[Federal Register Volume 67, Number 122 (Tuesday, June 25, 2002)]
[Notices]
[Pages 42814-42836]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-15683]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from May 31, 2002, through June 13, 2002. The 
last biweekly notice was published on June 11, 2002 (67 FR 40019).

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 July 25, 2002, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should

[[Page 42815]]

consult a current copy of 10 CFR 2.714,\1\ which is available at the 
Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
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    \1\ The most recent version of Title 10 of the Code of Federal 
Regulations, published January 1, 2002, inadvertently omitted the 
last sentence of 10 CFR 2.741(d) and subparagraphs (d)(1) and (2), 
regarding petitions to intervene and contentions. Those provisions 
are extant and still applicable to petitions to intervene. Those 
provisions are as follows: ``In all other circumstances, such ruling 
body or officer shall, in ruling on--
    (1) A petition for leave to intervene or a request for hearing, 
consider the following factors, among other things:
    (i) The nature of the petitioner's right under the Act to be 
made a party to the proceeding.
    (ii) The nature and extent of the petitioner's property, 
financial, or other interest in the proceeding.
    (iii) The possible effect of any order that may be entered in 
the proceeding on the petitioner's interest.
    (2) The admissibility of a contention, refuse to admit a 
contention if:
    (i) The contention and supporting material fail to satisfy the 
requirements of paragraph (b)(2) of this section; or
    (ii) The contention, if proven, would be of no consequence in 
the proceeding because it would not entitle petitioner to relief.''
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    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland, by the above date. Because of continuing 
disruptions in delivery of mail to United States Government offices, it 
is requested that petitions for leave to intervene and requests for 
hearing be transmitted to the Secretary of the Commission either by 
means of facsimile transmission to 301-415-1101 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and because of continuing disruptions in delivery of mail to United 
States Government offices, it is requested that copies be transmitted 
either by means of facsimile transmission to 301-415-3725 or by e-mail 
to [email protected]. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
System's (ADAMS) Public Electronic Reading Room on the Internet at the 
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].

[[Page 42816]]

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: May 15, 2002 (102-04701).
    Description of amendments request: The amendments would revise 
Limiting Condition for Operation (LCO) 3.9.3, ``Containment 
Penetrations,'' of the Technical Specifications. The amendments would 
(1) modify LCO 3.9.3.b on one door in each air lock being closed and 
(2) add a note to LCO 3.9.3 about containment penetration flow paths 
providing direct access from the containment to the outside atmosphere 
may be unisolated under administrative controls. The amendments would 
allow the containment air lock and other penetrations to be open during 
core alterations or movement of irradiated fuel assemblies within 
containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment[s] to Technical Specification (TS) 
3.9.3[,] ``Containment Penetrations,'' would allow the personnel air 
locks and other containment penetrations to remain open during CORE 
ALTERATIONS or movement of irradiated fuel assemblies within 
containment. The position of the personnel air locks and other 
containment penetrations (open or closed) are not an initiator of 
any accident.
    The fuel handling accident contained in the Updated Final Safety 
Analysis Report [for Palo Verde], Revision 11[,] assumes that the 
personnel air locks, containment penetrations, and the equipment 
hatch are open and the entire airborne radioactivity reaching the 
containment [from the damaged fuel] is released to the outside 
environment. Using these assumptions, the current analysis results 
in off site doses that are well within guideline values specified in 
10 CFR [Part] 100[,] ``[R]eactor Site Criteria[,]'' and calculated 
control room doses within the acceptance criteria specified in 
General Design Criteria 19[,] ``Control Room.''
    Therefore, the proposed amendment request to allow the personnel 
air locks and [other] containment penetrations to be open during 
CORE ALTERATIONS [or] movement of irradiated fuel assemblies in 
containment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment[s] to TS 3.9.3[,] ``Containment 
Penetrations,'' allowing the personnel air locks and other 
containment penetrations to be open during CORE ALTERATIONS [or] 
movement of irradiated fuel in containment does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed). It does[,] however, involve a minor 
change in the methods governing normal plant operation during 
refueling. This minor change in personnel air lock and containment 
penetration control does not create the possibility of a new or 
different kind of accident. [Containment penetration control is not 
an initiator of an accident.] The fuel handling accident [(FHA)] 
analysis contained in the Updated Final Safety Analysis Report, 
Revision 11[,] already assumes that the personnel air locks, [other] 
containment penetrations, and the equipment hatch are open and the 
entire airborne radioactivity released in containment following a 
FHA is transported to the outside environment. This analysis results 
in off site doses that are well within guideline values specified in 
10 CFR [Part] 100[,] ``Reactor Site Criteria[,]'' and calculated 
control room doses within the acceptance criteria specified in 
General Design Criteria 19[,] ``Control Room.''
    Thus, the proposed amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment[s] to TS 3.9.3[,] ``Containment 
Penetrations,'' allowing the personnel air locks and other 
containment penetrations to be open during CORE ALTERATIONS [or] 
movement of irradiated fuel in containment remains bounded by 
previously determined radiological dose consequences for a FHA 
inside containment. The previously analyzed dose consequences 
assumes that the personnel air locks, containment penetrations, and 
the equipment hatch are open and the entire airborne radioactivity 
released in containment is transported to the outside environment. 
The results of this analysis were determined to be within the limits 
of 10 CFR [Part] 100[,] ``Reactor Site Criteria[,]'' and [* * *] 
meets the acceptance criteria of NUREG-0800[, ``] Standard Review 
Plan for the Review of Safety Analysis Reports for Nuclear Power 
Plants[,''] Section 15.7.4[,] ``Radiological Consequences of Fuel 
Handling Accidents.'' The calculated control room doses are within 
the acceptance criteria specified in General Design Criteria 19[,] 
``Control Room.'' There are no changes in the assumptions made about 
the positions of the containment openings and penetrations. 
Therefore, there is no change in the analysis results and the 
proposed amendment request does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.5.3, ``Post Accident Sampling 
System (PASS),'' and thereby eliminate the requirements to have and 
maintain the PASS at Fermi 2. 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 May 23, 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 [Three Mile Island, Unit 2] 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 42817]]

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, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
delete Section 2.F of the Operating License which requires reporting 
violations of the requirements in Section 2.C of the Operating License. 
The licensee stated that the requirements in Section 2.F are adequately 
addressed by the reporting requirements identified in 10 CFR 50.72 and 
10 CFR 50.73, and therefore, Section 2.F is not required. The proposed 
amendment would also delete License Conditions 2.C.(19), 2.C.(20) and 
2.C.(21), which pertain to historical actions that have been met.
    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.
    This License Amendment request involves administrative changes 
only. No actual plant equipment or accident analyses will be 
affected by the proposed changes. The three License Conditions 
proposed for deletion pertain to actions that have been completed 
and are no longer applicable. The reporting requirements in Section 
2.F of the Operating License are not required because they are 
either adequately addressed by 10 CFR 50.72 and 10 CFR 50.73, or 
contained in the specific License Condition (2.C.(10)). Therefore, 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes have no impact on the design, function or 
operation of any plant structure, system or component. The changes 
are administrative in nature and do not affect plant equipment or 
accident analyses. License Conditions 2.C.(19), 2.C.(20) and 
2.C.(21) can be deleted because they are no longer applicable. The 
reporting requirements in the Fermi 2 Operating License can be 
deleted because they are either adequately addressed in 10 CFR 50.72 
and 10 CFR 50.73, or are included in the specific License Condition 
(2.C.(10)). Therefore, these changes cannot create a new failure 
mode, nor can they create the possibility of a new or different kind 
of accident than any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed changes do not relax the bases for any limiting 
condition of operation nor do they affect the design or operation of 
any fission product barrier. The changes are administrative in 
nature and result in the deletion of obsolete License Conditions and 
reporting requirements that are adequately addressed elsewhere. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. The NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan, Section Chief.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
change the Fermi 2 Technical Specifications (TSs) to allow a one-time 
deferral of the Type A primary containment integrated leak rate test. 
Specifically, TS 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' would be revised to extend

[[Page 42818]]

the current interval for performing the containment Type A test to 15 
years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed License Amendment involves a one-time extension of 
the testing frequency for the primary containment 10 CFR [Part] 50, 
Appendix J, Type A test. The current 10-year test interval would be 
extended on a one-time basis to no longer than 15 years. The 
proposed Technical Specification (TS) change does not involve a 
physical plant change or a change in the manner in which the plant 
is operated or controlled. The primary containment is designed to 
provide an essentially leak tight barrier against an uncontrolled 
release of radioactivity to the environment resulting from 
postulated design basis accidents. As such, the primary containment 
and the testing requirements do not affect accident initiation; 
therefore, the proposed TS change does not involve a significant 
increase in the probability of an accident previously evaluated.
    Type B and C containment local leak rate testing will continue 
to be performed at the frequency required by the TS. As documented 
in NUREG-1493, ``Performance-Based Containment Leakage Test 
Program,'' industry experience has shown that Type B and C tests 
have identified about 97 percent of containment leakage paths, and 
only about 3 percent have been detected by a Type A test. NUREG-1493 
also concluded, in part, that reducing the frequency of Type A 
containment leakage rate test to once per 20 years would result in 
an imperceptible increase in risk. The Fermi 2 risk-based assessment 
of the proposed extension supports this conclusion. The design and 
construction of the primary containment, combined with the 
containment inspection program in accordance with the American 
Society of Mechanical Engineers (ASME) Code, Section XI, and the 
Maintenance Rule program per 10 CFR 50.65 requirements, provide a 
high degree of confidence that the containment will not degrade in a 
manner that is detectable only by Type A testing. Additionally, the 
inherent feature of Boiling Water Reactor containments which 
provides on-line containment monitoring capability, allows for early 
detection of gross containment leakage during power operation.
    Based on the above, the proposed change does not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The primary containment is designed to contain energy and 
fission products during and following design basis accidents. The 
containment and testing requirements, invoked to periodically 
demonstrate the integrity of the containment, ensure the plant's 
ability to mitigate the consequences of an accident; however, the 
containment and testing do not involve accident initiation. In 
addition, the proposed change to the Type A test frequency does not 
involve a physical change to the facility. The change does not 
affect the operation of the plant such that a new failure mode 
involving the possibility of a new or different kind of accident is 
created. Therefore, the proposed change does not create the 
potential for a new or different kind of accident from any accident 
previously evaluated.
    3. The [proposed] change does not involve a significant 
reduction in the margin of safety.
    The NUREG-1493 generic study on the effects of extending 
containment leakage testing found that reducing the Type A test 
frequency to once per 20 years resulted in an imperceptible increase 
in risk to the public. The NUREG study concluded that Type B and C 
testing detect most potential containment leakage. The extension of 
[the] Type A test interval to 15 years has a minimal effect on 
leakage detection capability. The TS allowed leakage limit is not 
impacted by this change, and the frequency of local Type B and C 
testing will not be altered as a result of this extension. 
Additionally, the containment inspection program provides a high 
degree of assurance that the containment will not degrade in a 
manner only detectable by Type A testing. On-line containment 
monitoring provides additional assurance for detecting gross 
containment leakage during power operation. The combination of all 
these factors ensures that the safety margin will be maintained. 
Therefore, the proposed changes will not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to eliminate the response 
time testing requirements for certain instrumentations in TS 3.3.1.1 
and TS 3.6.1.1, based on NRC-approved licensing topical report, NEDO-
32291-A, ``System Analyses for Elimination of Selected Response Time 
Testing Requirements,'' dated October 1995, and its Supplement 1, dated 
October 1999. This licensing topical report shows that other periodic 
tests required by TSs, such as channel calibrations, channel checks, 
channel functional tests, and logic system functional tests, provide 
adequate assurance that instrument response times are within acceptance 
limits. Therefore, the proposed change to delete the specific response 
time testing requirements does not change the response time assumptions 
in the Updated Safety Analysis Report. Only the methodology of time 
response verification would be changed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the Technical Specifications does not 
result in the alteration of the design, material, or construction 
standards that were applicable prior to the change. The same Reactor 
Protection System (RPS) and Primary Containment Isolation 
Instrumentation instrumentation [sic] is used, and the response time 
assumptions in [the] Updated Final Safety Analysis Report (UFSAR) 
Chapter 15 analysis remain unchanged. Only the methodology of time 
response verification is changed. The proposed change will not 
result in the modification of any system interface that would 
increase the likelihood of an accident since these events are 
independent of the proposed change. The proposed amendment will not 
change, degrade, or prevent actions, or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the UFSAR. Therefore, the proposed amendment 
does not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not alter the performance of the Reactor 
Protection System (RPS) or Primary Containment Isolation 
Instrumentation systems. All RPS and Primary Containment Isolation 
Instrumentation channels will still have an initial response time 
verified by test before initially placing the channel in operational 
service and after any maintenance that could affect response time. 
Changing the method of periodically verifying instrument response 
for certain RPS and Primary Containment Isolation Instrumentation 
channels (assuring equipment operability) from time response testing 
to calibration and channel checks will

[[Page 42819]]

not create any new accident initiators or scenarios. Periodic 
surveillance of these instruments will detect significant 
degradation in the channel characteristic. Implementation of the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Implementation of NEDO 32291-A methodologies for eliminating 
selected response time testing does not involve a significant 
reduction in the margin of safety. The current response time limits 
are based on the maximum values assumed in the plant safety 
analyses. The analyses conservatively time testing does not affect 
the capability of the associated systems to establish the margin of 
safety. The elimination of the selected response perform their 
intended function within the allowed response time used as the basis 
for plant safety analyses. Plant and system response to an 
initiating event will remain in compliance within the assumptions of 
the safety analyses, and therefore, the margin of safety is not 
affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan, Section Chief.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to revise the requirements 
for system operability during movement of recently irradiated fuel 
assemblies in the secondary containment. Specifically, the 
Applicability of TS 3.3.7.1, ``Control Room Emergency Filtration (CREF) 
System Instrumentation,'' 3.7.3, ``Control Room Emergency Filtration 
(CREF) System,'' and 3.7.4, ``Control Center Air Conditioning (AC) 
System,'' during movement of recently irradiated fuel assemblies would 
be deleted.
    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.
    This License Amendment involves changes in the requirements for 
the operability of the CREF system, CREF system instrumentation, and 
Control Center Air Conditioning (AC) system. The functions of these 
systems provide configurations for mitigating the consequences of 
radiological accidents; however, they do not involve the initiation 
of any previously analyzed accident. Therefore, the proposed changes 
cannot increase the probability of any previously evaluated 
accident.
    The analysis of the Fuel Handling Accident (FHA) concludes that 
radiological consequences are within the regulatory acceptance 
criteria. The FHA analysis includes evaluations of the radiological 
consequences resulting from a limiting drop of a fuel assembly, 
using the Alternative Source Term (AST) and the Regulatory Guide 
1.25 methodologies, over the reactor core. The radiological 
consequences associated with this scenario, assuming no mitigation 
credit for the CREF System, have been shown to satisfy the 
regulatory acceptance criteria. Therefore, the proposed changes do 
not significantly increase the radiological consequences of any 
previously evaluated accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter the design function or 
operation of the systems involved. The CREF system will still 
provide protection to control room occupants in the case of a 
significant radioactive release. The revised Technical Specification 
(TS) requirements are supported by the FHA analysis. The 
radiological consequences of a FHA under the proposed TS 
requirements are well below the regulatory limits. The proposed 
changes do not introduce any new modes of plant operation and do not 
involve physical modifications to the plant. The original Licensing 
Basis for the FHA took no credit for CREF system mitigation. 
Therefore, the proposed changes do not create the potential for a 
new or different kind of accident from any accident previously 
evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed changes to the Fermi 2 TS requirements are 
supported by the design basis analysis and are established such that 
the radiological consequences are below the regulatory guidelines. 
Safety margins and analytical conservatisms are retained to ensure 
that the analysis adequately bounds all postulated event scenarios. 
The proposed TS requirements continue to ensure that the 
radiological consequences at both the control room and the exclusion 
area and low population zone boundaries are below the corresponding 
regulatory guidelines; therefore, the proposed changes will not 
result in a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Section Chief: L. Raghavan, Section Chief.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: May 7, 2002.
    Description of amendment request: The proposed amendment would 
change Technical Specifications (TSs) 2.2, ``Limiting Safety System 
Settings'' and 3/4.3, ``Instrumentation'' to more accurately reflect 
the existing plant design for the Reactor Protection System, the 
Engineered Safety Features Actuation System, and the Radiation 
Monitoring System instrumentation and to provide consistency within TS 
Tables 2.2-1, 3.3-1, and 4.3-1. Specifically, the proposed amendment 
would make the following changes:
    (1) The Reactor Coolant Pump Speed--low functional unit, also known 
as the Underspeed--Reactor Coolant Pumps functional unit, which is not 
credited by the facility accident analysis, would be deleted from the 
TSs.
    (2) The mode applicability for the Wide Range Logarithmic Neutron 
Flux Monitor functional unit would be revised consistent with a 
previously approved license amendment (Millstone Unit No. 2 License 
Amendment No. 38, dated April 19, 1978).
    (3) The Safety Limits And Limiting Safety System Settings TS would 
be revised for completeness and consistency with the Reactor Protection 
System Instrumentation TS to include those functional units which do 
not have specific trip or allowable values.
    (4) The Reactor Protection System Instrumentation TS would be 
revised to include operability requirements for the Reactor Protection 
System Logic functional unit.
    (5) The Reactor Protection System Instrumentation TS would be 
revised to

[[Page 42820]]

include operability requirements for the Reactor Trip Breakers 
functional unit.
    (6) The Engineered Safety Feature Actuation System Instrumentation 
TS would be revised to include operability, trip setpoint, and 
surveillance requirements for the Automatic Actuation Logic, as 
applicable, associated with the Safety Injection, Containment Spray, 
Containment Isolation, Main Steam Isolation, Enclosure Building 
Filtration, Containment Sump Recirculation, Loss of Power, and 
Auxiliary Feedwater functional units.
    (7) The Engineered Safety Feature Actuation System Instrumentation 
TS action statement for the Auxiliary Feedwater manual actuation 
functional unit would be revised such that the required actions are 
consistent with the applicability of the TS.
    (8) The Engineered Safety Feature Actuation System Instrumentation 
table, which identifies Engineered Safety Features Trip Values, would 
be revised for completeness and consistency to include those functional 
units which do not have specific trip or allowable values.
    (9) The Radiation Monitoring Instrumentation TS would be revised to 
include a new surveillance requirement which would verify that the 
response time for the control room isolation function is consistent 
with facility accident analysis assumptions.
    (10) The Noble Gas Effluent Monitor (high range) (Unit 2 stack) 
functional unit would be relocated within the applicable TS as a 
process monitor, consistent with its current (and original) design 
function.
    (11) The Remote Shutdown Instrumentation TS would be revised 
consistent with standard practices for TS format such that the action 
statement would not be entered unless the minimum channels of remote 
shutdown instrumentation that are required to be operable, as defined 
by this specification, are not maintained.
    (12) The Remote Shutdown Instrumentation TS would be revised by 
extending the restoration period for an inoperable channel of remote 
shutdown instrumentation from 7 days to 31 days.
    The TS Bases would also be revised, as applicable, to reflect these 
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 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes would not alter the way any structure, system, 
or component functions and would not alter the manner in which the 
plant is operated. There are no hardware changes associated with the 
proposed changes. Therefore, the Reactor Protection System, the 
Engineered Safety Features Actuation System, and the Radiation 
Monitoring System instrumentation would continue to perform within the 
bounds of the previously performed accident analyses. The proposed 
changes to the operability requirements would not affect the 
instrumentation's ability to mitigate the design-basis accidents. The 
design-basis accidents would remain the same postulated events 
described in the Millstone Unit No. 2 Final Safety Analysis Report, and 
the consequences of these events will not be affected. Therefore, the 
proposed changes would 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 would not alter the plant configuration (no 
new or different type of equipment would be installed) or require any 
new or unusual operator actions. The proposed changes would not alter 
the way any structure, system, or component functions and would not 
alter the manner in which the plant is operated. The proposed changes 
would not introduce any new failure modes. Therefore, the proposed 
changes would 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 would not reduce the margin of safety since 
the changes have no impact on any accident analysis assumption. The 
proposed changes would not decrease the scope of equipment currently 
required to be operable or subject to surveillance testing, nor would 
the proposed changes affect any instrument setpoints or equipment 
safety functions. The proposed changes would not alter the operation of 
any component or system, nor would the proposed changes affect any 
safety limits or safety system settings which are credited in a 
facility accident analysis. Therefore, the proposed changes would not 
result in a reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: 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. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: April 18, 2002.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to increase the boron concentration 
in the spent fuel pool from 730 ppm to 850 ppm, reduce the Boraflex 
credit from 50 percent to 40 percent, and change the storage criteria, 
fuel enrichment, and burnup requirements for Region 2A of this spent 
fuel pool.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No, based upon the following:

Dropped Fuel Assembly

    There is no significant increase in the probability of a fuel 
assembly drop accident in the spent fuel pools when considering the 
degradation of the Boraflex panels in the spent fuel pool racks 
coupled with the presence of soluble boron in the spent fuel pool 
water for criticality control. The handling of the fuel assemblies 
in the spent fuel pool has always been performed in borated water, 
and the quantity of Boraflex remaining in the racks has no effect on 
the probability of such a drop accident.
    The criticality analysis showed that the consequences of a fuel 
assembly drop accident in the spent fuel pools are not affected when 
considering the degradation of the Boraflex in the spent fuel pool 
racks and the presence of soluble boron.

Fuel Misloading

    There is no significant increase in the probability of the 
accidental misloading of spent fuel assemblies into the spent fuel 
pool racks when considering the degradation of the Boraflex in the 
spent fuel pool racks and the presence of soluble boron in the pool 
water for criticality control. Fuel assembly placement and storage 
will continue to be controlled pursuant to approved fuel handling 
procedures to ensure compliance with the Technical Specification 
requirements. These procedures will be revised as needed to comply 
with the revised Region 2A requirements which would be imposed by 
the proposed Technical

[[Page 42821]]

Specification changes. These revised storage limits were developed 
with input from station personnel. Their awareness, in conjunction 
with any procedure changes as described above, will provide 
additional assurance that an accidental misloading of a spent fuel 
assembly should not occur.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the spent fuel pool racks 
because criticality analyses demonstrate that the pool will remain 
subcritical following an accidental misloading if the pool contains 
an adequate soluble boron concentration. Current Technical 
Specification 3.7.14 will ensure that an adequate spent fuel pool 
boron concentration is maintained in the McGuire spent fuel storage 
pools. The McGuire Station UFSAR Chapter 16, ``Selected Licensee 
Commitments,'' provides for adequate monitoring of the remaining 
Boraflex in the spent fuel pool racks. If that monitoring identifies 
further reductions in the Boraflex panels which would not support 
the conclusions of the McGuire Criticality Analysis, then the 
McGuire TSs and design bases would be revised as needed to ensure 
that acceptable subcriticality are maintained in the McGuire spent 
fuel storage pools.

Significant Change in Spent Fuel Pool Temperature

    There is no significant increase in the probability of either 
the loss of normal cooling to the spent fuel pool water or a 
decrease in pool water temperature from a large emergency makeup 
when considering the degradation of the Boraflex in the spent fuel 
pool racks and the presence of soluble boron in the pool water for 
subcriticality control since a high concentration of soluble boron 
has always been maintained in the spent fuel pool water. Current 
Technical Specification 3.7.14 will ensure that an adequate spent 
fuel pool boron concentration is maintained in the McGuire spent 
fuel storage pools.
    A loss of normal cooling to the spent fuel pool water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in water density that would 
result in a decrease in reactivity when Boraflex neutron absorber 
panels are present in the racks. However, since a reduction in the 
amount of Boraflex present in the Region 2A racks is considered, and 
the spent fuel pool water has a high concentration of boron, a 
density decreases causes a positive reactivity addition. However, 
the additional negative reactivity provided by the current boron 
concentration limit, above that provided by the concentration 
required to maintain keff less than or equal to 0.95 
(1470 ppm), will compensate for the increased reactivity which could 
result from a loss of spent fuel pool cooling event. Because 
adequate soluble boron will be maintained in the spent fuel pool 
water, the consequences of a loss of normal cooling to the spent 
fuel pool will not be increased. Current Technical Specification 
3.7.14 will ensure that an adequate spent fuel pool boron 
concentration is maintained in the McGuire spent fuel storage pools.
    A decrease in pool water temperature from a large emergency 
makeup causes an increase in water density that would result in an 
increase in reactivity when Boraflex neutron absorber panels are 
present in the racks. However, the additional negative reactivity 
provided by the current boron concentration limit, above that 
provided by the concentration required to maintain keff 
less than or equal to 0.95 (1470 ppm), will compensate for the 
increased reactivity which could result from a decrease in spent 
fuel pool water temperature. Because adequate soluble boron will be 
maintained in the spent fuel pool water, the consequences of a 
decrease in pool water temperature will not be increased. Current 
Technical Specification 3.7.14 will ensure that an adequate spent 
fuel pool boron concentration is maintained in the McGuire spent 
fuel storage pools.
    2. Will the change create the possibility of a new or different 
kind of accident from any previously evaluated?
    No. Criticality accidents in the spent fuel pool are not new or 
different types of accidents. They have been analyzed in Section 
9.1.2.3 of the Updated Final Safety Analysis Report and in 
Criticality Analysis reports associated with specific licensing 
amendments for fuel enrichments up to 4.75 weight percent U-235. 
Specific accidents considered and evaluated include fuel assembly 
drop, accidental misloading of spent fuel assemblies into the spent 
fuel pool racks, and significant changes in spent fuel pool water 
temperature. The accident analysis in the Updated Final Safety 
Analysis Report remains bounding.
    The possibility for creating a new or different kind of accident 
is not credible. The amendment proposes to take credit for the 
soluble boron in the spent fuel pool water for reactivity control in 
the spent fuel pool while maintaining the necessary margin of 
safety. Because soluble boron has always been present in the spent 
fuel pool, a dilution of the spent fuel pool soluble boron has 
always been a possibility; however, a criticality accident resulting 
from a dilution accident was not considered credible. A spent fuel 
pool dilution evaluation * * * has demonstrated that a dilution of 
the boron concentration in the spent fuel pool water which could 
increase the rack keff to greater than 0.95 (constituting 
a reduction of the required margin to criticality) is not a credible 
event. The requirement to maintain a revised minimum boron 
concentration in the spent fuel pool water for reactivity control 
(at least 850 ppm) will have no effect on normal pool operations and 
maintenance. There are no changes in equipment design or in plant 
configuration. This revised requirement will not result in the 
installation of any new equipment or modification of any existing 
equipment. Therefore, the proposed amendment will not result in the 
possibility of a new or different kind of accident.
    3. Will the change involve a significant reduction in a margin 
of safety?
    No. The proposed Technical Specification changes and the 
resulting McGuire Region 2A spent fuel storage operating limits will 
provide adequate safety margin to ensure that the stored fuel 
assembly array will always remain subcritical. Those revised limits 
are based on a plant specific criticality analysis * * * based on 
the ``Westinghouse Spent Fuel Rack Criticality Analysis 
Methodology'' * * * The Westinghouse methodology for taking credit 
for soluble boron in the spent fuel pool has been reviewed and 
approved by the NRC * * * This methodology takes partial credit for 
soluble boron in the spent fuel pool and requires conformance with 
the following NRC acceptance criteria for preventing criticality 
outside the reactor:
    (1) keff shall be less than 1.0 if fully flooded with 
unborated water which includes an allowance for uncertainties at a 
95% probability, 95% confidence (95/95) level; and
    (2) keff shall be less than or equal to 0.95 if fully 
flooded with borated water, which includes an allowance for 
uncertainties at a 95/95 level.
    The criticality analysis utilized credit for soluble boron to 
ensure keff will be less than or equal to 0.95 under 
normal circumstances, and storage configurations have been defined 
using a 95/95 keff calculation to ensure that the spent 
fuel rack keff will be less than 1.0 with no soluble 
boron. Soluble boron credit is used to provide safety margin by 
maintaining keff less than or equal to 0.95 including 
uncertainties, tolerances and accident conditions in the presence of 
spent fuel pool soluble boron. The loss of substantial amounts of 
soluble boron from the spent fuel pool which could lead to exceeding 
a keff of 0.95 has been evaluated * * * and shown to be 
not credible. Accordingly, the required margin to criticality is not 
reduced.
    Previous evaluations * * * have shown that the dilution of the 
spent fuel pool boron concentration from the conservative assumed 
initial boron concentration (2475 ppm) to the minimum boron 
concentration required to maintain keff [le] 0.95 (850 
ppm) is not credible. The dilution analyses, along with the 95/95 
criticality calculation which shows that the spent fuel rack 
keff will remain less than 1.0 when flooded with 
unborated water, provide a level of safety comparable to the 
conservative criticality analysis methodology* * *
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the facility's margin of 
safety.

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

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

    Date of amendment request: April 19, 2002.
    Description of amendment request: The proposed change revises 
Technical

[[Page 42822]]

Specification (TS) 5.5.10, ``Technical Specification (TS) Bases Control 
Program,'' to provide consistency with changes to 10 CFR 50.59 as 
published in the Federal Register (64 FR 53582) on October 4, 1999, 
that became effective March 13, 2001. The proposed changes to TS 5.5.10 
are made to incorporate the change made in 10 CFR 50.59 to remove the 
phrase ``unreviewed safety question.'' The proposed changes to TS 
5.5.10 are consistent with NRC approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler TSTF-364, 
Revision 0, as amended by the Westinghouse Owners Group (WOG) editorial 
change WOG-ED-24.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change deletes the reference to ``unreviewed safety 
question'' as defined in 10 CFR 50.59. Deletion of the definition of 
``unreviewed safety question'' was approved by the NRC with the 
revision of 10 CFR 50.59. This change is administrative in nature. 
Consequently, the probability of an accident previously evaluated is 
not significantly increased. Changes to the TS Bases are still 
evaluated in accordance with 10 CFR 50.59. As a result, the 
probability or consequences of any accident previously evaluated are 
not significantly affected. There is no increase in the radiological 
dose at the site boundary for any previously evaluated accident. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. These changes are considered administrative in nature and 
do not modify, add, delete, or relocate any technical requirements 
in the TS. Therefore, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes will not reduce a margin of safety because 
it has no direct effect on any safety analyses assumptions. Changes 
to the TS Bases that result in meeting the criteria in paragraph 10 
CFR 50.59(c)(2) continue to require NRC approval pursuant to 10 CFR 
50.59. This change is administrative in nature based on the revision 
to 10 CFR 50.59. 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 14, 2002
    Description of amendment request: The proposed changes will amend 
the Operating License to revise the as-found safety function lift 
setpoint tolerances for the Safety and Relief Valves (S/RVs) for River 
Bend Station, Unit 1. The proposed amendment does not change the actual 
setpoint or the way the S/RVs are operated, would be limited to the 
lower tolerances and would not affect the upper limits, and would only 
apply to the as-found tolerance and not to the as-left tolerance which 
will remain unchanged. The as-found tolerances are used for determining 
operability and to increase sample sizes for testing. There will be no 
change to the valves as installed in the plant. The proposed amendment 
would also allow surveillance of the relief mode of operation of the S/
RVs without physically lifting the disk of a valve off the seat at 
power.
    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.
    These changes have no influence on the probability or 
consequences of any accident. The setpoint tolerance change does not 
[a]ffect the operation of valves that are installed in the plant or 
change the as-left tolerance which will remain at 1%. 
The setpoint tolerances for valves that have been tested or 
refurbished are not being changed. The change only has an [a]ffect 
on increased sampling for operability and for IST [in-service 
testing] purposes. The change to the tolerance only affects the 
lower limit for opening the valve and does not change the upper 
limit which is the limit that protects from overpressurization.
    There is no increase in the probability or consequences of any 
accident based on the changes to the remote actuation testing of the 
valves because the valve opening capability will continue to be 
bench tested and the actuator will be tested independently. The open 
and close capabilities will therefore be demonstrated 
satisfactorily.
    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 or different accidents are created because the proposed 
changes do not change the configuration or operation of the plant in 
any way. The setpoint tolerance changes only affect the criteria 
that determines when a valve test is considered to be a failure and 
is limited to the lower limit. It does not change the criteria for 
the upper limit that protects against overpressurization.
    The changes to the remote actuation testing continue to provide 
assurance that the valves have open and close capabilities and 
remain consistent with the intent of the present surveillance.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not change the configuration or 
operation of the plant in any way. The setpoint tolerance changes 
only affect the criteria that determines when a valve test is 
considered to be a failure and is limited to the lower limit. It 
does not change the criteria for the upper limit that protects 
against overpressurization.
    The changes to the remote actuation testing continue to provide 
assurance that the valve has open and close capability and is 
consistent with the intent of the present surveillance.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

[[Page 42823]]

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

    Date of amendment request: May 14, 2002.
    Description of amendment request: Entergy Operations, Inc. 
(Entergy) requests changes to the Degraded Voltage--Voltage basis and 
loss-of-coolant accident (LOCA) time delay allowable values (Technical 
Specification Table 3.3.8.1-1, Items 1.c and 1.e; and Items 2.c and 
2.e) to reflect the results of new calculations performed in 
association with a design basis reconstitution.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The change in the degraded voltage protection voltage and time 
delay allowable values allows the protection scheme to function as 
originally designed. The proposed allowable values ensure that the 
Class 1E distribution system remains connected to the offsite power 
system when adequate offsite voltage is available and motor starting 
transients are considered. Replacement of the Division 1 and 2 
degraded voltage relays provide operational flexibility to 
accommodate the proposed protection voltage allowable values, which 
are more conservative than the current limits. Calculations have 
demonstrated that adequate margin is present to support the decrease 
in the minimum allowable Division 3 degraded voltage. The small 
increase in the time delay allowable values more accurately reflects 
the actual load sequencing experienced during an accident condition. 
The proposed time delay continues to provide equipment protection 
while preventing a premature separation from offsite power. The 
diesel start due to a Loss of Coolant Accident signal is not 
impacted by this change. During an actual degraded voltage 
condition, the degraded voltage time delays will continue to isolate 
the Class 1E distribution system from offsite power before the 
diesel is ready to assume the emergency loads, which is the limiting 
time basis for mitigating system responses to the accident. For this 
reason, the existing Loss of Power / Loss of Coolant accident 
analysis continues to be valid.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change involves the revision of degraded voltage 
protection voltage and time delay allowable values to satisfy 
existing design requirements. Component replacement necessary to 
support these new values will be performed in accordance with plant 
procedures, which ensure adherence with all quality requirements. No 
additional failure mechanisms are introduced as a result of the 
changes to the allowable values.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed protection voltage allowable values are low enough 
to prevent inadvertent power supply transfer, but high enough to 
ensure that sufficient voltage is available to the required 
equipment. The small increase in the time delay allowable values 
more accurately reflects the actual load sequencing experienced 
during an accident condition. The proposed time delay continues to 
provide equipment protection while preventing a premature separation 
from offsite power. The diesel start due to a Loss of Coolant 
Accident signal is not impacted by this change. During an actual 
degraded voltage condition, the degraded voltage time delays will 
continue to isolate the Class 1E distribution system from offsite 
power before the diesel is ready to assume the emergency loads.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 14, 2002.
    Description of amendment request: Entergy Operations, Inc. is 
proposing to revise the River Bend Station, Unit 1 (RBS), 
Administrative Technical Specifications (TSs) regarding containment 
leak rate testing. The proposed change will revise RBS Administrative 
TS 5.5.13 to add an exception to the commitment to follow the 
guidelines for Regulatory Guide 1.163, ``Performance-Based Containment 
Leak-Test Program.'' The exception is taken to the interval guidance in 
NEI 94-01, Revision 0, ``Industry Guideline for Implementing 
Performance-Base Option of 10CFR50, Appendix J.'' The effect of this 
request will be a one-time extension of the interval between tests from 
10 years to 15 years.
    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.
    10CFR50, Appendix J was amended to incorporate provisions for 
performance-based testing in 1995. The proposed amendment to 
Technical Specification (TS) 5.5.13 adds a one-time extension to the 
current interval for Type A testing (i.e., the integrated leak rate 
test). The current interval of ten years, based on past performance, 
would be extended on a one-time basis to 15-years from the date of 
the last test. The proposed extension to the Type A test cannot 
increase the probability of an accident since there are no design or 
operating changes involved and the test is not an accident 
initiator. The proposed extension of the test interval does not 
involve a significant increase in the consequences since research 
documented in NUREG-1493, ``Performance Based Containment Leak Rate 
Test Program,'' has found that, generally, fewer than 3% of the 
potential containment leak paths are not identified by Type B and C 
testing. A risk evaluation of the interval extension for RBS is 
consistent with these results. In addition, at RBS, the testing and 
containment inspections also provide a high degree of assurance that 
the containment will not degrade in a manner detectable only by a 
Type A test. Inspections required by the Maintenance Rule 
(10CFR50.65) and by the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code are performed to identify 
containment degradation that could affect leaktightness.
    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 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed extension to the interval for the Type A test does 
not involve any design or operational changes that could lead to a 
new or different kind of accident from any accidents previously 
evaluated. The test itself is not being modified, but is only 
intended to be performed after a longer interval. The proposed 
change does not involve a physical alteration of the plant (no new 
or different type of equipment will be installed) or a change in the 
methods governing normal plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different

[[Page 42824]]

kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The generic study of the increase in the Type A test interval, 
NUREG-1493, concluded there is an imperceptible increase in the 
plant risk associated with extending the test interval out to twenty 
years. Further, the extended test interval would have a minimal 
effect on this risk since Type B and C testing detect 97% of 
potential leakage paths. For the requested change in the RBS ILRT 
(integrated leak rate testing) interval, it was determined that the 
risk contribution of leakage will increase 0.32%. This change is 
considered very small and does not represent a significant reduction 
in the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 14, 2002.
    Description of amendment request: The proposed modification of the 
River Bend Station Technical Specifications is to revise several of the 
Surveillance Requirements (SRs) pertaining to testing of the Division 3 
standby diesel generator (DG) and manual transfer test for offsite 
circuits. The proposed change would modify specific restrictions 
associated with these SRs that prohibit performing required testing in 
Modes 1, 2, or 3. The affected SRs are SR 3.8.1.8, SR 3.8.1.9, SR 
3.8.1.10, SR 3.8.1.11, SR 3.8.1.12, SR 3.8.1.13, SR 3.8.1.16, SR 
3.8.1.17, SR 3.8.1.18, and SR 3.8.1.19.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The DG and its associated emergency loads are accident 
mitigating features, not accident initiating equipment. Therefore, 
there will be no impact on any accident probabilities by the 
approval of the requested amendment.
    The design of plant equipment is not being modified by these 
proposed changes. As such, the ability of the DG to respond to a 
design basis accident will not be adversely impacted by these 
proposed changes. The capability of the DG to supply power in a 
timely manner will not be compromised by permitting performance of 
DG testing during periods of power operation. Additionally, limiting 
testing to only one DG at a time ensures that design basis 
requirements for backup power is met, should a fault occur on the 
tested DG. Therefore, there would be no significant impact on any 
accident consequences.
    Based on the above, the proposed change to permit certain DG 
surveillance tests to be performed during plant operation will have 
no [a]ffect on accident probabilities or consequences. 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 causal mechanisms would be created as a result 
of NRC [Nuclear Regulatory Commission] approval of this amendment 
request since no changes are being made to the plant that would 
introduce any new accident causal mechanisms. Equipment will be 
operated in the same configuration with the exception of the plant 
mode in which the testing is conducted. This amendment request does 
not impact any plant systems that are accident initiators; neither 
does it adversely impact any accident mitigating systems.
    Based on the above, implementation of the proposed changes would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes to the testing requirements for the DG 
do not affect the operability requirements for the DG, as 
verification of such operability will continue to be performed as 
required. Continued verification of operability supports the 
capability of the DG to perform its required function of providing 
emergency power to plant equipment that supports or constitutes the 
fission product barriers. Consequently, the performance of these 
fission product barriers will not be impacted by implementation of 
this proposed amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 30, 2002.
    Description of amendment request: The proposed amendment would 
revise the requirements in several administrative programs in Technical 
Specification (TS) Section 6.0, ``Administrative Controls.'' 
Specifically, the proposed amendment would: (1) Replace the specific 
management titles for several organizational positions with generic 
titles, (2) replace the title of the Quality Assurance Program 
Description with a reference to the quality assurance program described 
or referenced in the Updated Final Safety Analysis Report (UFSAR), and 
(3) delete the functions of the Station Nuclear Safety and the Nuclear 
Facilities Safety Committees and the Vice President-Nuclear Power since 
their duties and responsibilities are described in the Quality 
Assurance Program Description. The proposed changes reflect the 
organizational integration at the Indian Point Energy Center.
    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 license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The proposed change eliminates the redundant controls on 
elements of the managerial and administrative controls implemented 
by the quality assurance program described or referenced in the 
UFSAR. There are no changes proposed to the design, operation, 
maintenance or testing

[[Page 42825]]

of the plant's systems, structures or components. Therefore, the 
assumptions of the operability or performance of systems, 
structures, or components in accident analyses are unchanged.
    The adequacy of the managerial and administrative controls used 
to assure safe operation were previously accepted by the NRC 
[Nuclear Regulatory Commission] in their approval of the quality 
assurance program description. The changes to the existing controls 
were evaluated under 10 CFR 50.54 to ensure the changes would not 
reduce the commitments in the quality assurance program description 
previously accepted by the NRC. Therefore, there is no increase in 
the probability or in the 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?
    The proposed changes do not affect the design, operation, 
maintenance, or testing of a plant system, structure or component. 
No new or unanalyzed conditions can be created through the proposed 
replacement of specific administrative position titles with generic 
position titles, since the authority, responsibility and 
qualification for each required position are specified in the 
quality assurance program described or referenced in the UFSAR.
    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 amendment involve a significant reduction 
in a margin of safety?
    The proposed changes do not affect a design function or 
operation of any plant structure, system, or component. The change 
does not affect the method of ENO's [Entergy Nuclear Operations'] 
compliance with any regulation. The changes to the quality assurance 
program as described or referenced in the UFSAR were evaluated under 
10 CFR 50.54 and it was determined that the changes do not reduce 
any commitments from the quality assurance program description that 
was previously evaluated and accepted by the NRC.
    Therefore, the proposed changes do not result in a change to any 
of the safety analyses or any margin of safety.

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

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

    Date of amendment request: March 19, 2002.
    Description of amendment request: The proposed amendment would 
revise the method of controlling the fuel cycle unfavorable exposure 
time (UET) related to an anticipated transient without scram (ATWS) 
event. The current methodology controls UET by limiting the value of 
the moderator temperature coefficient (MTC) inherent in the reactor 
core design. The proposed license amendment would utilize the 
Configuration Risk Management Program to administratively control the 
availability of ATWS risk significant equipment to minimize core UET. 
By removing the UET MTC constraint, reload cores may be designed with a 
more positive MTC as allowed by the TS, therefore resulting in 
significant benefits including reduced fuel cost, reduced outage time, 
and reduced amount of spent fuel.
    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 proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The change in the methodology of controlling the UET associated 
with an ATWS event will not increase the probability of any accident 
previously evaluated, including an ATWS event. All systems, 
including the existing ATWS Mitigating Systems Actuation Circuitry 
(AMSAC), will continue to be operated in accordance with current 
design requirements, and no new components or system interactions 
have been identified that could lead to an increase in the 
probability of any accident previously evaluated in the UFSAR.
    Currently, the UET for a given fuel cycle must be less than 5% 
of the operating cycle under a ``base case'' set of plant conditions 
(i.e., 100% power-operated relief valve (PORV) capacity, 100% AFW 
system availability, no control rod insertion capability, and AMSAC 
available). The proposed license amendment would replace the 5% fuel 
cycle limit on UET with the requirement to administratively control 
ATWS risk significant equipment when core conditions are 
``unfavorable'' over the entire operating cycle. The goal of the 
administrative control program is to minimize the UET at all times. 
The methodology used to determine the UET will remain the same as 
the currently approved methodology. The Configuration Risk 
Management Program (CRMP), currently described in the Byron Station 
and Braidwood Station Technical Requirements Manual (TRM), Appendix 
T, will be used to manage the availability of ATWS risk significant 
equipment. The CRMP will provide a proceduralized process to perform 
a configuration risk assessment of the plant equipment configuration 
and availability prior to planned on-line maintenance of the ATWS 
risk significant equipment and/or functions. The CRMP is currently 
used as a tool to manage maintenance activities to minimize any 
increase in the consequences of an abnormal event or accident. 
Development of the Byron Station and Braidwood Station CRMP is 
consistent with 10 CFR 50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants,'' paragraph 
(a)(4), and is governed by Work Control Procedure, WC-AA-101, ``On-
Line Work Control Process.''
    The ATWS risk significant equipment which will be monitored by 
the CRMP includes the:
     Rod control system;
     AFW system;
     Pressurizer PORVs; and
     ATWS Mitigating Systems Actuation Circuitry (AMSAC)
    This change in methodology will also have no effect on the 
consequences of any accident previously evaluated including an ATWS 
event. Should an ATWS occur during an ``unfavorable'' fuel cycle 
period, the consequences of this event will remain unchanged under 
the new methodology which only administratively controls plant 
equipment availability associated with the UET. Also, the 
consequences of an ATWS event with the core designed with a more 
positive MTC remain acceptable. Although the time to RCS [reactor 
coolant system] overpressure and resultant loss-of-coolant accident 
(LOCA) may decrease, the consequences of the LOCA remain unchanged.
    Based on this evaluation, it is concluded that the proposed TS 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The configuration, operation and accident response of the Byron 
Station and the Braidwood Station systems, structures or components 
are unchanged by the proposed TS change which would utilize an 
alternate method of controlling the UET of a fuel cycle. No 
transient event would result in a new sequence of events that could 
lead to a new accident scenario.
    No new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode was identified as a result of 
utilizing the CRMP to monitor ATWS risk significant equipment. In 
addition, this methodology does not create any new failure modes 
that could lead to a different kind of accident. Software changes to 
the existing CRMP will be made to monitor the above mentioned ATWS 
risk significant equipment.

[[Page 42826]]

    Based on this analysis, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed change. The proposed TS 
change does not have an adverse effect on any safety-related system. 
Therefore, the proposed TS change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The newly proposed methodology of monitoring and controlling the 
UET during an operating cycle is more conservative than the 
currently approved method and; therefore, will increase the margin 
of safety.
    Currently, the UET for a given fuel cycle is limited to less 
than 5% of the operating cycle and is only evaluated for a ``base 
case'' set of plant conditions (i.e., 100% PORV capacity, 100% AFW 
system availability, no control rod insertion capability, and AMSAC 
available). The UET is currently limited by constraining the value 
of the MTC inherent in the reload reactor core design.
    The proposed methodology will utilize the CRMP as a tool to 
monitor the availability of ATWS risk significant equipment during 
the entire operating cycle. By effectively managing the planned on-
line maintenance of ATWS risk significant equipment, the cycle UET 
will be minimized at all times. This methodology also analyzes 
different combinations of ATWS risk significant equipment 
availability in addition to the ``base case'' conditions. The 
proposed license amendment would replace the 5% fuel cycle limit on 
UET with the requirement to administratively control ATWS risk 
significant equipment when core conditions are ``unfavorable'' over 
the entire operating cycle. The goal of the administrative program 
is to minimize the UET at all times. The methodology used to 
determine the UET will remain the same as the currently approved 
methodology. The Configuration Risk Management Program (CRMP) 
currently described in the Byron Station and Braidwood Station 
Technical Requirements Manual (TRM), Appendix T, will be used to 
manage the availability of ATWS risk significant equipment. The CRMP 
will provide a proceduralized process to perform a configuration 
risk assessment of the plant equipment configuration and 
availability prior to planned on-line maintenance of the ATWS risk 
significant equipment and/or functions. The CRMP is currently used 
as a tool to manage maintenance activities to minimize any increase 
in the consequences of an abnormal event or accident. Development of 
the Byron Station and Braidwood Station CRMP is consistent with 10 
CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' paragraph (a)(4), and is 
governed by Work Control Procedure, WC-AA-101, ``On-Line Work 
Control Process.''
    Based on this evaluation, the proposed TS changes do not involve 
a significant reduction in a margin of safety.
    Based upon the above analyses and evaluations, we have concluded 
that the proposed change to the TS involve no 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 5, 2002.
    Description of amendment request: The amendment would revise the 
Improved Technical Specifications to increase the maximum allowed rated 
thermal power for Crystal River Unit 3 from 2544 MegaWatts-thermal 
(MWt) to 2568 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 not involve a significant increase in the probability 
or consequences of an accident previously analyzed.
    The proposed change will increase the maximum core power level 
from 2544 MWt to 2568 MWt. This increase will only require 
adjustments and calibrations of existing plant instrumentation and 
control systems. No hardware upgrades or equipment replacements are 
needed to implement the proposed change.
    Nuclear steam supply systems (NSSS) and balance-of-plant (BOP) 
systems and components that could be affected by the proposed change 
have been evaluated using revised NSSS design parameters based on a 
core power level of 2568 MWt. The results of these evaluations, 
which used well-defined analysis input assumptions/parameter values 
and currently approved analytical techniques, indicate that CR-3 
systems and components will continue to function within their design 
parameters and remain capable of performing their required safety 
functions at 2568 MWt. Since the revised NSSS parameters remain 
within the design conditions of the reactor coolant system (RCS) 
functional specification, the proposed change will not result in any 
new design transients or adversely affect the current CR-3 design 
transient analyses.
    The accidents analyzed in Chapter 14 of the CR-3 Final Safety 
Analysis Report (FSAR) have been reviewed for the impact of the 
uprate. Based on the power levels assumed in the current safety 
analyses, it has been determined that all FSAR and supporting 
analyses bound the uprate. This includes the dose calculations for 
the design basis radiological accidents, which assume a power level 
of 2619 MWt (2568 MWt plus an assumed 2 percent measurement 
uncertainty).
    Based on the above, the change will not increase the probability 
or consequences of an accident previously evaluated.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    As discussed above, no hardware upgrades or equipment 
replacements are required to implement the proposed change. All CR-3 
systems and components will continue to function within their design 
parameters and remain capable of performing their required safety 
functions. The proposed change does not impact current CR-3 design 
transients or introduce any new transients. The design, physical 
configuration and operation of the plant will not be changed; as a 
result, no new equipment failure modes will be introduced. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Does not involve a significant reduction in the margin of 
safety.
    Challenges to the fuel, reactor coolant system (RCS) pressure 
boundary and containment were evaluated for uprate conditions. Core 
analyses show that the implementation of the power uprate will 
continue to meet the current nuclear design basis. Impacts to 
components associated with RCS pressure boundary structural 
integrity, and factors such as pressure/temperature limits, vessel 
fluence, and pressurized thermal shock (PTS) were determined to be 
bounded by current analyses. Mass and energy release to the 
containment from a loss-of-coolant accident (LOCA) or main steam 
line break are also bounded by current analyses, which assume an 
initial power level of 2619 MWt.
    As discussed above, all systems will continue to operate within 
their design parameters and remain capable of performing their 
intended safety functions following implementation of the proposed 
change. Finally, the current CR-3 safety analyses, including the 
design basis radiological accident dose calculations, bound the 
uprate.
    Therefore, this 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Acting Section Chief: Kahtan N. Jabbour.

[[Page 42827]]

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

    Date of amendment request: May 22, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.9.1.11.b to add two NRC-approved 
topical reports to the Core Operating Limits Report (COLR) methodology 
list, and delete superseded reports. Also, the method of listing 
topical reports would be revised to be consistent with Technical 
Specifications Task Force (TSTF) 363, which has been approved by the 
NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment updates the list of COLR methodologies 
and would allow the use of two new NRC approved methodologies, EMF-
2310(P)(A), ``SRP [Standard Review Plan] Chapter 15 Non-LOCA [Loss-
of-Coolant Accident] Methodology for Pressurized Water Reactors 
[(PWR)],'' and EMF-2328 (P)(A), ``PWR Small Break LOCA Evaluation 
Model, S-RELAP5 Based,'' for the St. Lucie Unit 1 safety analyses. 
The proposed changes have no adverse impact on the operation of the 
plant and have no relevance to the accident initiators. There are no 
changes to the plant configuration, and thus the frequency of 
occurrence of previously analyzed accidents is not affected by the 
proposed changes.
    With the updated methodologies, the safety analysis would 
continue to meet the analysis acceptance criteria consistent with 
the design basis requirements. The proposed changes have no adverse 
effect on the safety analysis and thus would not involve a 
significant increase in the consequences of design basis accidents. 
Changes to the COLR limits would continue to be controlled per 
Generic Letter 88-16 under the provisions of 10 CFR 50.59 and the 
requirements of TS 6.9.1.11.c.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed amendment updates the list of approved 
methodologies in TS 6.9.1.11.b. These changes would not create the 
possibility of a new kind of accident since there is no change to 
the plant configuration, systems or components, which would create 
new failure modes. The modes of operation of the plant remain 
unchanged.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes have no adverse impact on the safety 
analysis. The changes proposed would continue to provide margin to 
the acceptance criteria for Specified Acceptable Fuel Design Limits 
(SAFDL), 10 CFR 50.46(b) requirements, primary and secondary 
overpressurization, peak containment pressure, potential radioactive 
releases, and existing limiting conditions for operation. The future 
use of updated approved methodologies would follow all design basis 
requirements to ensure that a safety margin to the acceptance 
criteria would continue to remain available at all power levels for 
operation of St. Lucie Unit 1.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) associated with refueling 
operations to remove the requirement for operability of certain systems 
(containment penetrations, spent fuel pool and shield building 
ventilation, and containment isolation) when handling fuel assemblies 
that have decayed a sufficient period of time such that dose 
consequences of the postulated fuel handling accident (FHA) remain 
below the limits of 10 CFR Part 100 and the NRC Standard Review Plan 
with these systems unavailable. The proposed changes are consistent 
with the Standard TS for Combustion Engineering plants and a portion of 
Nuclear Energy Institute TS Task Force change traveler TSTF-51, 
Revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to the St. Lucie Units 1 and 2 TSs 
incorporate line item improvements that are based on assumptions in 
the postulated fuel handling accident analyses. These proposed 
changes remove the applicability of TSs regarding operability of 
certain systems (containment penetrations, spent fuel pool and 
shield building ventilation, and containment isolation) when 
handling fuel assemblies that have decayed a sufficient period of 
time. The results of the FHA analyses demonstrate that sufficient 
radioactive decay has occurred after 72 hours such that the 
resulting dose consequences are well within the limits given in 10 
CFR 100 and within the limits given in the Standard Review Plan, 
NUREG-0800. The systems that have been included in these proposed 
changes will have administrative controls in place to assure that 
systems are available and can be promptly returned to operation to 
further reduce dose consequences. These administrative controls will 
include a single normal or contingency method to promptly close the 
primary or secondary containment penetrations. These prompt methods 
need not completely block the penetrations nor be capable of 
resisting pressure, but are to enable the ventilation systems to 
draw the release from the postulated FHA such that it can be treated 
and monitored. This will result in lower doses than those calculated 
for the FHA.
    The equipment or systems involved are not initiators of an 
accident. Operability of these systems or equipment during fuel 
movement and/or core alterations has no affect on the probability of 
any accident previously evaluated.
    The proposed changes do not significantly increase the 
consequences of a fuel handling accident as previously evaluated. 
The calculated doses are well within the limits given in 10 CFR Part 
100 and within the limits given in the Standard Review Plan, NUREG-
0800. In addition, the calculated doses are larger than the expected 
doses because the calculations do not credit any filtration or 
containment of the source term that will occur by the administrative 
controls that will be in place.
    The changes being proposed do not affect assumptions contained 
in other plant safety analyses or the physical design of the plant, 
nor do they affect other TSs that preserve safety analysis 
assumptions. Therefore, operation of the facility in accordance with 
the proposed amendments would not involve a significant increase in 
the probability or consequences of an accident previously analyzed.

[[Page 42828]]

    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The proposed changes to the TSs do not affect or create a 
different type of fuel handling accident. The fuel handling accident 
analyses assume that all of the iodine and noble gases that become 
airborne, escape, and reach the exclusion area boundary and low 
population zone with no credit taken for filtration, containment of 
the source term, or for decay or deposition. The proposed changes do 
not involve the addition or modification of equipment nor do they 
alter the design of plant systems. The revised operations are 
consistent with the fuel handling accident analyses. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The calculated doses are well within the limits given in 10 CFR 
Part 100 and within the limits given in the Standard Review Plan, 
NUREG-0800. The proposed changes do not alter the bases for 
assurance that safety-related activities are performed correctly or 
the basis for any TS that is related to the establishment of or 
maintenance of a safety margin.
    The systems that have been included in the proposed change will 
have administrative controls in place to assure that the systems are 
available and can be promptly returned to operation to further 
reduce dose consequences. These administrative controls will include 
a single normal or contingency method to promptly close the primary 
or secondary containment penetrations. These prompt methods need not 
completely block the penetrations nor be capable of resisting 
pressure, but are to enable the ventilation systems to draw the 
release from the postulated FHA such that it can be treated and 
monitored.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: July 30, 2001.
    Description of amendment request: Proposed amendment revises the 
Cooper Nuclear Station licensing basis with respect to containment 
overpressure contribution to emergency core cooling system pump net 
positive suction head.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below. The requested amendment:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The requested license amendment does not result in any new 
accident initiators, nor are there changes being proposed to other 
plant systems or equipment postulated to initiate an accident 
previously evaluated. Thus, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated in the USAR [Updated Safety Analysis Report].
    The containment overpressure evaluation conservatively 
demonstrates that adequate margin between the available containment 
overpressure and the overpressure required to assure adequate low 
pressure ECCS [emergency core cooling system] pump NPSH [net 
positive suction head] are such that ECCS pump operation, as 
credited in the CNS [Cooper Nuclear Station] accident analysis, 
remains unchanged. Thus, the ECCS pumps continue to be available to 
perform the safety functions previously evaluated, and the proposed 
change does not involve a significant increase in the consequences 
of an accident previously evaluated in the USAR.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The proposed license amendment does not introduce any new 
equipment or hardware changes. The only equipment affected by this 
license amendment are the low pressure ECCS pumps. These pumps 
retain their ability to function following any accident previously 
evaluated and no new accidents are created as a result of increased 
reliance on overpressure or methodology changes. Thus, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated in the USAR.
    3. Does not create a significant reduction in the margin of 
safety.
    Although there is an increased reliance on containment 
overpressure, adequate low pressure ECCS pump NPSH is assured, and 
sufficient margin is conservatively determined to be maintained 
between the available overpressure and the required overpressure to 
provide confidence that the ECCS pumps will operate as required. The 
calculations are revised to show an increased absolute containment 
overpressure consideration from [sim]5 psi (original license 
application) to [sim]9.5 psi at the time of the peak suppression 
pool temperatures following a design basis LOCA [loss-of-coolant 
accident]. At this containment overpressure, the CS [core spray] and 
RHR [residual heat removal] pumps will utilize only [sim]4.45 psi 
and [sim]6.47 psi, respectively, of the available overpressure. This 
provides a margin of [sim]5 psi and [sim]3 psi, respectively, for 
the CS and RHR pumps at the peak suppression pool temperature. The 
calculations also address both short-term and long-term reliance on 
containment overpressure.
    In the short-term (<600 seconds), the RHR pumps do not depend on 
containment overpressure for adequate NPSH. However, during this 
short-term period following initiation of the event, the CS pump is 
conservatively calculated to require as much as [sim]4.94 psi of 
containment overpressure to assure adequate NPSH. At the time this 
overpressure is needed, [sim]6.85 psi of containment overpressure is 
available, providing a margin of [sim]1.9 psi. For the time periods 
following the peak suppression pool temperature, the required 
overpressure reliance reduces with time and suppression pool 
temperature.
    During the accident, beyond the time period of the peak 
suppression pool temperature, a minimum margin of [sim]0.6 psi is 
provided for ECCS pump NPSH. However, this minimum margin occurs 
just prior to 100 hours into the event at a point when no 
containment overpressure is required for ECCS pump NPSH. During 
times when containment overpressure is credited, there is a minimum 
of [sim]1 psi containment overpressure available.
    The analysis also utilizes three new methods for evaluation of 
the previously evaluated accidents. These are the SHEX code for the 
containment pressure and temperature response analysis, the ANS 5.1-
1979 model for determination of core decay heat, and the use of 
spatial evaluation of the suppression pool safety relief valve 
discharge quenchers relative to the ECCS pump intake strainers for 
prevention of steam bubble ingestion. A benchmark evaluation of the 
SHEX code is provided which indicates that the results are 
comparable to previous analysis. The ANS 5.1-1979 model is less 
conservative than the previously used May-Witt model. However, this 
change in conservatism is offset by the use of other input parameter 
changes such as reduced RHR heat exchanger heat removal assumptions 
and increased service water and suppression pool temperature 
assumptions. Additionally, both the SHEX code and the ANS 5.1 decay 
heat model have been previously accepted by NRC as sufficiently 
conservative analysis methods. The spatial evaluation of the 
suppression pool safety relief valve discharge quenchers relative to 
the ECCS pump intake strainers shows steam bubble ingestion is not 
predicted. This supports the elimination of a local suppression pool 
temperature limit.
    Therefore, sufficient margin and adequate NPSH are demonstrated 
with the conservatism of a two sigma (two standard

[[Page 42829]]

deviations) uncertainty in the decay heat model, increased suction 
strainer debris loading, decreased RHR heat exchanger minimum 
performance criteria, and increases in SW [service water] and 
suppression pool temperatures. Thus, the proposed activity does no 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: May 7, 2002.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements for meeting 
surveillances in TS 4.0.a, TS requirements for missed surveillances in 
TS 4.0.c, and TS requirements for a Bases control program consistent 
with TS Bases Control Program described in Section 5.5 of NUREG-1431, 
Standard Technical Specifications for Westinghouse Plants, Revision 2. 
The delay period would be extended from the current limit of `` * * * 
up to 24 hours or up to the limit of the time interval, whichever is 
less'' to `` * * * up to 24 hours or up to the limit of the time 
interval, whichever is greater.'' In addition, the following 
requirement would be added to surveillance requirement 4.0.E: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line-item 
improvement process. 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 May 7, 2002.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

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

    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, 
Unit No. 1 (VCSNS), Fairfield County, South Carolina

    Date of amendment request: May 8, 2002.
    Description of amendment request: The proposed change will exclude 
the control room normal and emergency air handling system from the 
requirement to apply Technical Specification (TS) 3.0.4 to actions 
required by Limiting Condition for Operation 3.7.6 in Modes 5 and 6.
    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:

    South Carolina Electric & Gas Company (SCE&G) has evaluated the 
proposed changes to the VCSNS TS described above against the 
significant Hazards Criteria of 10 CFR 50.92 and has determined that 
the changes do not involve any significant hazard. The following is 
provided in support of this conclusion.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 42830]]

    Response: No.
    The proposed change to Technical Specification 3.7.6 does not 
contribute to the initiation of any accident previously evaluated. 
The actions within the VCSNS TS associated with the control room 
normal and emergency air handling system during shutdown (i.e., 
Modes 5, 6, and defueled) and during the handling of irradiated fuel 
does not require any physical modification to plant components or 
systems. Implementing the proposed action has no impact on the 
probability of an accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to Technical Specification 3.7.6 does not 
contribute to the initiation of any accident previously evaluated. 
The actions within the VCSNS TS associated with the control room 
normal and emergency air handling system during shutdown (i.e., 
Modes 5, 6, and defueled) and during the handling of irradiated fuel 
do not introduce any new accident initiator mechanisms. The 
exclusion of the provisions of Specification 3.0.4 requirements from 
Specification 3.7.6 Mode 5 and 6, action requirements does not cause 
the initiation of any accident nor create any new credible limiting 
single failure nor result in any event previously deemed incredible 
being made credible. As such, it does not create the possibility of 
an accident different than any evaluated in the FSAR [Final Safety 
Analysis Report].
    3. Does this change involve a significant reduction in margin of 
safety?
    Response: No.
    When invoked, the proposed change will allow operational 
transitions involving Modes 5 and 6 within the remedial measures 
currently defined in the specification, including the following when 
one train is inoperable:
     A 7-day AOT [allowed outage time] to restore an 
inoperable train to OPERABLE status.
     Operation of the OPERABLE control room emergency air 
cleanup system in the recirculation mode.
    Although the overall reliability of the system is reduced 
because a single failure in the OPERABLE train could result in a 
loss of function, the 7-day AOT provides adequate margins of safety 
because of the low probability of a design basis accident (DBA) 
occurring during this time period and the ability of the remaining 
train to provide the required capability. Adequate margins of safety 
are also provided by the alternative action that places the unit in 
a protected condition because this ensures the remaining train is 
operating, that no failure preventing automatic actuation will 
occur, and that any active failure can be readily detected.
    With two trains inoperable, action must be taken immediately to 
suspend activities that could result in a release of radioactivity 
that might enter the control room. This places the unit in a 
condition that minimizes accident risk. This does not preclude the 
movement of fuel to a safe position.
    Given the degree of protection provided by the current 
specification, exclusion * * * of the provisions of Specification 
3.0.4 is judged to not result in a significant reduction in the 
margin of safety as described in the bases of any Technical 
Specification.
    Pursuant to 10 CFR 50.91, the preceding analyses provides a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment revises 
the Shutdown Margin limits to Core Operating Limits Report and does not 
change any requirements that are currently in place.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to relocate the Shutdown Margin limits to 
the Core Operating Limits Report [COLR] does not change any 
requirements that are currently in place. No actual plant equipment 
or accident analyses will be affected by the proposed change. The 
Shutdown Margin limits in the COLR will continue to be controlled by 
the STP [South Texas Project] programs and procedures. The safety 
analysis addressed in the UFSAR [updated final safety analysis 
report] will be examined with respect to changes in these limits, 
which are obtained using NRC-[Nuclear Regulatory Commission] 
approved methodologies. Changes to the COLR will be conducted per 
the requirements of 10 CFR 50.59.
    The proposed changes to modify the Specification action 
requirements changing the structure of the specifications to be more 
consistent with NUREG 1431, Westinghouse Improved Standard Technical 
Specifications have no technical impact. The changes clarify time 
requirements and remove details that remain consistent with the 
UFSAR safety analysis. The changes have no effect on the reactivity 
control systems to perform their design functions and involve no 
change to the accident analyses.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes have no influence or impact on, nor do they 
contribute in any way to the probability or consequences of an 
accident. No safety-related equipment or safety function will be 
altered as a result of these proposed changes. The SDM [shutdown 
margin] will continue to be calculated using the NRC-approved 
methods that will be submitted to the NRC. The Technical 
Specifications will continue to require operation within these 
reactivity limits.
    The proposed change modifies the Specification action 
requirements but does not change the way the system is operated. 
When the limiting condition for operation is exceeded, the boration 
control system will continue to be operated in a manner consistent 
with the safety analyses. The details concerning boron flow rate and 
concentration that are removed from the Specifications will be added 
to the TS [technical specification] Bases for the purposes of 
providing an example.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The relocation of the Shutdown Margin limits to the COLR will 
not change any requirements. The values for SDM will remain 
consistent with the UFSAR and will continue to provide their safety 
function through the Shutdown Margin Specification. Actions required 
to be taken to restore SDM will remain in the TS. Therefore, the 
proposed change will not affect the limits on reactivity control, 
and will not permit operations that could result in exceeding these 
limits.
    The proposed change modifies action requirements for restoring 
shutdown margin or refueling boron concentration. The combination of 
parameters currently in the Specification that are being removed 
discuss one means, where as several system lineups and boration 
sources have been evaluated in the safety analysis as acceptable to 
restore Shutdown Margin. Also, the time requirements for the action 
were modified to be consistent with the safety analysis assumptions. 
No actual accident analyses will be affected by these proposed 
changes. The proposed change will not affect reactivity control 
limits and will not permit operations that could result in exceeding 
these limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 42831]]

10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the request for amendments involves no significant 
hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 23, 2002
    Description of amendment request: The proposed amendment revises 
the Unit 2 Operating License and several sections of Technical 
Specifications to delete information differentiating between Unit 1 and 
Unit 2 specific to Model E steam generators.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Operating Licenses currently reflect plant operation with 
both Delta 94 and Model E SGs [steam generators], but all Model E 
SGs will be replaced with Delta 94 SGs by the end of 2002. The 
proposed administrative change deletes information associated with 
the Model E SGs and deletes references to Delta 94 SGs. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The Operating Licenses currently reflect plant operation with 
both Delta 94 and Model E SGs, but all Model E SGs will be replaced 
with Delta 94 SGs by the end of 2002. The proposed administrative 
change deletes information associated with the Model E SGs and 
deletes references to Delta 94 SGs. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Operating Licenses currently reflect plant operation with 
both Delta 94 and Model E SGs, but all Model E SGs will be replaced 
with Delta 94 SGs by the end of 2002. The proposed administrative 
change deletes information associated with the Model E SGs and 
deletes references to Delta 94 SGs. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 23, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications Limiting Conditions for Operation 3.7.1.5, 
Main Steam Isolation Valves, and 3.7.1.7, Main Feedwater Isolation 
Valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the action completion time for one 
MSIV [main steam isolation valve] in Mode 1, and one or more in Mode 
2 and 3, from 4 hours to 8 hours. Extending the completion time is 
not an accident initiator and thus does not change the probability 
that an accident will occur. However, it could potentially affect 
the consequences of an accident if an accident occurred during the 
extended unavailability of the inoperable MSIV. The increase in time 
that the MSIV is unavailable is small and the probability of an 
event occurring during this time period, which would require 
isolation of the main steam flow paths, is low.
    The proposed change extends the action completion time for one 
or more MFIVs [main feedwater isolation valves] from 4 hours to 72 
hours. Extending the completion time is not an accident initiator 
and thus does not change the probability that an accident will 
occur. However, it could potentially affect the consequences of an 
accident if an accident occurred during the extended unavailability 
of the inoperable MFIV. The increase in time that the MFIV is 
unavailable is small and the probability of an event occurring 
during this time period, which would require isolation of the main 
feedwater flow paths, is low.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    Closure of the MSIVs is required to mitigate the consequences of 
large Steam Line Break inside containment. The proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Closure of the MFIVs is required to mitigate the consequences of 
the Main Steam Line Break and Main Feedwater Line Break accidents. 
The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed changes do not change any Technical Specification 
Limit or accident analysis assumption. Therefore it does not involve 
a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis, & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.
    Virginia Electric and PowerCompany, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
    Date of amendment request: May 14, 2002.
    Description of amendment request: The proposed changes would revise 
the Technical Specifications and associated Bases to revise the 
surveillance frequency of the containment spray and recirculation spray 
system spray header nozzles from a periodic surveillance to a 
performance-based surveillance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed revision to Technical Specifications changes the 
frequencies of the surveillance requirements for the Containment 
Spray and Recirculation Spray nozzles. The frequency is being 
changed from every 10 years to ``following maintenance which could 
result in nozzle blockage.'' In accordance with the requirements of 
10 CFR 50.92, the enclosed application is judged to

[[Page 42832]]

involve no significant hazards based upon the following information:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change revises the surveillance frequencies from 
every 10 years to ``following maintenance which could result in 
nozzle blockage.'' Analyzed events are initiated by the failure of 
plant structures, systems, or components. The Containment Spray and 
Recirculation Spray Systems are not considered to be initiators of 
any analyzed event. The proposed change does not have a detrimental 
impact on the integrity of any plant structure, system, or component 
that initiates an analyzed event. The proposed change will not alter 
the operation of or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident. As a 
result, the probability of any accident previously evaluated is not 
significantly increased.
    The proposed change revises the surveillance frequencies. 
Reduced testing is justified where operating experience has shown 
that routinely passing a surveillance test performed at a specified 
interval has no apparent connection to overall component 
reliability. In this case, routine surveillance testing at the 
specified frequency is not connected to any activity, which may 
initiate reduced component reliability, and therefore has been of 
limited value in ensuring component reliability. Thus, the proposed 
frequency change is not significant from a reliability standpoint. 
The proposed containment spray and recirculation spray nozzle 
surveillance frequencies have been established based on achieving 
acceptable levels of equipment reliability.
    This change does not affect the plant design. Due to the plant 
design, the spray ring headers are maintained dry. Formation of 
significant corrosion products is unlikely. Due to their location at 
the top of the containment, introduction of foreign material from 
exterior to the headers is unlikely. Since maintenance that could 
introduce foreign material is the most likely cause for obstruction, 
testing or inspection following such maintenance would verify the 
nozzle(s) remain unobstructed and the systems' continued capability 
to perform their safety function(s). As a result, the consequences 
of any accident previously evaluated are not significantly affected 
by the proposed change in surveillance frequencies.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The margin of safety for this system is based on the capacity of 
the spray headers. The system is not susceptible to corrosion 
induced obstruction or obstruction from external sources to the 
system. Performance of maintenance on a spray ring header would now 
require evaluation of the potential for nozzle blockage and the need 
for a test or inspection. Consequently, the spray header nozzles 
should remain unblocked and available in the event that the safety 
function is required. Hence, the change in surveillance frequencies 
does not involve a significant reduction in the margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: July 9, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications to be consistent with changes made to 10 CFR 50.59, 
``Changes, tests, and experiments.''
    Date of issuance: June 4, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 151.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44162). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 11, 2001, as 
supplemented on April 8, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications, deleting the cycle-specific footnote regarding the 
safety limit minimum critical power ratio in Section 2.1.A, and making 
associated administrative changes.
    Date of Issuance: May 31, 2002.
    Effective date: May 31, 2002, and shall be implemented within 30 
days of issuance.
    Amendment No.: 228.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.

[[Page 42833]]

    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59501). The April 8, 2002, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated May 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 4, 2001, as supplemented 
July 20, 2001.
    Brief Description of amendments: The proposed license amendments 
change the Technical Specifications Surveillance Frequency and Action 
Requirements for the suppression chamber-to-drywell vacuum breakers at 
the Brunswick Steam Electric Plant, Units 1 and 2.
    Date of issuance: June 3, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days from date of issuance.
    Amendment Nos.: 223 and 248.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34280). The July 20, 2001, supplement contained clarifying information 
only, and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 3, 2002.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: August 27, 2001.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) 3/4.6.1.3, ``Containment Systems--Containment Air 
Locks'' and the associated TS Bases section.
    Date of issuance: June 7, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 267.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55010). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 6, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specification 3.3.1 allowable values for the reactor trip 
system instrumentation overtemperature delta temperature and overpower 
delta temperature set points.
    Date of issuance: May 23, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 202 and 183.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2920). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 23, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 8, 2002, as supplemented 
on April 15, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.1, ``Main Steam Safety Valves,'' to reduce the 
maximum allowable power range neutron flux high setpoint when one or 
more main steam line safety valves are inoperable. The amendment also 
revises the associated TS Basis to incorporate a more conservative 
equation to calculate this setpoint.
    Date of issuance: June 4, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 228.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10012). The April 15, 2002, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 4, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 8, 2002.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.8, ``Refueling, Fuel Storage and Operations with 
the Reactor Vessel Head Bolts Less Than Fully Tensioned,'' and TS 
4.5.F, ``Fuel Storage Building Air Filtration System,'' by deleting the 
requirements for the Fuel Storage Building Air Filtration System. The 
amendment also revised the associated Basis sections.
    Date of issuance: June 5, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 229.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10013). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 5, 2002.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 13, 2002, as supplemented by 
letter dated May 23, 2002.
    Brief description of amendment: The amendment corrects several 
errors that were found subsequent to Nuclear Regulatory Commission 
issuance of Amendment No. 215, which converted the Plant Technical 
Specifications (TSs) for Arkansas Nuclear One, Unit 1 to Improved TSs.
    Date of issuance: June 10, 2002.
    Effective date: As of the date of issuance and shall be implemented 
in conjunction with the implementation of Amendment No. 215.
    Amendment No.: 218.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications/license.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21287).

[[Page 42834]]

The supplemental letter dated May 23, 2002, provided additional 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 2, 2001, as supplemented by 
letter dated March 20, 2002.
    Brief description of amendment: The amendment relocated the 
requirements for the containment recirculation system from the 
Technical Specifications to the Technical Requirements Manual.
    Date of issuance: May 31, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 245.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29352). The March 20, 2002, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear 
Power Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: September 5, 2001.
    Brief description of amendment: The amendment revises the battery 
terminal voltage on float charge for the alternate battery.
    Date of issuance: June 6, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-19: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2002 (67 FR 
10013). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 6, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 9, 2001.
    Brief description of amendments: Replace the phrase ``involves an 
unreviewed safety question as defined in'' with ``requires NRC approval 
pursuant to,'' maintaining reference to 10 CFR 50.59, ``Changes, tests, 
and experiments,'' in order to provide consistency with changes to 10 
CFR 50.59 as published in the Federal Register (64 FR 53582) dated 
October 4, 1999.
    Date of issuance: June 4, 2002.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment Nos.: 182 and 169.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44170). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 4, 2002.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: February 20, 2002.
    Brief description of amendments: These amendments revised TS 3/
4.6.5, ``Vacuum Relief Valves,'' to make the Limiting Condition for 
Operation applicable to vacuum relief ``lines'' and extend the allowed 
outage time for the containment vacuum relief lines from 4 hours to 72 
hours. Also, some specific requirements for surveillance testing and 
valve actuation setpoints are relocated to the TS Bases documents.
    Date of Issuance: May 30, 2002.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 182 and 125.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12602). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 12, 2002
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 4.0.E to extend the delay period, prior to having to 
declare the subject equipment inoperable, 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 time interval, whichever 
is less'' to `` * * * up to 24 hours or up to the limit of the time 
interval, whichever is greater.'' In addition, the following 
requirement is added to SR 4.0.E: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: May 31, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 127.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2002 (67 FR 
15625). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 27, 2002, as supplemented by 
letter dated May 9, 2002.
    Brief description of amendment: The amendment revises the maximum 
allowable value of the reactor protective system (RPS) variable high 
power trip (VHPT) setpoint from 107.0% to 109.0%. Specifically, TS 
Table 1-1, ``RPS Limiting Safety System Settings,'' in the Trip 
Setpoints column for Trip Number 1 [High Power Level (A) 4-Pump 
Operation] has been revised from 107.0% to 109.0%. In addition, TS 
Section 1.3(1), ``Basis,'' describing the high power trip initiation, 
has been revised from 107.0% to 109.0%.
    Date of issuance: May 29, 2002.
    Effective date: May 29, 2002, to be implemented within 30 days from 
the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards

[[Page 42835]]

consideration: Yes (67 FR 34478 dated May 14, 2002). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by June 13, 2002, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, consultation with the State of Nebraska and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated May 29, 2002.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: March 18, 2002.
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period is extended from the current limit of ``* * * up to 24 
hours or up to the limit of the specified Frequency, whichever is 
less'' to ``* * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is greater.'' In addition, the following 
requirement is added to SR 3.0.3: ``A risk evaluation shall be 
performed for any Surveillance delayed greater than 24 hours and the 
risk impact shall be managed.''
    Date of issuance: June 12, 2002.
    Effective date: June 12, 2002.
    Amendment No.: 82.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21293). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 12, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: January 24, 2002.
    Brief Description of amendments: The amendments delete Technical 
Specifications Section 5.5.3, ``Post Accident Sampling,'' for Farley 
Nuclear Plant, Units 1 and 2, and thereby eliminated the requirements 
to have and maintain the post-accident sampling systems.
    Date of issuance: May 22, 2002.
    Effective date: As of the date of issuance and shall be implemented 
by December 31, 2002.
    Amendment Nos.: 156 and 148.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2002 (67 FR 
21293). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 22, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 8, 2001, as amended by 
your letter dated April 8, 2002.
    Brief description of amendments: The amendments deleted various 
reporting requirements from the Sequoyah Technical Specifications (TSs) 
because they are duplicative to the requirements of 10 CFR 50.72 and 10 
CFR 50.73. One exception was reporting of steam generator tube 
inspection results, TS 4.4.5.5.c, which is more stringent than 10 CFR 
50.72 and 10 CFR 50.73. Therefore, the request to delete this TS was 
denied.
    Date of issuance: May 24, 2002.
    Effective date: Date of issuance, to be implemented within 45 days 
of issuance.
    Amendment Nos.: 276 and 267.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5339). The supplemental letter provided clarifying information that was 
within the scope of the initial notice and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 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: August 24, 2001, as supplemented by 
letter dated April 15, 2002.
    Brief description of amendments: The amendments extend the 
surveillance test interval from ``92 days'' to ``18 months'' for 
Westinghouse Electric Company Type AR relays with alternating current 
coils used as Solid State Protection System slave relays, in 
Surveillance Requirement (SR) 3.3.2.6 and auxiliary (i.e., interposing) 
relays in the containment ventilation isolation system in SR 3.3.6.5.
    Date of issuance: May 31, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 96 and 96.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52804). The April 15, 2002, supplement provided clarifying information 
and did not change the original no significant hazards determination 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 31, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 31, 2001, as supplemented 
by letters dated October 17, 2001, and March 5, 2002.
    Brief Description of amendments: These amendments revise the 
Technical Specifications to add a 14-day allowed outage time for the 
power-operated relief valve backup air supply, and additional 
surveillance, functional testing, and calibration requirements.
    Date of issuance: May 31, 2002.
    Effective date: May 31, 2002.
    Amendment Nos.: 231 and 231.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64310). The supplements dated October 17, 2001, and March 5, 2002, 
provided clarifying information that did not change the scope of the 
May 31, 2001, application nor the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 31, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of June 2002.


[[Page 42836]]


    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-15683 Filed 6-24-02; 8:45 am]
BILLING CODE 7590-01-P