[Federal Register Volume 68, Number 13 (Tuesday, January 21, 2003)]
[Notices]
[Pages 2796-2810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-1161]


=======================================================================
-----------------------------------------------------------------------

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 27, 2002 through January 9, 2003. 
The last biweekly notice was published on January 7, 2003 (68 FR 798).

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 20, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request

[[Page 2797]]

for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714,\1\ which is available at the 
Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    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].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: November 27, 2002.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS). Licensees were generally required to implement PASS upgrades as 
described

[[Page 2798]]

in NUREG-0737, ``Clarification of TMI [Three Mile Island] Action Plan 
Requirements,'' and Regulatory Guide 1.97, ``Instrumentation for Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
Conditions During and Following an Accident.'' Implementation of these 
upgrades was an outcome of the lessons learned from the accident that 
occurred at TMI Unit 2. Requirements related to PASS were imposed by 
Order for many facilities and were added to or included in the TS for 
nuclear power reactors currently licensed to operate. Lessons learned 
and improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System 
(PASS).'' The NRC staff issued a notice of opportunity for comment in 
the Federal Register on December 27, 2001 (66 FR 66949), on possible 
amendments concerning TSTF-413, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on March 20, 2002 (67 FR 13027). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated November 27, 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: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: November 27, 2002.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) 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 TSs for nuclear power reactors currently licensed to 
operate. However, 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 Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-413, ``Elimination of Requirements 
for a

[[Page 2799]]

Post Accident Sampling System (PASS).'' The NRC staff issued a notice 
of opportunity for comment in the Federal Register on December 27, 2001 
(66 FR 66949), on possible amendments concerning TSTF-413, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the consolidated line item improvement 
process. The NRC staff subsequently issued a notice of availability of 
the models for referencing in license amendment applications in the 
Federal Register on March 20, 2002 (67 FR 13027). The licensee affirmed 
the applicability of the following NSHC determination in its 
application dated November 27, 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.

    Based upon the reasoning presented above and the previous 
discussion of the proposed amendment, the NRC staff proposes to 
determine that the requested amendment does not involve a significant 
hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New York

    Date of amendment request: December 16, 2002.
    Description of amendment request: The licensee proposed to amend 
the Oyster Creek Nuclear Generating Station (OCNGS) Technical 
Specifications (TSs) regarding the safety limit minimum critical power 
ratio (SLMCPR) to reflect the results of cycle-specific calculations 
performed for the current fuel cycle (i.e., Cycle 19), using Nuclear 
Regulatory Commission (NRC)-approved methodology for determining SLMCPR 
values. Specifically, the licensee proposed to revise TS 2.1.A, 
changing the SLMCPR from 1.12 to 1.10 for three-recirculation-loop 
operation, and from 1.11 to 1.09 for four-or five-recirculation-loop 
operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Before commencement of Cycle 19, the licensee used NRC-approved 
methods and procedures in Topical Report NEDE-24011-P-A-14, ``General 
Electric Standard Application for Reactor Fuel'' (GESTAR II) and U.S. 
Supplement, NEDE-24011-P-A-14-US, dated June 2000, to derive the SLMCPR 
values for OCNGS, Cycle 19. The revised values were approved by the NRC 
staff via Amendment No. 233, dated September 26, 2002. Subsequently, 
the licensee recalculated these SLMCPR values using the methodology in 
Topical Report NEDC-32694-P-A, ``Power Distribution Uncertainties for 
Safety Limit MCPR Evaluation,'' and requested to revise these values 
further by the December 16, 2002, application.
    The analysis methodology incorporates cycle-specific parameters. 
These calculations do not change the operating procedures of OCNGS and 
have no effect on the probability of an accident initiating event or 
transient. The basis of the SLMCPR is to ensure no mechanistic fuel 
damage is calculated to occur if the limit is not violated. The new 
SLMCPR values preserve the existing margin to transition boiling and

[[Page 2800]]

the probability of fuel damage is not increased (i.e., in the event of 
an accident or transient, the amount of fuel damaged would not be 
increased as a result of the new SLMCPR values). Furthermore, the 
proposed new SLMCPR values do not lead to, nor do they arise as a 
result of, plant design or procedural changes. 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?
    The new SLMCPR values for OCNGS Cycle 19 core have been calculated 
in accordance with the methods and procedures described in NRC-approved 
topical reports. The proposed new SLMCPR values do not lead to, nor do 
they arise as a result of, plant design or procedural changes. The 
changes do not involve any new method for operating the facility and do 
not involve any facility modifications. As a result, no new initiating 
events or transients could develop from the proposed changes. 
Therefore, the proposed TS 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?
    The margin of safety as defined in OCNGS's licensing basis will 
remain the same. The new, cycle-specific SLMCPR values are calculated 
using NRC-approved methods and procedures that are in accordance with 
the current fuel design and licensing criteria. The SLMCPR values will 
remain high enough to ensure that greater than 99.9% of all fuel rods 
in the core are expected to avoid transition boiling if the limits are 
not violated, thereby preserving the fuel cladding integrity. 
Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Based on the above review, it appears that the three standards of 
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the requested amendment involves no significant hazards 
consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear 
Plant, Charlevoix County, Michigan

    Date of amendment request: November 20, 2002.
    Description of amendment request: The amendment request reflects 
organizational changes due to the transfer of the Palisades Plant from 
Consumers Energy to Nuclear Management Company. The revision reduces 
redundancy between the Defueled Technical Specifications (DTS) and the 
Big Rock Point Quality Program Description for Nuclear Power Plants. 
Other changes are being proposed to correct minor typographical, 
grammatical, and spelling errors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Because this proposed change involves only a change in reporting 
relationships, and no accidents previously evaluated consider 
administrative controls, the change does not involve a significant 
increase in the probability or consequences of an accident 
previously considered.
    2. Will the proposed change create the possibility of a new or 
different kind of accident [from any other accident] previously 
evaluated?
    The proposed change would result in moving requirements for 
certain reporting relationships from the Defueled Technical 
Specifications to the Consumers Energy Quality Program Description 
for Nuclear Power Plants, Part 1--Big Rock Point Plant (CPC-2A). 
Because the Topical Report, CPC-2A, requires prior NRC [U.S. Nuclear 
Regulatory Commission] approval for any changes which would reduce 
the level of commitment in that document, an equivalent level of NRC 
oversight is applied to changes to CPC-2A as are applied to changes 
to Chapter 6 (Administrative Controls) of the Defueled Technical 
Specifications. Therefore, no changes in administrative controls 
defined in CPC-2A that might create the possibility of a new or 
different kind of accident previously evaluated would be permitted 
by the proposed change.
    3. Will the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change stipulates that individuals who perform 
audits, surveillances and independent safety reviews will report as 
indicated in CPC-2A, which states that independent safety reviews 
are performed by the Restoration Safety Review Committee (RSRC). The 
proposed change involves no significant change in a margin of safety 
because margins of safety (in the Defueled Technical Specifications) 
are directly controlled by system design and operation in accordance 
with Limiting Conditions of Operation, Surveillances and Design 
Features specified in the Defueled Technical Specifications are 
affected by this proposed change.
    To the extent that design and operation of systems having safety 
margins might be affected by independent oversight, the following is 
offered as evidence that no significant reduction in margin of 
safety will result from the proposed change:
    [sbull] The Manager, Nuclear Performance Assessment Department 
(NPAD) and the RSRC both report their findings directly to the 
Senior Nuclear Officer; therefore there will be no change in the 
ultimate reporting relationship.
    [sbull] The membership of NPAD and RSRC consists of individuals 
who are independent of the plant organization.
    [sbull] Changes to the Topical Report, CPC-2A that would result 
in a reduction in level of commitment in the Quality Program 
Description require a review and approval process equivalent to 
proposed changes to the administrative controls specified in the 
Defueled Technical Specifications.
    [sbull] The requirements for performing onsite and offsite 
reviews and audits are specified in CPC-2A; the proposed change to 
the DTS to place the reporting relationship for individuals 
performing these audits and reviews eliminates redundancy between 
the Defueled Technical Specifications and CPC-2A.
    The NRC staff has reviewed the licensee's significant hazards 
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: David A. Mikelonis, Esquire, Consumers 
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Robert A. Gramm.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: December 11, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) related to N-1 loop operation. 
Specifically, the proposed changes would eliminate N-1 loop operation 
from particular sections of the TS and would make other changes that 
are clarifying and/or administrative in nature. In addition, the TS 
Bases would be revised to address the proposed changes.
    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

[[Page 2801]]

consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not alter the way any structure, system, 
or component functions and would not alter the way [in] which the 
plant is operated. The proposed changes do not involve any physical 
plant modifications. The proposed changes incorporate existing plant 
operational restrictions into the facility Technical Specifications 
and provide for the removal of information which is not applicable 
to plant operation.
    The proposed allowed outage times (i.e. the required action 
times for Specification 3.4.1.5) are reasonable and consistent with 
the existing technical specification outage times and consistent 
with industry guidelines, thereby ensuring affected components are 
restored in a timely manner. The proposed changes to surveillance 
requirements are also consistent with existing surveillance 
frequencies and focus the Technical Specifications on verifying 
normal plant configurations are maintained. The design basis 
accidents, including the uncontrolled rod withdrawal from 
subcritical and boron dilution events, will remain the same 
postulated events described in the Millstone Unit No. 3 Final Safety 
Analysis Report (FSAR), and the consequences of these events will 
not be affected.
    Therefore, the proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The proposed changes do not alter 
the way any structure, system, or component functions and do not 
alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
they have no impact on any accident analysis assumption. The 
proposed changes do not decrease the scope of equipment currently 
required to be OPERABLE or subject to surveillance testing, nor do 
the proposed changes affect any instrument setpoints or equipment 
safety functions. The effectiveness of Technical Specifications will 
be maintained since the changes will not alter the operation of any 
component or system, nor will the proposed changes affect any safety 
limits or safety system settings which are credited in a facility 
accident analysis. Therefore, there is no 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 4, 2002.
    Description of amendment request: The amendments revise Technical 
Specification 3.7.6 by changing the minimum combined inventory for 
Emergency Feedwater from 72,000 gallons to 155,000 gallons and 
eliminating the condensate storage tank as a source of this inventory.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the 
determination that this amendment request involves a No Significant 
Hazards Consideration by applying the standards established by the 
NRC regulations in 10 CFR 50.92. This ensures that operation of the 
facility in accordance with the proposed amendment would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. This revision to Technical Specification (TS) 3.7.6 changes 
the inventory requirements for the Upper Surge Tank (UST) and 
hotwell. These components provide a suction source to the Emergency 
Feedwater System (EFW). This increase in inventory from 72,000 
gallons to 155,000 gallons increases the required available 
inventory. This increase in inventory does not affect the 
probability or consequences of any previously evaluated accident.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. This revision to the combined UST and hotwell inventory 
increases the required amount of water available to the EFW system. 
No new or different kind of accident is created by this change as 
only the required inventory is revised.
    (3) Involve a significant reduction in a margin of safety:
    No. The increase in required UST and hotwell inventory does not 
reduce the margin of safety. The increase provides the required 
inventory to ensure that the EFW can provide a Reactor Coolant 
System cooldown at a rate of 50[deg] F/hour to decay heat removal 
entry conditions following a reactor trip.
    Duke has concluded, based on the above, that there is no 
significant hazards considerations involved in this amendment 
request.

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

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

    Date of amendment request: May 1, 2002, as supplemented December 4, 
2002.
    Description of amendment request: The proposed amendment would 
extend the applicability of the current Pilgrim Nuclear Power Station 
(Pilgrim) reactor pressure vessel pressure-temperature (P-T) curves 
through the end of Operating Cycle (OC) 16. The current P-T curves were 
approved for use in License Amendment 190, dated April 13, 2001, and 
are limited to use through the end of OC 14. The proposed change would 
delete the 20 and 32 Effective Full Power Year (EFPY) curves and 
replace the wording of the title blocks to allow use through the end of 
OC 16. The proposed amendment would change Pilgrim Technical 
Specification Figures 3.6.1, 3.6.2 and 3.6.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves a request to extend the use of the 
current reactor pressure vessel P-T curves for two additional OCs. The 
P-T curves were generated in accordance with the fracture toughness 
requirements of 10 CFR Part 50, Appendix G, and American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 
Section XI, Appendix G and Regulatory Guide 1.99, Revision 2, Radiation 
Embrittlement of Reactor Vessel Materials, and were established in

[[Page 2802]]

compliance with the methodology used to calculate and predict effects 
of radiation on embrittlement of reactor pressure vessel beltline 
materials. There are no physical changes to the plant or new modes of 
operation being introduced by the proposed change. Further, the 
proposed change does not involve a change to any activities or 
equipment and is not assumed in the safety analysis to initiate any 
accident sequence. The proposed change does not adversely affect the 
integrity of the reactor coolant pressure boundary such that its 
function in the containment of radioactive materials is affected. 
Additionally, the proposed change will not create any failure mode not 
bounded by previously evaluated accidents. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The current P-T curves were generated in accordance with the 
fracture toughness requirements of 10 CFR Part 50, Appendix G, and ASME 
Code, Section XI, Appendix G, and were approved by the U.S. Nuclear 
Regulatory Commission for use through OC 14. The proposed change would 
extend use of the P-T curves for two additional OCs. No new modes of 
operation are introduced by the proposed change. Plant operation in 
compliance with the current P-T curves ensures conditions in which 
brittle fracture of primary coolant pressure boundary materials is 
avoided. Accidents involving a breach of the primary coolant pressure 
boundary have previously been evaluated and no other types of accidents 
associated with the proposed change have been identified. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed curves were established in compliance with the 
methodology used to calculate and predict effects of radiation on 
embrittlement of reactor pressure vessel beltline materials and are 
estimated for 48 effective full-power years. The current curves are 
approved for use through the end of OC 14 ([sim]19 EFPYs) which 
provides a conservatism factor of 1.7 between the actual EFPYs at the 
end of OC 14 and the end-of-life curve (32 EFPY). The change would 
extend the use of the proposed curves to the end of OC 16 ([sim]23 
EFPYs) which provides a conservatism factor of approximately 2.0. The 
actual EFPYs at the end of OC 16 is bounded by the 48 EFPYs estimated 
for the current curves. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 10, 2002
    Description of amendment request: The proposed Technical 
Specification (TS) amendment request changes the diesel fuel 
specification to a more current revision in TS 4.10.C. The change would 
also makes administrative revisions to reflect generic position titles 
in TS 6.0, correct page numbers and titles in the Table on Contents, 
and delete the General Table of Contents. Bases pages were also revised 
to reflect the fuel specification revision as well as to make 
administrative changes to provide clarity and correct a misspelling.
    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 operation of the Vermont Yankee Nuclear Power Station 
[VY] in accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    VY has determined that the probability of occurrence of a 
previously evaluated accident is not increased because the proposed 
changes do not impact any accident initiating conditions. The 
proposed changes will have no significant impact on any safety 
related structures, systems or components. Additionally, the 
administrative changes do not affect any system operation or 
function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    VY has determined that the proposed changes do not involve any 
physical alteration of plant equipment and do not change the method 
by which any safety-related system performs its function. No new or 
different types of equipment will be installed. The proposed changes 
do not create any new accident initiators or involve an activity 
that could be an initiator of an accident of a different type.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    VY has determined that the proposed changes do not alter the 
basic operation of process variables, systems, or components as 
described in the safety analysis. No new equipment is introduced.
    The proposed changes do not impact design margins of any system 
to perform its intended safety functions. There is no physical or 
operational change being made which would alter the sequence of 
events, plant response, or margins in existing safety analyses. The 
proposed changes result in no impact on analyzed accident event 
precursors or effects. These proposed changes do not alter the 
physical design of the plant. There is no change in methods of 
operation.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: November 27, 2002.

[[Page 2803]]

    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 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 27, 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 requests 
involve no significant hazards consideration.
    Attorneys for licensees: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chiefs: Anthony J. Mendiola, James W. Clifford.

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

    Date of amendment request: October 30, 2002.
    Description of amendment request: The proposed amendment deletes 
requirements from the technical specifications (TS) and other elements 
of the licensing bases to maintain a Post Accident Sampling System 
(PASS).
    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 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 Three Mile Island (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

[[Page 2804]]

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 23 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, the Emergency Operating Procedures (at PNPP, these 
procedures are titled the Plant Emergency Instructions), and site 
survey monitoring that support modification of Emergency Plan 
Protective Action Recommendations (PARs). Therefore, the elimination 
of PASS requirements from Technical Specifications does not involve 
a significant increase in the consequences of any accident 
previously evaluated.
    2. The proposed change would 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.
    3. The proposed change will 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 do 
not rely on PASS are designed to provide rapid assessment of current 
reactor core conditions and the trending 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.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

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

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

    Date of amendment request: March 22, 2002, as supplemented May 13, 
June 24, July 29, and December 20, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Surveillance Requirement (SR) 
4.0.3 to extend the delay period, before entering a Limiting Condition 
for Operation (LCO), following a missed surveillance. The delay period 
would be extended from the current limit of ``* * * up to 24 hours to 
permit the completion of the surveillance when the allowable 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 NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process (CLIIP). The NRC 
staff subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated March 22, 2002, as supplemented May 13, June 24, July 29, and 
December 20, 2002.
    The proposed amendment would also add a requirement for a TS Bases 
Control Program to the administrative controls section of TSs. This 
change is necessary to be consistent with the CLIIP and is also 
consistent with the TS Bases Control Program presented in Section 5.5 
of NUREG-1431, Revision 2, ``Standard Technical Specifications 
Westinghouse Plants.'' The licensee provided its analysis of the issue 
of NSHC for this proposed change in its application.
    The proposed amendment would also modify SR 4.0.1, and its 
associated Bases, to link it with SR 4.0.3. The modification to SR 
4.0.1 is consistent with NUREG-1431, Revision 2, ``Standard Technical 
Specifications Westinghouse Plants.'' The licensee provided its 
analysis of the issue of NSHC for this proposed change in its 
application.
    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.

[CLIIP Change]

    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

[[Page 2805]]

consequences of an accident previously evaluated.
    [Addition of TS Bases Control Program and Changes to SR 4.0.1] 
The proposed changes to adopt the ITS [Improved Standard Technical 
Specifications] wording for Specification 4.0.1 and formally adopt a 
[TS] Bases Control Program are administrative in nature and do not 
adversely affect accident initiators or precursors nor alter the 
design assumptions, conditions, configuration of the facility or the 
manner in which it is operated. The proposed changes do not alter or 
prevent the ability or structures, systems, or components to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the 
Seabrook Station Updated Final Safety Analysis Report (UFSAR).
    Future changes to the TS Bases will continue to be 
administratively controlled pursuant to the provisions of 10 CFR 
50.59. The TS Bases is a licensee-controlled document that contains 
bases information for the [TS]. Future changes to the information 
contained in the TS Bases will be reviewed and approved in 
accordance with the FPLE Seabrook Regulatory Compliance Manual and 
TS Section 6.7.6j (TS Bases Control Program) of the Seabrook Station 
[TS]. Therefore, the proposed 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.

[CLIIP Change]

    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.

[Addition of TS Bases Control Program and Changes to SR 4.0.1]

    The proposed changes do not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated. There are no changes to the source term or 
radiological release assumptions used in evaluating the radiological 
consequences in the Seabrook Station UFSAR. The proposed changes 
have no adverse impact on component or system interactions. The 
proposed changes will not adversely degrade the ability of systems, 
structures and components important to safety to perform their 
safety function nor change the response of any system, structure or 
component important to safety as described in the UFSAR. The 
proposed changes are administrative in nature and do not change the 
level of programmatic and procedural details of assuring operation 
of the facility in a safe manner. Since there are no changes to the 
design assumptions, conditions, configuration of the facility, or 
the manner in which the plant is operated and surveilled, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

[CLIIP Change]

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

[Addition of TS Bases Control Program and Changes to SR 4.0.1]

    There is no adverse impact on equipment design or operation and 
there are no changes being made to the [TS] required safety limits 
or safety system settings that would adversely affect plant safety. 
The proposed changes are administrative in nature and do not reduce 
the level of programmatic or procedural controls associated with the 
activities presently performed via the aforementioned surveillance 
requirements.
    Future changes to the TS Bases information will be reviewed and 
approved in accordance with Seabrook Station [TS], Section 6.7, and 
as outlined in [FPLE Seabrook's] Regulatory Compliance programs. 
Specifically, changes to the Seabrook Station [TS] Bases require an 
evaluation pursuant to the provisions of 10 CFR 50.59 and review and 
approval by the Station Operation Review Committee (SORC) prior to 
implementation.
    Therefore, formal adoption of a TS-required TS Bases Control 
Program and adoption of ITS wording for Specification 4.0.1 do not 
involve a significant reduction in the margin of safety provided in 
the existing specifications.

    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 NSHC.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: November 15, 2002.
    Description of amendment request: The proposed amendment would 
revise the Donald C. Cook Nuclear Plant, Unit 2, operating license and 
Technical Specifications to increase the licensed power level to 3468 
Mega Watts Thermal (MWt), or 1.66 percent greater than the current 
level of 3411 MWt.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    Probability of Occurrence of an Accident Previously Evaluated--
In support of this Measurement Uncertainty Recapture (MUR) power 
uprate, a comprehensive evaluation was performed for nuclear steam 
supply system (NSSS) and balance of plant systems and components and 
analyses that could be affected by this change. A power calorimetric 
uncertainty calculation was performed, and the effect of increasing 
plant power by 1.66 percent on the plant's design and licensing 
basis was evaluated. The result of these evaluations is that all 
plant components will continue to be capable of performing their 
design function at an uprated core power of 3468 MWt. In addition, 
an evaluation of the accident analyses demonstrates that applicable 
analysis acceptance criteria continue to be met. No accident 
initiators are affected by this uprate and no challenges to any 
plant safety barriers are created by this change.
    Consequences of an Accident Previously Evaluated--This change 
does not affect the release paths, the frequency of release, or the 
source term for release for any accidents previously evaluated in 
the Updated Final Safety Analysis Report. Structures, systems, and 
components (SSC) required to mitigate transients remain capable of 
performing their design functions, and thus were found acceptable. 
The reduced uncertainty in the feedwater flow input to the power

[[Page 2806]]

calorimetric measurement ensures that applicable accident analyses 
acceptance criteria continue to be met, to support operation at a 
core power of 3468 MWt. Analyses performed to assess the effects of 
mass and energy remain valid. The source terms used to assess 
radiological consequences have been reviewed and determined to bound 
operation at the uprated condition.
    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.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed changes. The 
installation of the Caldon Leading Edge Flow Meter 
CheckPlusTM system has been analyzed, and failures of 
this system will have no adverse effect on any safety-related system 
or any SSCs required for transient mitigation. SSCs previously 
required for the mitigation of a transient remain capable of 
fulfilling their intended design functions. The proposed changes 
have no adverse effects on any safety-related system or component 
and do not challenge the performance or integrity of any safety-
related system.
    This change does not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than previously 
evaluated. Operating at a core power level of 3468 MWt does not 
create any new accident initiators or precursors. The reduced 
uncertainty in the feedwater flow input to the power calorimetric 
measurement ensures that applicable accident analyses acceptance 
criteria continue to be met, to support operation at a core power of 
3468 MWt. Credible malfunctions continue to be bounded by the 
current accident analysis of record or evaluations that demonstrate 
that applicable acceptance criteria continue to be met.
    Therefore, the proposed changes do 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 margins of safety associated with this MUR Uprate Program 
are those pertaining to core power. This includes those associated 
with the fuel cladding, Reactor Coolant System pressure boundary, 
and containment barriers. A comprehensive engineering review was 
performed to evaluate the 1.66 percent increase in the licensed core 
power from 3411 MWt to 3468 MWt. The 1.66 percent increase required 
that revised NSSS design thermal and hydraulic parameters be 
established, which then served as the basis for all of the NSSS 
analyses and evaluations. This engineering review concluded that no 
design transient modifications are required to accommodate the 
revised NSSS design conditions. NSSS systems and components were 
evaluated and it was concluded that the NSSS equipment has 
sufficient margin to accommodate the 1.66 percent power uprate. NSSS 
accident analyses were evaluated for the 1.66 percent power uprate. 
In all cases, the evaluations demonstrate that the applicable 
analyses acceptance criteria continue to be met. As such, the 
margins of safety continue to be bounded by the current analyses of 
record for this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In summary, based upon the above evaluation, [Indiana Michigan 
Power Company] has concluded that the proposed amendment involves no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and, accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: December 19, 2002.
    Description of amendment request: The proposed amendment would 
update and clarify the Technical Specifications (TSs) requirements for 
demonstrating shutdown margin (SDM). The proposed changes incorporate 
new, more restrictive, SDM limits; add the required limiting condition 
for operation (LCO) actions if the SDM is not met; and also add the 
surveillance requirements for verifying the SDM. These LCO actions and 
surveillance requirements are not currently specified in the TSs. The 
revised SDM limits account for the uncertainty in the demonstration of 
adequate SDM analytically or by measurement. The proposed changes also 
eliminate the unnecessary restriction requiring SDM demonstration in 
the cold shutdown condition. The option for SDM demonstration in the 
cold shutdown condition is retained consistent with the existing 
special test exception.
    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.
    Adequate SDM provides assurance that inadvertent criticalit[y] 
and potential control rod drop accidents (CRDAs) involving high 
worth control rods will not cause significant fuel damage. The SDM 
is not an accident initiator and, as such, will have no effect on 
the probability of an accident. The proposed changes incorporate 
more restrictive SDM limits and provide the necessary actions and 
verifications to assure that there will be no adverse effect on the 
initial conditions and assumptions of the accidents previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes do not involve physical changes to the plant or 
introduce any new modes of operation. Accordingly, continued 
assurance is provided that the process variables, structures, 
systems, and components are maintained such that there will be no 
degradation of any fission product barrier which could increase the 
radiological consequences of an accident. 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 proposed changes to the SDM limits and requirements will 
have no adverse effect on the design or assumed accident performance 
of any structure, system, or component, or introduce any new modes 
of system operation or failure modes. Moreover, the proposed changes 
will have no impact on conformance to 10 CFR [Code of Federal 
Regulations] 50, Appendix A, General Design Criterion 26 (GDC 26), 
in that the control rods will continue to satisfy the SDM 
requirements and provide assurance that the reactor can be made 
subcritical from all applicable operating conditions, transients, 
and design basis events. 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 provide separate SDM limits for testing 
consistent with the Improved Standard Technical Specifications 
(NUREG-1433 and NUREG-1434) where the highest worth control rod is 
determined analytically (0.38% Dk/k) or by measurement (0.28% Dk/k). 
The proposed SDM limits are more restrictive than the current limit 
(0.25% Dk/k) and account for the uncertainty in the demonstration of 
SDM by testing. The SDM will continue to account for changes in core 
reactivity during the fuel cycle. Therefore, the margin of safety is 
increased relative to the SDM assumptions for the control rod 
withdrawal error transient and CRDA analyses.

[[Page 2807]]

    Accordingly, 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 19, 2002.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) reporting requirements for the discovery of defective or degraded 
steam generator tubes.
    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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not have any effect on structures, 
systems, and components (SSCs) of the Kewaunee Nuclear Power Plant. 
The changes do not affect plant operations, any design function or 
an analysis that verifies the capability of an SSC to perform a 
design function. The changes do not change any previously evaluated 
accidents in the updated safety analysis report (UFSAR). As these 
changes are administrative, there is no increase in the probability 
and consequences of analyzed accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are administrative and do not change the 
design function or operation of any plant SSCs. The proposed changes 
do not create the possibility of a new or different kind of accident 
due to credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes modify NRC reporting requirements only. The 
changes do not exceed or alter a design basis or safety limit or 
significantly reduce the margin of safety.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Esq., Shaw Pittman, 
Potts & Trowbridge, 2300 N. Street, NW, Washington, DC 20037-1128.
    NRC Section Chief: L. Raghavan.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of amendment request: December 4, 2002.
    Description of amendment request: The proposed amendment changes 
the Hatch Unit 2 turbine building high temperature primary containment 
isolation value specified in Technical Specification Table 3.3.6.1-1, 
Item 1f.
    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[?]
    This TS [Technical Specification] revision request changes the 
allowable value for the turbine building high temperature primary 
containment isolation. The setpoint at which the isolation occurs 
has nothing to do with preventing a system break; therefore, this 
proposed change will not change the probability of occurrence of a 
small primary coolant system break.
    For the turbine building high temperature primary containment 
isolation, the analytical limit has been calculated at 207[deg]F 
with the allowable value at 200[deg]F. The calculation supporting 
these values accounts for instrument uncertainties thus confirming 
that adequate margin exists between the allowable value and the 
analytical limit. Accordingly, the consequences of a small primary 
system break are not significantly increased.
    2. [Does] the [* * *] proposed [* * *] change create the 
possibility of a new or different kind of accident from any 
previously evaluated[?]
    Changing an allowable value does not introduce any new operating 
modes for any plant system or piece of equipment. All plant systems 
will continue to be operated, tested and maintained as before, and 
within their licensing and design basis. As a result, no new failure 
modes are introduced and the possibility of a new or different type 
of accident is not created.
    3. [Does] the [* * *] proposed [* * *] change involve a 
significant decrease in the margin of safety[?]
    Increasing the allowable value by 6[deg]F does not result in a 
significant reduction in a margin of safety. A formal calculation 
was performed which justified an analytical limit of 207[deg]F. This 
calculation determined the analytical limit based on a primary leak 
into the turbine building and confirmed that the allowable value 
adequately protects the analytical limit. As a result, the margin of 
safety is not significantly reduced.

    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.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 2808]]

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 email to [email protected].

Dominion Nuclear Connecticut, Inc. Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendment: February 14, 2002, as 
supplemented on September 9, 2002.
    Brief description of amendment: The amendments revised the 
Millstone Power Station, Unit No. 2 (MP2) and 3 (MP3) Technical 
Specifications (TSs) by relocating selected MP2 and MP3 TSs related to 
the Reactor Coolant System and Plant Systems to the respective unit's 
Technical Requirements Manual.
    The amendment does not address changes to MP2 TS 3/4.7.10, 
``Snubbers,'' and MP3 TSs 3/4.7.10, ``Snubbers,'' and 3/4.7.14, ``Area 
Temperature Monitoring,'' as described by the application dated 
February 14, 2002, because these proposed TSs changes were withdrawn by 
the supplement dated September 9, 2002.
    Date of issuance: January 2, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 272 and 214.
    Facility Operating License Nos. DPR-65 and NPF-49: This amendment 
revised the TSs.
    Date of initial notice in Federal Register: April 16, 2002 (67 
FR18645). The September 9, 2002, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 2, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 12, 2002, as 
supplemented by letter dated December 30, 2002.
    Brief description of amendments: The amendments temporarily revised 
Technical Specifications (TS) 3.5.2, ``Emergency Core Cooling System;'' 
TS 3.6.6, ``Containment Spray System;'' TS 3.7.5, ``Auxiliary Feedwater 
System;'' TS 3.7.7, ``Component Cooling Water System;'' TS 3.7.8, 
``Nuclear Service Water System;'' and TS 3.8.1, ``AC Sources.''
    Date of issuance: January 7, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 203 & 196.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63692). The supplement dated December 30, 2002, provided clarifying 
information that did not change the scope of the September 12, 2002, 
application, nor the initial no significant hazard consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 7, 2003.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: August 29, 2002. Brief 
description of amendments: The amendments revise Technical 
Specification (TS) 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.
    Date of issuance: January 9, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 204 & 197.
    Facility Operating License Nos. NPF-35 and NPF-52: 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 January 9, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 22, 2002.
    Brief description of amendment: The amendment deletes a reference 
to Section 2.E in Section 2.F of Facility Operating License No. NPF-21. 
Section 2.E requires the licensee to fully implement and maintain in 
effect all provisions of the Commission-approved physical security, 
guard training and qualification, and safeguards contingency plans. 
Section 2.E is redundant because the reporting requirements and 
criteria for the physical security programs are specified in 10 CFR 
73.71 and Appendix G of 10 CFR Part 73.
    Date of issuance: January 9, 2003.
    Effective date: January 9, 2003 to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 183.
    Facility Operating License No. NPF-21: The amendment revised the 
operating license.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75871). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 9, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 21, 2002.
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency,

[[Page 2809]]

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: January 2, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 127.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61679). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 2, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request October 15, 2001, as supplemented by 
letter dated August 27, 2002.
    Brief description of amendment: The amendment provides additional 
information to support a modification to Technical Specification 3.4.7 
and limits Reactor Coolant System activity permitted by the ACTION 
statement to 60 microcuries per gram at all power levels. The letdown 
line break accident analysis in the Final Safety Analysis Report was 
also changed.
    Date of issuance: January 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 184.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications and Final Safety Analysis Report.
    Date of initial notice in Federal Register: October 28, 2002 (67 FR 
66009). The August 27, 2002, supplemental letter provided additional 
information and revised the no significant hazards consideration 
determination. The original Federal Register notice was published on 
November 28, 2001 (66 FR 56504), but was superceded by the October 28, 
2002 publication.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 26, 2002, as supplemented 
September 12, 2002.
    Brief description of amendments: Extend the use of the pressure-
temperature limits in Technical Specification Figure 3.4.6.1-1 to 32 
effective full power years.
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: January 2, 2003.
    Amendment Nos. 163 and 125.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50953). The supplement dated September 12, 2002, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 2, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-171, Peach Bottom Atomic 
Power Station, Unit 1, York County, Pennsylvania

    Date of application for amendment: May 21, 2002.
    Brief description of amendment: This proposed amendment will revise 
the Peach Bottom Atomic Power Station, Unit 1, License and Technical 
Specifications (TS) to: (1) Delete License Condition C(4) to reflect 
satisfaction of the minimum decommissioning trust fund amount at the 
time of transfer of the Facility Operating License; (2) revise License 
Condition C(5)(d) to reflect 30 days prior written notification to the 
Director of Nuclear Material Safety and Safeguards before modification 
of the decommissioning trust agreement in any material respect; (3) 
delete TS 2.1(B)3 and TS 2.4(b) to eliminate inconsistencies with 
reporting requirements in 10 CFR 20.2202, 50.73, and 73.71; (4) revise 
TS 2.2 to refer to the Facility Operating License; and (5) revise TS 
2.3 to refer to the radiological hazards associated with the facility.
    Date of Issuance: December 26, 2002.
    Effective Date: On the date of issuance of this amendment and must 
be fully implemented no later than 30 days from the date of issuance.
    Amendment No.: 11.
    Facility Operating License No. DPR-12: Amendment revised the 
License and TS with respect to administrative procedures or 
requirements.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61682). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 26, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 1, 2001, as supplemented 
by letters dated June 13, 2001, May 20, 2002, and June 28, 2002.
    Brief description of amendments: These amendments revised TS 3.7.1, 
``Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat 
Sink (UHS),'' to add operability requirements and surveillance 
requirements for the UHS spray bypass and large array valves, and 
reduce the allowed Completion Times for the conditions applicable to 
the RHRSW system.
    Date of issuance: December 30, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 206 and 180.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46481). The June 13, 2001, May 20, 2002, and June 28, 2002, letters 
provided additional information that clarified the application, but did 
not expand the scope of the application as originally noticed, and did 
not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 30, 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 amendment request: December 3, 2001, as supplemented by 
letter dated August 29, 2002.
    Brief description of amendments: The amendments revise TS 3.7.1.2,

[[Page 2810]]

``Auxiliary Feedwater System,'' to better reflect the four train 
auxiliary feedwater (AFW) system design at STP. Specifically, the 
changes specify the same allowed outage time (AOT) for any one 
inoperable motor-driven pump, regardless of train. The amendments also 
extend the AOT for one inoperable motor-driven pump from 72 hours to 28 
days. A sentence has also been added to Action d. stating that Limiting 
Condition for Operation (LCO) 3.0.3 and all other LCO actions requiring 
Mode changes are suspended until one of the four inoperable AFW pumps 
is restored to operable status. There is also an administrative change 
in the wording of the LCO to clarify that there are only four AFW pumps 
in each STP unit.
    Date of issuance: December 31, 2002.
    Effective date: December 31, 2002.
    Amendment Nos.: Unit 1--146; Unit 2--134.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2930). The supplement provided additional information that clarified 
the application, did not expand the scope as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 8, 2002.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.4.16, ``RCS [Reactor Coolant System] Specific 
Activity,'' to lower the Limiting Condition For Operation and 
associated Surveillance Requirements for Dose Equivalent Iodine-131 in 
the RCS from a specific activity of 1.0 [mu]Ci/gm to 0.45 [mu]Ci/gm.
    Date of issuance: January 6, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 102 and 102.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40026). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 6, 2003.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-338, North Anna 
Power Station, Unit 1, Louisa County, Virginia

    Date of application for amendment: December 7, 2001, as 
supplemented by letters dated June 28 and July 25, 2002.
    Brief description of amendment: This amendment permits a one-time 
extension of the current 10-year Title 10 of the Code of Federal 
Regulations Part 50, Appendix J, Option B, Type A test interval from 
April 3, 2003, to April 2, 2008.
    Date of issuance: December 31, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 234.
    Facility Operating License No. NPF-4: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21295). The supplemental letters dated June 28 and July 25, 2002, 
contained clarifying information only and did not change the proposed 
no significant hazards consideration determination or expand the scope 
of the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 31, 2002.
    No significant hazards consideration comments received: No.

    For the Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 13th day of January 2003.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-1161 Filed 1-17-03; 8:45 am]
BILLING CODE 7590-01-P