the use of a single-failure-
proof-handling system for the Spent Fuel Building and to remove the 
restriction on travel of crane loads in excess of 1800 pounds.
    Date of issuance: December 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 198.
    Facility Operating License No. DPR-61: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10009).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 17, 2002.
    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: August 29, 2002.
    Brief description of amendments: The amendments revised the 
Technical Specifications 3.8.4.7, to modify the note to eliminate the 
``once per 60 months'' restriction on replacing the battery service 
test by the battery modified performance discharge test. Associated 
changes to the TS Bases are also included.
    Date of issuance: December 17, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 209 & 190.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68733).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 30, 2002, as supplemented on 
October 31, 2002.
    Brief description of amendment: The amendment revised the 
requirements in several administrative programs in Technical 
Specification Section 6.0, ``Administrative Controls.'' Specifically, 
the amendment: (1) Replaced the specific management titles for several 
organizational positions with generic titles, (2) replaced the title of 
the Quality Assurance Program Description

[[Page 815]]

with a reference to the quality assurance program described or 
referenced in the Updated Final Safety Analysis Report, and (3) deleted 
the functions of the Station Nuclear Safety and the Nuclear Facilities 
Safety Committees and the Vice President-Nuclear Power since their 
duties and responsibilities are described in the Quality Assurance 
Program Description.
    Date of issuance: December 17, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42824).
    The October 31 supplemental letter provided clarifying information 
that did not expand the scope of the amendment or change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 17, 2002.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: June 28, 2002, as supplemented 
on October 15 (two separate letters), October 17, November 15, and 
December 6, 2002.
    Brief description of amendment: The amendment increases the 
licensed reactor core power level by 1.66 percent from 3250 megawatts 
thermal (MWt) to 3304 MWt. The power level increase is considered a 
measurement uncertainty recapture power uprate.
    Date of issuance: December 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 273.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48219).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 8, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification 2.7, ``Electrical Systems,'' to increase the amount of 
diesel fuel oil required for seven days of emergency diesel generator 
operation.
    Date of issuance: December 16, 2002.
    Effective date: December 16, 2002, and to be implemented within 30 
days of issuance.
    Amendment No.: 213.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68741).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 24, 2002, as supplemented 
by letter dated September 24, 2002.
    Brief description of amendments: The amendments delete Technical 
Specification 5.5.3, ``Post Accident Sampling System (PASS),'' and 
thereby eliminate the requirements to have and maintain the PASS at 
Plant Hatch.
    Date of issuance: December 18, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 235 & 177.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50958).
    The supplement dated September 24, 2002, provided clarifying 
information that did not change the scope of the June 24, 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 December 18, 2002.
    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 amendments request: October 24, 2001, as supplemented by 
correspondent e-mails dated August 27, 2002, and September 24, 2002.
    Brief description of amendments: The amendments consist of 
relocating various Technical Specifications (TSs) to the Technical 
Specification Requirements Manual (TRM). The amendments will relocate 
TSs 3/4.1.3.3, 3/4.3.3.2, 3/4.3.3.11, 3/4.4.7, 3/4.4.9.2, 3/4.3.4.11, 
3/4.7.2, 3/4.7.10, 3/4.9.3, 3/4.9.5, 3/4.9.7, 3/4.10.5, and 3/4.11.2.5 
to the TRM. Their associated bases will also be relocated to the TRM to 
be consistent with relocation of the various TSs. In addition, the 
proposed amendment corrects various typographical and page numbering 
errors, deletes an outdated one-time exception, and makes minor formal 
changes to improve consistency.
    Date of issuance: The license amendment is effective as of its date 
of issuance and shall be implemented within 6 months from the date of 
issuance.
    Effective date: December 17, 2002.
    Amendment Nos.: Unit 1--145; Unit 2--33.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5334).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 3, 2002.
    Description of amendment request: The proposed amendment revised 
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period is extended from the current limit of 
``* * * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for

[[Page 816]]

any Surveillance delayed greater than 24 hours and the risk impact 
shall be managed.''
    Date of issuance: December 23, 2002.
    Effective date: Date of issuance, to be implemented within 45 days.
    Amendment Nos.: 243, 278, 237.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63698).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2002.
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 3, 2002.
    Description of amendment request: The proposed amendment revises 
Surveillance Requirement (SR) 3.0.3 to extend the delay period, before 
entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period is extended from the current limit of 
``* * * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``* * * up to 24 hours or up to the limit of the 
specified Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: December 11, 2002.
    Effective date: Date of issuance, to be implemented within 45 days.
    Amendment No.: 42.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63699).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 11, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket No. 50-280, Surry 
Power Station, Unit 1, Surry County, Virginia

    Date of application for amendment: October 15, 2001, as 
supplemented November 8, 2001, June 28, 2002, and July 25, 2002 .
    Brief Description of amendment: This amendment revises the 
Technical Specifications to allow a one-time change in the Appendix J 
Type A containment integrated leakage rate test interval from the 
required 10 years to a test interval of 15 years at Surry Power 
Station, Unit 1.
    Date of issuance: December 16, 2002.
    Effective date: December 16, 2002.
    Amendment No.: 233.
    Facility Operating License No. DPR-32: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64309). The November 8, 2001, June 28, 2002, and July 25, 2002, 
supplements contained clarifying information only and did not change 
the initial no significant hazards consideration determination or 
expand the scope of the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of December 2002.

    For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 03-156 Filed 1-6-03; 8:45 am]
BILLING CODE 7590-01-P