[Federal Register Volume 66, Number 153 (Wednesday, August 8, 2001)]
[Notices]
[Pages 41609-41631]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-19746]


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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 July 16, 2001 through July 27, 2001. The 
last biweekly notice was published on July 25, 2001 (66 FR 38756).

[[Page 41610]]

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 NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By September 7, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). 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.
    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,

[[Page 41611]]

Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Branch, or may be delivered to the Commission's Public Document Room, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852, by the above date. A copy of the petition 
should also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and 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 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 Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-
415-4737 or by email to [email protected].

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

    Date of amendment request: May 24, 2001.
    Description of amendment request: Generic Letter 96-04 informed all 
licensees of the issues concerning the use of Boraflex in spent fuel 
storage racks. In an October 15, 1996, response to the generic letter, 
the licensee stated that a reevaluation of the criticality analysis for 
the Oyster Creek fuel racks would be performed to consider Boraflex 
degradation including boron carbide loss. A reevaluation of the Oyster 
Creek criticality analysis including consideration of Boraflex 
degradation has been performed and the licensee is asking for review 
and approval of the proposed change to its licensing basis of spent 
fuel racks containing Boraflex.
    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 amendment does not:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The accident of concern is a fuel bundle drop onto the top of a 
storage rack as described in DPR-16 License Amendment No. 76 dated 
September 17, 1984 and DPR-16 License Amendment No. 121. This 
accident was previously considered in an analysis that calculated 
the reactivity of two unpoisoned fuel assemblies separated only by 
water. The analysis shows a separation of 2.5 inches results in a 
reactivity k of 0.90. For a fuel assembly lying 
horizontally on the top of a rack, the separation distance would be 
 14 inches. Since only water separation is considered and 
no credit is taken for Boraflex, there is no effect on this accident 
as described in the SAR [Safety Analysis Report].
    The SAR identifies that keff for the spent fuel shall 
not exceed 0.95 accounting for uncertainties. This criticality 
analysis, which includes consideration of Boraflex degradation, 
shows the spent fuel pool Keff will remain below 0.95 
with a 95% probability at the 95% confidence level. Therefore, the 
revised criticality analysis for Boraflex degradation does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The change does not involve any plant systems associated with 
plant operation so safe plant operation will not be affected. This 
analysis does include a new consideration (dissolution of the 
Boraflex in the fuel racks) that had not been previously considered.
    Nuclear safety is not effected [affected] since the required 
margin to criticality is maintained with consideration of Boraflex 
degradation. The current analysis uses the conservative assumption 
of coplanar gaps (i.e., all gaps occurring at the same axial plane). 
This is a very conservative assumption given gap measurement data at 
Oyster Creek and in the industry that shows an axial distribution of 
gaps.
    The proposed criticality analysis utilizes an axial distribution 
of gaps. The analysis is based on the same fuel design and 
enrichment as the previous analysis, a GE7 8x8 fuel design having 
4.0% enrichment and seven rods containing 3.0% 
Gd3O8 depleted to peak reactivity. The 
analysis assumes shrinkage up to 4.2% of panel length, gaps of 5.89 
inches occurring in 75% of the panels, and 10% thinning (4 mils) of 
the panel thickness. The analysis conforms to regulatory and 
industry guidelines for criticality analyses and the calculated 
keff provides 95% probability at the 95% confidence 
level. The design limit is 0.9410 (5.0% design margin plus 
calculational uncertainity) and the spent fuel pool keff 
is 0.9381 including manufacturing uncertainties. This establishes 
the acceptability of the assumed Boraflex degradation against design 
limits.
    The analysis, which includes the effect of Boraflex degradation, 
demonstrates that keff in the fuel racks remains below 
the license requirement of 0.95. The possibility of a new or 
different kind of accident from any accident previously evaluated is 
not created since keff remains below 0.95 when Boraflex 
degradation mechanisms are considered and the change does not 
involve any plant systems or procedures associated with plant 
operation.
    (3) Involve a significant reduction in a margin of safety.
    As stated in Oyster Creek Technical Specification Section 5.3.1, 
the fuel pool keff is limited to 0.95 to assure [ensure] 
an ample margin to criticality. The new analysis demonstrates this 
margin is maintained given the Boraflex degradation assumed in the 
analysis that is based on industry and Oyster Creek specific 
observations and testing. The new analysis revises the Boraflex gap 
assumption to use a random axial distribution of gaps rather than a 
more conservative coplanar (gaps in same location in all fuel 
bundles) distribution. The axial distribution is more representative 
of actual gap locations observed at Oyster Creek (based on Blackness 
and BADGER testing) and other plants with similar rack designs. The 
assumption remains conservative since all Boraflex gaps are assumed 
to occur in the upper three-quarters of the rack height. This 
results in an over estimation of gaps in a smaller area that 
increases the reactivity penalty. Since the required keff 
limit of 0.95 is not exceeded, and the analysis remains 
conservative, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard Correia, Acting.

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: June 15, 2001.
    Description of amendments request: The proposed amendments delete 
requirements from the Technical Specifications (TSs) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were

[[Page 41612]]

generally required to implement PASS upgrades as described in NUREG-
0737, ``Clarification of TMI [Three Mile Island] Action Plan 
Requirements,'' and Regulatory Guide 1.97, ``Instrumentation for Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs 
Conditions During and Following an Accident.'' Implementation of these 
upgrades was an outcome of the lessons learned from the accident that 
occurred at TMI, Unit 2. Requirements related to PASS were imposed by 
Order for many facilities and were added to or included in the TSs for 
nuclear power reactors currently licensed to operate. Lessons learned 
and improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, 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 October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated June 15, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The NRC staff proposes to determine that the amendment request 
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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: July 17, 2001.
    Description of amendment request: By letters dated October 4, 2000, 
and December 14, 2000, Carolina Power & Light Company (CP&L) submitted 
license amendment requests to revise the Harris Nuclear Plant (HNP) 
Operating License and Technical Specifications (TS) to support steam 
generator replacement (SGR) and to allow operation at an uprated 
reactor core power level of 2900 megawatts thermal (Mwt). CP&L, in its 
letter of July 17, 2001, proposed to revise the Final Safety Analysis 
Report (FSAR) Chapter 15 accident analyses, which had been previously 
submitted as part of the October 4, 2000, SGR amendment request. The 
proposed revision to the accident analyses would adopt the alternate 
source term (AST) methodology, using the guidance of Nuclear Regulatory 
Commission (NRC) Regulatory Guide 1.183.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below. The licensee's analysis is 
limited to its request to use the AST methodology for the accident 
analyses. The NRC staff's proposed no significant hazards consideration 
determination for the SGR amendment request, published in the Federal 
Register on November 1,

[[Page 41613]]

2000 (65 CFR 65338), and the Notice of Consideration of Issuance to 
Amendment to Facility Operating License and Opportunity for Hearing for 
the power uprate, published on February 6, 2001 (66 CFR 9110), remain 
valid for the other aspects of the SGR and power uprate amendment 
requests.

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    An alternative source term calculation has been performed for 
HNP which demonstrates that dose consequences remain below limits 
specified in NRC Regulatory Guide 1.183 and 10 CFR 50.67. The 
proposed change does not modify the design or operation of the 
plant.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect plant structures, systems, 
or components. The operation of plant systems and equipment will not 
be affected by this proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed change is the implementation of the alternate 
source term methodology consistent with NRC Regulatory Guide 1.183. 
The proposed change does not significantly affect any of the 
parameters that relate to the margin of safety as described in the 
Bases of the TS or FSAR. Accordingly, NRC Acceptance Limits are not 
significantly affected by this change.
    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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Section Chief: Patrick M. Madden, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 16, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 3.1.A, ``Reactor Coolant 
System Operational Components,'' 3.1.B, ``Reactor Coolant System [RCS] 
Heatup and Cooldown,'' 3.2, ``Chemical and Volume Control System,'' 
3.3.A, ``Engineered Safety Feature Safety Injection and Residual Heat 
Removal Systems,'' and 4.3, ``Reactor Coolant System Integrity 
Testing,'' to incorporate revised reactor pressure vessel pressure-
temperature limits to allow operation up to 25 effective full-power 
years (EFPY). The proposed amendment would also make changes to the 
associated TS Bases sections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no physical changes to the plant being introduced by the 
proposed changes to the heatup and cooldown limitation curves. The 
proposed changes do not modify the RCS pressure boundary. That is, 
there are no changes in operating pressure, materials, or seismic 
loading. The proposed changes do not adversely affect the integrity 
of the RCS pressure boundary such that its function in the control 
of radiological consequences is affected. The proposed heatup and 
cooldown limitation curves were generated in accordance with the 
fracture toughness requirements of 10CFR50 Appendix G, and ASME B&PV 
Code [American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code], Section XI, Appendix G in conjunction with ASME Code 
Cases N-640 and N-588. The proposed heatup and cooldown limitation 
curves were established in compliance with the methodology used to 
calculate and predict effects of radiation on embrittlement of RPV 
[reactor pressure vessel] beltline materials. Use of this 
methodology provides compliance with the intent of 10CFR50 Appendix 
G and provides margins of safety that ensure non-ductile failure of 
the RPV will not occur.
    The proposed heatup and cooldown limitation curves prohibit 
operation in regions where it is possible for non-ductile failure of 
carbon and low alloy RCS materials to occur. Hence, the primary 
coolant pressure boundary integrity will be maintained throughout 
the limit of applicability of the curves, 25 EFPY. Operation within 
the proposed OPS [overpressure protection system] limits ensures 
that overpressurization of the RCS at low temperatures will not 
result in component stresses in excess of those allowed by the ASME 
B&PV Code Section XI Appendix G.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or the consequences of an accident 
previously evaluated.
    2. 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.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes to the heatup and cooldown 
limitation curves were generated in accordance with the fracture 
toughness requirements of 10CFR50 Appendix G and ASME B&PV Code, 
Section XI, Appendix G in conjunction with ASME Code Cases N-588 and 
N-640. Compliance with the heatup and cooldown limitation curves 
will ensure that conditions in which non-ductile failure of the RCS 
pressure boundary materials is possible will be avoided. Compliance 
with the proposed OPS limits will ensure that the RCS will be 
physically protected against overpressurization events during low 
temperature operation when the fracture toughness properties of the 
carbon and low alloy components are at their lowest.
    No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Further, the proposed 
changes to the heatup and cooldown limitation curves and the OPS 
limits do not affect any activities or equipment other than the RCS 
pressure boundary and are not assumed in any analysis to initiate or 
mitigate any accident sequence. 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 
amendment would not involve a significant reduction in the margin of 
safety.
    The proposed TS changes do not involve a significant reduction 
in the margin of safety.
    The revised heatup and cooldown limitation curves and OPS limits 
provide more operating flexibility than the current heatup and 
cooldown limitation curves. Industry experience since the inception 
of pressure-temperature limits in the 1970s confirms that some of 
the original methodologies used to develop the heatup and cooldown 
limitation curves are overly conservative. Accordingly, ASME Code 
Cases N-588 and N-640 take advantage of the acquired knowledge by 
establishing more realistic methodologies for development of the 
heatup and cooldown limitation curves. Therefore, operational 
flexibility is gained and an acceptable margin of safety to reactor 
pressure vessel non-ductile type fracture is maintained.
    The revised heatup and cooldown limitation curves and OPS limits 
are established in accordance with current regulations and the ASME 
B&PV Code 1996 version. These proposed changes are acceptable 
because the ASME B&PV Code maintains the margin of safety required 
by

[[Page 41614]]

10CFR50.55(a). Because operation will be within these limits, the 
RCS materials will continue to behave in a ductile manner consistent 
with the original design bases.
    The proposed changes to the allowable operation of charging and 
safety injection pumps when OPS is required to be operable is 
consistent with the IP2 [Indian Point 2] licensing bases but 
implements the licensing bases in a more conservative manner than 
the current TS. The change in OPS surveillance frequency has been 
previously evaluated by the NRC to involve an insignificant increase 
in risk. That insignificant increase in risk is offset by the 
adverse effects of the alternatives of either 1.) delaying forced 
cooldowns until OPS testing is complete; 2.) complicating cooldown 
operations by imposition of limits required when OPS is inoperable; 
or 3.) conducting OPS testing periodically while at power.
    Therefore, Con Edison has concluded that the proposed changes do 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia (Acting).

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

    Date of amendment request: May 31, 2001.
    Description of amendment request: The proposed amendment would 
increase the allowed outage time (AOT) for one inoperable emergency 
diesel generator (EDG) from 72 hours to 14 days to allow the 
performance of various maintenance and repair activities while the 
plant is operating.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification change to increase the EDG 
AOT from 72 hours to 14 days will not cause an accident to occur and 
will not result in any change in the operation of the associated 
accident mitigation equipment. The EDGs are not accident initiators, 
and extending the EDG AOT will not impact the frequency of any 
previously evaluated accidents. The design basis accidents will 
remain the same postulated events described in the Millstone Unit 
No. 2 Final Safety Analysis Report (FSAR). In addition, extending 
the EDG AOT will not impact the consequences of an accident 
previously evaluated. The consequences of previously evaluated 
accidents will remain the same during the proposed 14 day AOT as 
during the current 72 hour AOT. The ability of the remaining EDG to 
mitigate the consequences of an accident will not be affected since 
no additional failures are postulated while equipment is inoperable 
within the Technical Specification AOT. The remaining EDG is 
sufficient to mitigate the consequences of any design basis 
accident. Therefore, the proposed change will not increase the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification change to allow 
verification of offsite circuit(s) within 1 hour prior to or after 
entering the condition of either an inoperable offsite source or 
inoperable EDG will not cause an accident to occur and will not 
result in any change in the operation of the associated accident 
mitigation equipment. Performing a verification of the offsite 
circuits does not require any equipment manipulations or operator 
actions that could cause a previously evaluated accident to occur. 
Providing the flexibility to verify offsite circuit availability 
before removing equipment from service will reduce the potential to 
establish an adverse plant configuration. The design basis accidents 
will remain the same postulated events described in the Millstone 
Unit No. 2 FSAR. In addition, allowing an early verification of 
offsite circuit(s) will not impact the consequences of an accident 
previously evaluated. The consequences of previously evaluated 
accidents will remain the same whether the verification is performed 
immediately after, or just before, an EDG or offsite circuit is 
removed from service. The ability of the remaining power sources to 
mitigate the consequences of an accident will not be affected since 
no additional failures are postulated while equipment is inoperable 
within the Technical Specification AOT. The remaining power sources 
are sufficient to mitigate the consequences of any design basis 
accident. Therefore, the proposed change will not increase the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification changes associated with the 
requirements for the pressurizer heaters to be supplied by emergency 
power will not result in any change in plant design. These 
components will continue to be powered from Class 1E power sources. 
As a result, the operation and reliability of the pressurizer 
heaters will not be affected by the proposed changes. In addition, 
operation of the pressurizer heaters is not assumed to mitigate any 
design basis accident. The proposed changes will not cause an 
accident to occur and will not result in a change in the operation 
of any accident mitigation equipment. The design basis accidents 
remain the same postulated events described in the Millstone Unit 
No. 2 FSAR. Therefore, the proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    The additional change to add the requirement to verify that the 
steam driven auxiliary feedwater (SDAFW) pump is operable when one 
EDG is inoperable will ensure sufficient auxiliary feedwater 
capability is available if a loss of offsite power were to occur. 
Operation of the SDAFW pump will not be affected by the proposed 
change, and the SDAFW pump is not an accident initiator. Verifying 
operability of the SDAFW pump will not impact the frequency of any 
previously evaluated accidents. The design basis accidents will 
remain the same postulated events described in the Millstone Unit 
No. 2 FSAR. The ability of the SDAFW pump to mitigate the 
consequences of an accident will not be affected. Therefore, the 
proposed change will not increase the probability or consequences of 
an accident previously evaluated.
    The additional proposed changes to renumber action requirements 
and remove a footnote that is no longer valid will not result in any 
technical changes to the current requirements. Therefore, these 
additional proposed change[s] will not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the Technical Specifications do not 
impact any system or component in a manner that could cause an 
accident. The proposed changes will not alter the plant 
configuration (no new or different type of equipment will be 
installed) or require any unusual operator actions. The proposed 
changes will not alter the way any structure, system, or component 
functions, and will not significantly alter the manner in which the 
plant is operated. There will be no adverse effect on plant 
operation or accident mitigation equipment. The response of the 
plant and the operators following an accident will not be 
significantly different. In addition, the proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification change to increase the EDG AOT 
from 72 hours to 14 days and allow verification of offsite circuit(s) 
within 1 hour prior to or after entering the condition of an inoperable 
offsite source or inoperable EDG does not adversely affect equipment 
design or operation, and there are no changes being made to the 
Technical Specification required safety limits or safety system 
settings that would adversely affect plant safety. The proposed 
Technical Specification change, in conjunction with the administrative 
controls, provides adequate assurance of the capability to supply power 
to the safety related Class 1E electrical loads thereby ensuring the 
accident mitigation functions will be

[[Page 41615]]

maintained. The availability of offsite power combined with the 
availability of the Millstone Unit No. 3 Station Blackout diesel 
generator and the use of the Configuration Risk Management Program 
required by 10 CFR 50.65(a)(4) provide adequate compensation for the 
small incremental increase in plant risk of the proposed EDG AOT 
extension. This small increase in plant risk while operating is offset 
by a reduction in shutdown risk resulting from the increased 
availability and reliability of the EDGs during refueling outages, and 
avoiding transition risk incurred during unplanned plant shutdowns. In 
addition, the calculated risk measures associated with the proposed AOT 
are below the acceptance criteria defined in Regulatory Guide 1.177, An 
Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' dated August 1998. Therefore, the proposed change 
will not result in a significant reduction in a margin of safety.
    The proposed Technical Specification changes associated with the 
requirements for the pressurizer heaters to be supplied by emergency 
power do not adversely affect equipment design or operation, and there 
are no changes being made to the Technical Specification required 
safety limits or safety system settings that would adversely affect 
plant safety. The emergency power requirement for the pressurizer 
heaters, which came from the Three Mile Island (TMI) action item 
requirement item ll.E.3.1, Emergency Power Requirements for Pressurizer 
Heaters,'' of NUREG-0737, ``A Clarification of TMI Action Plan 
Requirements,'' will continue to be met. The pressurizer heaters are 
permanently connected to Class 1E power supplies as described in the 
Millstone Unit No. 2 FSAR. Therefore, these changes will not result in 
a significant reduction in a margin of safety.
    The additional more restrictive change to add the requirement to 
verify that the SDAFW pump is operable when one EDG is inoperable will 
not adversely affect equipment design or operation, and there are no 
changes being made to the Technical Specification required safety 
limits or safety system settings that would adversely affect plant 
safety. Operation of the SDAFW pump will not be affected by the 
proposed change. Therefore, this change will not result in a 
significant reduction in a margin of safety.
    The additional proposed changes to renumber action requirements and 
remove a footnote that is no longer valid will not result in any 
technical changes to the current requirements. Therefore, these 
additional changes will not result in 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385 Hartford, Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: July 2, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, 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 October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 2, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).

[[Page 41616]]

    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 radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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: July 2, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, 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 October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 2, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

[[Page 41617]]

procedures, and programs that provide effective mitigation of and 
recovery from reactor accidents, results in a neutral impact to the 
margin of safety. Methodologies that are not reliant on PASS are 
designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. 
The intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    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: Richard L. Emch, Jr.

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

    Date of amendment request: July 9, 2001.
    Description of amendment request: This Technical Specification (TS) 
change removes TS requirements that will no longer be applicable 
following replacement of the part-length control element assemblies 
(PLCEAs) with five-element full-length control element assemblies 
(CEAs) and removal of the four-element CEAs on the core periphery.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed changes are administrative in nature and maintain 
the conservative restrictions on the operation of the CEAs. Chapter 
15 of the Waterford 3 [Waterford Steam Electric Station, Unit 3] 
Safety Analysis Report identifies the analyses associated with the 
CEAs. These analyses are evaluated in the development of the Reload 
Analysis for each fuel cycle, and the appropriate limitations to 
insure acceptable analysis results are incorporated in the Core 
Operating Limits Report (COLR) for the fuel cycle. The modifications 
replacing the part-length CEAs with full-length CEAs and removing 
the four-element CEAs will be evaluated under the 10 CFR 50.59 
process prior to implementation. The Reload Analysis and changes to 
the COLR are also evaluated under the 10 CFR 50.59 process prior to 
incorporating the identified changes.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed changes introduce no new mode of plant operation 
and are considered to be administrative in nature. Operating 
experience has shown that the full-length CEAs are capable of 
controlling the axial power distribution function intended for the 
part-length CEAs. The part-length CEAs will be replaced with the 
same type of full-length CEAs used in the shutdown and regulating 
CEA groups. Removal of the four-element CEAs provides no mechanism 
for creating a new or different kind of accident.
    Therefore, this 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 this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed changes may improve overall safety margins. 
Replacements of the part-length CEAs with full-length CEAs will 
allow Entergy [Operations, Inc.,] to credit these CEAs in the 
shutdown margins calculations. The worth of the four-element CEAs 
being removed from the core is relatively small in the modern low-
leakage core design used at Waterford 3. Therefore, the removal of 
the four-element CEAs in combination with the replacement of the 
part-length CEAs with full-length CEAs will result in an overall 
increase in the net CEA worth. This will result in an increase in 
the available shutdown margin during reactor operation.
    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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: June 26, 2001.
    Description of amendment request: The proposed amendments would 
extend the dates specified in Operating License Sections 2.C(8) and 
3.P, ``Pressure-Temperature Limit Curves,'' for Dresden Nuclear Power 
Station, Units 2 and 3, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to the operating license to extend the 
limitations on the use of the [Pressure-Temperature] P-T limits does 
not affect the operations or configuration of any plant equipment. 
Thus, no new accident initiators are created by this change.
    The proposed change extends the use of the pressure-temperature 
(P-T) limits for an additional cycle of operation on each unit. The 
P-T limits are based on the projected reactor pressure vessel (RPV) 
neutron fluence at 32 effective full power years (EFPYs) of 
operation. At the end of the next cycle of operation, the Dresden 
Nuclear Power Station (DNPS) units will have attained a maximum of 
67.5% of the 32 EFPY operating times. Separately, we submitted a 
license amendment request to permit operation with an extended power 
uprate (EPU). Even with an approximately 17% increase in reactor 
power for one cycle due to the EPU, this provides significant margin 
to ensure that the current 32 EFPY fluence projection of 5.1  x  
1017 n/cm2 will not be exceeded. This ensures 
that the basis for proposed applicability of the P-T limits is 
conservative and that the RPV integrity is protected under all 
operating conditions. Therefore, neither the probability nor the 
consequences of an accident are increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to the operating license to extend the 
limitations on the use of the P-T limits does not affect the 
operation or configuration of any plant equipment. The current P-T 
limits will remain valid and conservative during the proposed 
extension. Thus, no new or different accidents are created by this 
proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different

[[Page 41618]]

kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change extends the use of the P-T limits for an 
additional cycle of operation on each unit. The P-T limits are based 
on the projected RPV neutron fluence at 32 EFPYs of operation. At 
the end of the next cycle of operation, the DNPS units will have 
attained a maximum of 67.5% of the 32 EFPY operating times. In a 
separate license amendment request, ComEd submitted a license 
amendment request to permit operation with an extended power uprate 
(EPU). Even with an approximately 17% increase in reactor power for 
one cycle due to the EPU, this provides sufficient margin to ensure 
that the current 32 EFPY fluence projection of 5.1  x  
1017 n/cm2 will not be exceeded. This ensures 
that the basis for the P-T limits is conservative and therefore 
ensures that the reactor pressure vessel integrity is protected 
under all operating conditions. 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 15, 2001 (RS-01-117).
    Description of amendment request: The proposed amendments would 
allow the use of ATRIUM 10 fuel from Framatome Advanced Nuclear Fuel, 
Inc.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to LaSalle County Station, Unit 1 and Unit 
2 Technical specification (TS), add the fuel analytical methods to 
TS Section 5.6.5, ``Core Operating Limits Report (COLR),'' that 
support insertion of Framatome Advanced Nuclear Fuel, Inc. (i.e., 
Framatome) ATRIUM 10 fuel.
    LaSalle County Station Unit 1, is scheduled to load ATRIUM 10 
fuel during its upcoming outage in November 2001. The proposed 
changes to TS Section 5.6.5 will add the fuel analytical methods 
that support the initial insertion of ATRIUM 10 fuel tot he list of 
methods used to determine the core operating limits. The addition of 
approved methods to TS Section 5.6.5 has no effect on any accident 
initiator or precursor previously evaluated and does not change the 
manner in which the core is operated. The NRC approved methods have 
been reviewed to ensure that the output accurately models predicted 
core behavior, have no affect on the type or amount of radiation 
released, and have no affect on predicted offsite doses in the event 
of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to TS Section 5.6.5 do not affect the 
performance of any LaSalle County Station structure, system, or 
component credited with mitigating any accident previously 
evaluated. The insertion of a new generation of fuel which has been 
analyzed with NRC approved methodologies will not affect the control 
parameters governing unit operation or the response plant equipment 
to transient conditions. The proposed changes do not introduce any 
new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to TS Section 5.6.5 will add the ATRIUM 10 
fuel analytical methods to the list of methods used to determine the 
core operating limits. The additional methods have been previously 
approved by the NRC for use by licensees. The proposed changes do 
modify the safety limits or setpoints at which protective actions 
are initiated, and do not change the requirements governing 
operation or availability of safety equipment assumed to operate to 
preserve the margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Robert Helfrich, Senior Counsel, 
Nuclear, Mid-West regional Operating Group, Exelon Generation Company, 
LLC, 1400 Opus Place, Suite 900, Downers Grove, IL 60515.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 15, 2001 (RS-01-118).
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) Section 3.5.1, ``ECCS-Operating,'' 
Surveillance Requirement (SR) 3.5.1.8. The proposed changes will 
eliminate the requirement that the Automatic Depressurization System 
(ADS) designated Safety/Relief Valves (S/Vs) open during the manual 
actuation of the ADS and changes the SR frequency to require the 
testing of all required ADS manual actuation solenoids during the 
performance of SR 3.5.1.8 in lieu of on a staggered basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes modify Technical Specifications (TS) 
Section 3.5.1, ``ECCS--Operating,'' Surveillance Requirement (SR) 
3.5.1.8. The proposed changes will eliminate the TS requirement that 
the Automatic Depressurization System (ADS) designated Safety/Relief 
Valves (S/RVs) open during the manual actuation of the ADS and 
rewords the SR frequency to require the testing of all required ADS 
manual actuation solenoids during the performance of SR 3.5.1.8 in 
place of testing on a staggered basis. The performance of ADS valve 
testing is not a precursor to any accident previously evaluated and 
does not change the manner in which the ADS is operated. Thus, the 
proposed changes to the performance of SR 3.5.1.8 do not have any 
affect on the probability of an accident previously evaluated.
    The testing provides assurance that the ADS will functions as 
designed when actuated to depressurize the Primary Coolant System 
(PCS). The proposed changes to the surveillance requirement provide 
the same level of assurance regarding ADS reliability as the 
previous surveillance requirements. Accordingly, the consequences of 
an accident previously evaluated where the ADS was credited with 
mitigation is unchanged.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to SR 3.5.1.8 do not affect the performance 
of any LaSalle County Station structure, system, or component 
credited with mitigating any accident

[[Page 41619]]

previously evaluated since the proposed changes will provide the 
same level of confidence concerning the functioning of the ADS as 
the current requirements. Furthermore, the proposed changes do not 
install any new equipment, introduce any new modes of system 
operation or failure mechanisms.
    Therefore, the proposed changes do not create the possiblility 
of a new or different kind of accident from any previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to SR 3.5.1.8 will allow the uncoupling of 
the ADS valve lever from the other components associated with the 
manual actuation of the ADS valve. The proposed changes will allow 
the testing of the manual actuation electrical circuitry, manual 
actuation solenoid and air control valve, and the actuator without 
causing the ADS valve to open. The ADS valves will continue to be 
manually actuated by the bench-test valve control system of the 
setpoint testing program. The proposed changes do not affect the 
valve setpoint or the operational criteria that directs the ADS 
valves to be manually opened during plant transients.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Robert Helfrich, Senior Counsel, 
Nuclear, Mid-West Regional Operating Group, Exelon Generation Company, 
LLC, 1400 Opus Place, Suite 900, Downers Grove, IL 60515.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 23, 2001
    Description of amendment request: The proposed change would delete 
the loose parts monitoring system (LPMS) and the associated Technical 
Specifications (TSs) and Bases currently contained in the Limerick 
Generating Station, Units 1 and 2 Technical Specifications. The 
licensee bases its proposal to delete the LPMS on the conclusions of 
the Boiling Water Reactor Owners' Group Tropical Report NEDC-32975P, 
``Regulatory Relaxation for BWR Loose Parts Monitoring Systems''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This Technical Specification (TS) Change Request will delete the 
Loose Parts Monitoring System and the associated Technical 
Specifications and Bases currently contained in the Limerick 
Generating Station (LGS), Units 1 and 2, Technical Specifications. 
The Loose Parts Monitoring System (LPMS) is not an accident 
initiating system. The LPMS was designed in conformance with 
Regulatory Guide 1.133 (``Loose-Parts Detection Program for the 
Primary System of Light-Water-Cooled Reactors,'' Revision 1, May 
1981), to detect and alarm for loose parts in the reactor coolant 
system. A secondary function of the system is to assist the 
operators in locating the detected loose parts. The LPMS is used for 
information purposes only and is not a safety-related system. The 
operators do not rely solely on this system or information provided 
by this system for the performance of any safety-related action. 
Review of the Updated Final Safety Analysis (UFSAR) indicates that 
this system is not relied upon by other systems for input or data. 
This is a monitoring system that does not perform any automatic or 
control functions, and is not relied upon for any accident or 
transient evaluation. The removal of the LPMS from operation will 
not increase the need for operator intervention or increase operator 
burden to support any system used to mitigate an accident under 
normal or off normal conditions. Therefore, the proposed changes 
will not significantly increase the probability of an accident 
previously evaluated.
    The removal of the LPMS will not change or degrade the physical 
barriers or systems designed to contain radiation, and will have no 
affect on the on-site or off-site radiological conditions. 
Therefore, the proposed TS changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This TS Change Request will delete the Loose Parts Monitoring 
System and the associated Technical Specifications and Bases 
currently contained in the LGS, Units 1 and 2, Technical 
Specifications. Removal of this system will not create a new mode of 
operation of the plant. The LPMS is a nonsafety-related monitoring 
system. The proposed changes do not create a system-level failure 
mode different than those that already exist. In addition, there are 
no operation or failure modes of the LPMS that are accident 
initiators. Therefore, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    This TS Change Request does not affect any safety limits or 
analytical limits. Also there are no changes to accident or 
transient core thermal hydraulic conditions, or fuel or reactor 
coolant boundary design limits, as a result of these proposed 
changes. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: June 1, 2001.
    Description of amendment request: The proposed amendment would 
revise the Limerick Generating Station (LGS), Units 1 and 2, Technical 
Specifications (TS) 3.6.1.7 drywell average air temperature limit from 
135  deg.F to 145  deg.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed TS change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The increase in the allowable drywell average air temperature does 
not make any physical changes to the plant; it only permits the plant 
to operate at a higher drywell average air temperature, and therefore, 
does not increase the probability of an accident previously evaluated.
    The LGS Mark II containment design was evaluated during Power 
Rerate using an initial temperature of 150 degrees Fahrenheit for the 
Loss-of-Coolant Accident (LOCA) due to an instantaneous double-ended 
rupture of a recirculation suction line. The results of this evaluation 
showed that the drywell air temperature does not exceed the limit of 
340 degrees Fahrenheit post-accident and that the peak drywell pressure 
does not exceed the design

[[Page 41620]]

limit of 55 psig. In addition, the containment analysis performed for 
Power Rerate also bounds the small break LOCA.
    Since the proposed change allows a drywell air temperature that 
remains within the design analysis value, this proposed change does not 
increase the consequences of an accident previously evaluated in the 
Safety Analysis Report (SAR). This proposed change does not adversely 
affect mitigating systems, structures or components (SSC), and does not 
adversely affect the initial conditions of any accidents. Redundancy 
and diversity of mitigating systems are unchanged as a result of this 
proposed change. This proposed change does not affect onsite or offsite 
radiological consequences of any accident previously evaluated in the 
SAR.
    Based on the above discussion, this proposed TS change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The increase in the drywell average air temperature proposed by 
this TS change does not change any SSC of the plant. This TS change 
does not create new operating or failure modes.
    Based on the above discussion, this proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant reduction 
in the margin of safety.
    This proposed change will allow the plant to operate at a higher 
drywell average temperature during normal operation. This change does 
not create additional heat loads or change the way any of the equipment 
is operated. The equipment will remain within the limitations of the 
equipment qualification (EQ) program, which is qualified/maintained 
based on operation at an average annual temperature of 145 degrees 
Fahrenheit.
    Therefore, this proposed change does not involve a significant 
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: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of amendment request: March 28, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements to credit the 
soluble boron in the fuel storage pool analyses. This amendment would 
revise the index, modify TS 3.9.14, ``Fuel Storage--Spent Fuel Storage 
Pool,'' add TS 3.9.15, ``Fuel Storage Pool Boron Concentration,'' 
modify applicable Bases and revise Design Feature Section 5.3.1.1, 
``Criticality.'' TS 3.9.14 would be modified by separating this 
specification into two specifications to support crediting soluble 
boron in the fuel storage pool. The revised TS 3.9.14 would provide 
controls for fuel assembly enrichment and burnup in the spent fuel pool 
and also include an increase in the maximum enrichment from 4.85 weight 
percent (w/o) to 5.0 
w/o. A new TS 3.9.15 would provide control for soluble boron 
requirements in the spent fuel pool. Separating this specification into 
two specifications follows the general guidance provided in the 
improved standard TS (ISTS) of NUREG-1431.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Because of the Boraflex deterioration that has been observed, 
the spent fuel racks have been reanalyzed neglecting the presence of 
Boraflex to allow storage of Westinghouse 17 x 17 fuel assemblies 
with nominal enrichments up to 5.0 weight percent (w/o) using credit 
for checkerboarding, burnup and soluble boron. The proposed changes 
will not have a significant impact on the safety of the plant or on 
the spent fuel storage pool and are consistent with the NRC approved 
changes identified for other plants (i.e., Prairie Island Units 1 
and 2, Vogtle Units 1 and 2). Criteria set forth in Table 3.9-1 
provide qualification requirements for fuel assembly storage to 
ensure the NRC acceptance criteria and accident analysis assumptions 
are satisfied. Increasing the enrichment from 4.85 w/o up to and 
including 5.0 w/o U-235 [uranium 235] has minor effects on the 
radiological source terms and subsequently the potential releases, 
both normal and accidental, are not significantly affected.
    The proposed Technical Specification changes credit the use of 
soluble boron in the spent fuel pool criticality analyses. These 
criticality analyses were performed using the NRC approved 
methodology developed by the Westinghouse Owners Group (WOG) and 
described in WCAP-14416-NP-A, Revision 1, ``Westinghouse Spent Fuel 
Rack Criticality Analysis Methodology,'' November 1996. The analysis 
includes evaluations that factor in the axial burnup bias correction 
and utilizing identified conservatisms in the analysis demonstrate 
that Keff remains less than or equal to the design 
limits.
    The proposed changes do not involve a change to plant equipment 
and do not affect the performance of plant equipment used to 
mitigate an accident. They do not affect the operation of the spent 
fuel pool cooling system or any other system and are consistent with 
applicable analyses including [those associated with postulated] 
fuel handling accidents. They will not affect the ability of any 
system to perform its design function; therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    There are no hardware changes associated with this license 
amendment nor are there any changes in the method by which any 
safety-related plant system performs its safety function. No new 
accident scenarios, transient precursors, failure mechanisms or 
limiting single failures are introduced as a result of the proposed 
changes. The proposed changes do not introduce any adverse effects 
or challenges to any safety-related systems.
    The potential criticality accidents have been reanalyzed to 
demonstrate that the pool remains subcritical. Soluble boron has 
been maintained in the fuel storage pool water since its initial 
operation. The possibility of a fuel storage pool dilution is not 
affected by the proposed changes to the Technical Specifications. 
Therefore, implementation of Technical Specification controls for 
the soluble boron will not create the possibility of a new or 
different kind of accidental pool dilution.
    With credit for soluble boron now a major factor in controlling 
subcriticality, an evaluation of fuel storage pool dilution events 
was completed. This evaluation concluded that no credible events 
would result in a reduction of the criticality margin below the 5% 
margin recommended by the NRC. In addition, the No Soluble Boron 95/
95 probability/confidence level criticality analysis assures that 
dilution to 0 ppm [parts per million] will not result in 
criticality.
    The proposed Technical Specification changes ensure the 
maintenance of the fuel pool boron concentration and storage 
configuration. Therefore, the proposed changes will not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not affect the acceptance criteria for 
any analyzed event

[[Page 41621]]

nor impact any plant safety analyses since the analysis assumptions 
are not changed. The safety limits assumed in the accident analyses 
and the design function of the equipment required to mitigate the 
consequences of any postulated accidents will not be changed since 
the proposed changes do not affect equipment required to mitigate 
design basis accidents described in the Updated Final Safety 
Analysis Report. The Technical Specifications continue to assure 
that applicable operating parameters are maintained within required 
limits.
    The proposed changes to the fuel storage pool boron 
concentration and storage requirements will provide adequate margin 
to assure that the fuel storage array will always remain subcritical 
by the 5% margin recommended by the NRC. These limits are based on a 
criticality analysis performed in accordance with NRC approved 
Westinghouse fuel storage rack criticality analysis methodology.
    While criticality analysis utilized credit for soluble boron, 
the storage configurations have been defined using Keff 
calculations to ensure that the spent fuel rack Keff will 
be less than 1.0 with no soluble boron. Soluble boron credit is used 
to offset off-normal conditions (such as a misplaced assembly) and 
to provide subcritical margin such that the fuel storage pool 
Keff is maintained less than or equal to 0.95.
    The spent fuel pool boron dilution analysis concludes that an 
unplanned or inadvertent event which would result in dilution of the 
spent fuel pool boron concentration from 2000 ppm to 450 ppm is not 
a credible event. This conclusion is based on the substantial volume 
of unborated water required to dilute the pool and the fact that a 
large dilution event would be readily detected by plant personnel 
via alarms, flooding in the fuel handling building or detected 
during normal operator rounds through the spent fuel pool area.
    The margin of safety depends upon maintenance of specific 
operating parameters within design limits. The Technical 
Specifications continue to require that these limits be maintained 
and provide appropriate remedial actions if a limit is exceeded. The 
maintenance of these limits continues to be assured through 
performance of surveillances. Therefore, the plant will be 
maintained within the analyzed limits and the proposed changes will 
not involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: May 22, 2001.
    Description of amendment request: The proposed amendment would 
revise the once-through steam generator (OTSG) tube repair roll 
requirements to (1) Utilize updated limiting tensile tube loads, (2) 
define new exclusion zones within the steam generator in which the 
application of the repair roll is prohibited, (3) allow the repair roll 
to be used in the lower tubesheet area, (4) remove the limitation of 
only one repair roll per OTSG tube, and (5) replace the requirement 
that the repair roll be one inch in length with a requirement that the 
repair roll be installed in accordance with Framatome Technologies 
Incorporated Report BAW-2303P, revision 4, ``OTSG Repair Roll 
Qualification Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because testing and analysis have 
shown the once-through steam generator (OTSG) tube repair roll 
process under the proposed revised Technical specification (TS) 
Surveillance Requirement (SR) 4.4.5.4 ensures the new pressure 
boundary joint created by the repair roll process provides adequate 
structural and leakage integrity for all normal operating and 
accident conditions. In addition, the removal of the name ``Babcock 
& Wilcox'' is an administrative change to reflect that Framatome ANP 
has succeeded the Babcock & Wilcox Company. Therefore, the proposed 
changes to SR 4.4.5.4 will not increase the probability of a 
previously evaluated accident.
    The proposed change to TS Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not involve an increase in 
the probability of an accident previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the repair roll process under 
the proposed revised SR 4.4.5.4 ensures the new pressure boundary 
joint created by the repair roll process provides adequate 
structural and leakage integrity under all accident conditions. Any 
leakage resulting from repair roll joint slippage under accident 
conditions will be accounted for to ensure that the post-accident 
OTSG leakage will not exceed that assumed in the accident analyses. 
Should a repaired tube fail, the radiological consequences would be 
bounded by the existing Steam Generator Tube Rupture analysis.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not involve an increase to 
the consequences of an accident previously evaluated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because there will 
be no change in the operation of the steam generators or connecting 
systems as a result of the repair roll process added by the proposed 
changes to SR 4.4.5.4. The physical changes in the steam generators 
associated with the repair roll process have been evaluated and do 
not create the possibility for a new or different kind of accident 
from any accident previously evaluated, i.e., the physical change in 
the steam generators is limited to the location and accident slip 
behavior of the primary to secondary boundary within the tubesheet. 
Accordingly, these changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not create the possibility 
of any new or different kind of accident.
    3. Not involve a significant reduction in a margin of safety 
because tubes with primary system to secondary system boundary 
joints created by the repair roll process have been shown by testing 
and analysis to satisfy all structural, leakage, and heat transfer 
requirements. The additional testing of tubes repaired by the repair 
roll process under existing SR 4.4.5.9 provides continuing inservice 
monitoring of these tubes such that inservice degradation of tubes 
repaired by the repair roll process will be detected. Therefore, the 
changes to SR 4.4.5.4 to modify the repair process do not reduce and 
margin of safety.
    The proposed change to Bases 3/4.4.5 reflects the changes 
proposed to its associated SR, and does not reduce the margin of 
safety.
    On the basis of the above, the Davis-Besse Nuclear Power Station 
has determined that the License Amendment Request does not involve a 
significant hazards consideration. As this License Amendment Request 
concerns a proposed change to the Technical Specifications that must 
be reviewed by the Nuclear Regulatory Commission, this License 
Amendment Request does not constitute an unreviewed safety question.

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

[[Page 41622]]

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

    Date of amendment request: July 18, 2001.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.9.4, Containment Penetrations. TS 
3.9.4.a. requires that the containment equipment door be closed during 
core alterations or movement of irradiated fuel within containment. The 
proposed changes to TS 3.9.4.a. would allow the containment equipment 
door to be open during core alterations and movement of irradiated fuel 
in containment provided: (a) The equipment door is capable of being 
closed with four bolts, (b) the plant is in MODE 6 with at least 23 
feet of water above the reactor vessel flange, and (c) a designated 
crew is available to close the door. The basis for the proposed changes 
is a reanalysis of the limiting design basis Fuel Handling Accident, 
using an Alternate Source Term in accordance with 10 CFR 50.67 and 
Regulatory Guide (RG) 1.183.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to TS 3.9.4 would allow the containment 
equipment door and both doors of each containment airlock to be open 
during fuel movement or core alterations. Currently, the equipment 
door is closed with four (4) bolts during fuel movement or core 
alterations to prevent the escape of radioactive material in the 
event of an in-containment fuel handling accident. The containment 
equipment door is not an initiator of an accident. Whether the 
containment equipment door is open or closed during fuel movement 
and core alterations has no effect on the probability of any 
accident previously evaluated.
    Allowing the containment equipment door to be open during fuel 
movement or core alterations does not significantly increase the 
consequences from a fuel handling accident. The calculated offsite 
doses are well within the limits of 10 CFR Part 50.67 and RG 1.183. 
In addition, the calculated doses are larger than the expected doses 
because the calculation does not incorporate the closing of the 
containment equipment door after the containment is evacuated, which 
would occur in much less than the two hours assumed in the analysis.
    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 Technical Specifications 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.
    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 accident previously evaluated.
    The proposed changes to Technical Specification 3.9.4, 
``Containment Building Penetrations,'' affect a previously evaluated 
fuel handling accident. Both the current and the revised fuel 
handling accident analyses assume that all of the iodine and noble 
gases that become airborne, escape and reach the site boundary and 
low population zone with no credit taken for filtration, for the 
containment building barrier, or for decay or deposition. Since the 
proposed changes do not involve the addition or modification of 
equipment nor alter the design of plant systems, 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 margin of safety has not been significantly reduced. The 
calculated dose is well within the limits given in 10 CFR Part 50.67 
and RG 1.183. The proposed changes do not alter the bases for 
assurance that safety-related activities are performed correctly or 
the basis for any Technical Specification that is related to the 
establishment of or maintenance of a safety margin. 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: Patrick M. Madden.

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

    Date of amendment request: July 18, 2001.
    Description of amendment request: The proposed amendments would 
change the title of the corporate executive responsible for overall 
plant nuclear safety from ``President-Nuclear Division'' to ``Chief 
Nuclear Officer,'' in Technical Specification (TS) Section 6.0.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature, changing 
the title of the corporate executive responsible for overall plant 
nuclear safety, and would not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
These amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because they do not affect assumptions contained in plant safety 
analyses, the physical design and/or operation of the plant, nor do 
they affect TS that preserve safety analysis assumptions. Therefore, 
the proposed changes do not affect the probability or consequences 
of accidents previously analyzed.
    (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 accident previously evaluated.
    The proposed changes to the TS are administrative in nature, 
changing the title of the corporate executive responsible for 
overall plant nuclear safety in the Turkey Point Units 3 and 4 TS, 
and would not create the possibility of a new or different kind of 
accident from any previously evaluated. The proposed amendments will 
not change the physical plant or the modes of plant operation 
defined in the facility operating license. No new failure mode is 
introduced due to the administrative changes since the proposed 
changes do not involve the addition or modification of equipment, 
nor do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes are administrative in nature, changing the 
title of the corporate executive responsible for overall plant 
nuclear safety in the Turkey Point Units 3 and 4 TS, and would not 
reduce any of the margins of safety. The operating limits and 
functional capabilities of the affected systems, structures, and 
components remain unchanged by the proposed amendments.

    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

[[Page 41623]]

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: Patrick M. Madden.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: April 11, 2001.
    Description of amendment request: The proposed action would modify 
Technical Specification (TS) sections 4.2, ``Fuel Storage,'' and 5.6.5, 
``Spent Fuel Pool Water Chemistry Program,'' by adding applicability 
statements that these sections apply only when irradiated fuel is 
stored in the fuel storage pool. The applicability statements will 
allow timely dismantlement of the fuel storage pool following removal 
of the last irradiated fuel assembly from the fuel storage pool to the 
onsite independent spent fuel storage installation.
    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 change does not:

    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The requested license amendment involves addition of 
applicability statements to the design features for fuel storage and 
the program requirements for the spent fuel pool water chemistry 
program. These applicability statements will make the respective 
design features and program requirements applicable whenever 
irradiated fuel is stored in the fuel storage pool. Once irradiated 
fuel has been completely removed from the fuel storage pool and 
transferred to certified dry storage containers under a general [10 
CFR] Part 72 license, these design features and program requirements 
for the fuel storage pool are no longer necessary. These design 
features include: the maximum allowable Uranium-235 enrichment in 
fuel assemblies stored in racks; minimum acceptable margin to 
criticality allowed in the design of the spent fuel racks; the 
nominal fuel cell spacing in the spent fuel rack design; the minimum 
allowable drainage prevention design elevation; the fuel assembly 
loading capacity of the fuel storage pool and the specified storage 
locations within the fuel storage pool for different fuel 
enrichments and burnup periods; and the maximum allowable number of 
standard fuel assemblies in consolidated form. The program 
requirements consist of the establishment, implementation and 
maintenance of a water chemistry program for the fuel storage pool 
to minimize the potential effects of corrosion.
    The corresponding design features and program requirements for 
fuel storage in dry storage containers are specified in the 
container's certificate of conformance and safety analysis report. 
The corresponding design features currently include: fuel loading 
positions; fuel assembly limits including consolidated fuel and 
minimum cooling times versus burnup/initial enrichments. 
Descriptions of other design features of the UMS [Universal 
Multipurpose System] Storage System are found in the NAC [Nuclear 
Assurance Corporation]--UMS  SAR [Safety Analysis Report]. 
The corresponding program requirements currently include 
specifications for canister vacuum drying pressure and helium 
backfill pressure which ensure that a sufficiently inert environment 
is produced within the canister to preclude or inhibit corrosion.
    Since the design features and program requirements associated 
with fuel storage in the fuel storage pool do not significantly 
contribute to accident prevention or mitigation following the 
complete removal of irradiated fuel and since the corresponding 
design features and program requirements for fuel storage in dry 
storage containers are specified and controlled under other 
applicable license documents, these changes do not significantly 
increase the probability or the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The requested amendment involves the addition of applicability 
statements which will have the effect of making certain design 
features and program requirement associated with the fuel storage 
pool inapplicable when the fuel storage pool is no longer used for 
fuel storage. The corresponding design features and program 
requirements for fuel storage in dry storage containers are 
adequately specified in applicable license documents. The 
elimination of these design features and program requirements 
following complete removal of irradiated fuel from the fuel storage 
pool does not result in any new or different accident initiators 
from those already assumed in accidents previously evaluated, nor 
does it exacerbate any such accidents. Therefore, these changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The safety margins produced as a result of the specification of 
design features and program requirements for fuel storage in the 
fuel storage pool are adequately maintained in corresponding design 
features and program requirements associated with fuel storage in 
dry storage containers. These corresponding design features and 
program requirements are specified in the dry storage container's 
certificate of conformance and safety analysis report. Therefore, 
these changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Joseph Fay, Esquire, Maine Yankee Atomic 
Power Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: April 6, 2001.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 5.5.10, ``Technical Specifications 
Bases Control Program,'' to provide consistency with the changes to 10 
CFR 50.59 as published in the Federal Register (64 FR 53582) dated 
October 4, 1999. TS 5.5.10.b.2. would be revised to state: ``A change 
to the updated final safety analysis report or Bases that requires 
Nuclear Regulatory Commission approval pursuant to 10 CFR 50.59.'' In 
TS 5.5.10.b, a minor editorial change replaces the phrase ``changes do 
not involve'' with ``changes do not require.'' This change is 
consistent with the Nuclear Energy Institute Technical Specification 
Task Force (TSTF) Standard Technical Specification Change Traveler, 
TSTF-364 Revision 0, ``Revision to TS Bases Control Program to 
Incorporate Changes to 10 CFR 50.59.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will 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. 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 consequences of any accident previously evaluated are 
not significantly affected. Therefore, this change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

[[Page 41624]]

    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, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed change 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) will still 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 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: Al Gutterman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: September 26, 2000, as supplemented on 
October 6, 2000, and May 21, 2001. This notice supersedes a previous 
notice (65 FR 69065) published on November 15, 2000, that was based on 
the licensee's application for amendment dated September 26, 2000, as 
supplemented on October 6, 2000.
    Description of amendment request: The proposed change would amend 
the Salem Nuclear Generating Station (Salem) Unit Nos. 1 and 2 
Technical Specifications (TSs) to increase the as-found setpoint 
tolerance for the Pressurizer Safety Valves (PSV) from 1% 
to 3%; increase the as-found setpoint tolerance for the 
Main Steam Safety Valves (MSSV) from 1% to 3%; 
change the required action for reducing power when one or two MSSVs are 
inoperable; change the required action for three inoperable MSSVs to 
include a requirement to decrease the Power Range Neutron Flux High 
trip setpoint in addition to reducing power; and remove specifications 
and references related to plant operation with three Reactor Coolant 
System loops. The associated TS Bases sections will also be amended to 
reflect the TS 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 NRC staff's review is 
presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    Changing the pressurizer and main steam safety relief valve lift 
setpoint tolerance from 1% to 3% does not 
significantly increase the probability of any accident previously 
evaluated. The only events initiated by the opening of these safety 
valves are the accidental depressurization of the Reactor Coolant 
System and accidental depressurization of the Main Steam System. These 
events are a result of an inadvertent lifting of these valves and do 
not depend on the safety valve lift setpoint or tolerance. Therefore, 
the likelihood that either of these events will occur has not been 
increased.
    Analyses associated with the limiting overpressurization transients 
(Loss of External Electrical Load and/or Turbine Trip, and Single 
Reactor Coolant Pump Locked Rotor) have been performed that demonstrate 
that increasing the Pressurizer Safety Valve and Main Steam Safety 
valve lift setpoint tolerance to 3% would result in primary 
and secondary side pressure responses less than the acceptance criteria 
of 110% of the design pressure. Therefore, since the proposed setpoint 
tolerance increase would not adversely impact current accident analysis 
assumptions, the proposed change would not result in an increase in 
consequences of an accident previously evaluated.
    For operation with one or two inoperable main steam safety valves 
in one or more steam generators, changing the required action from a 
reduction of the power range high neutron flux trip setpoint to a 
reduction of the allowable reactor power level will not increase the 
consequences of any accident. With one or two inoperable Main Steam 
Safety Valves, the Loss of External Electrical Load and/or Turbine Trip 
event becomes limiting in terms of secondary side pressurization. The 
high flux trip does not provide any mitigation for this event. Other 
events limiting at power, that require the power range trip for 
mitigation, assume a safety analysis trip setpoint of 118% (based on a 
nominal trip setpoint of 109%) regardless of the initial power level. 
Therefore, the proposed change does not impact any of the accident 
analysis assumptions.
    During an RCCA (Reactor Cluster Control Assembly) Bank Withdrawal 
at Power event, the Main Steam Safety Valves may lift to ensure 
secondary side pressure remains below the allowable limit. This is 
especially true for events initiated from partial power conditions and 
slow reactivity insertion rates, where the reactor trip is from 
Overtemperature T (OTDT). Protection for this event is 
provided by a reactor trip on OTDT, not by the power range--high 
neutron flux trip. Thus, the proposed change does not affect the 
mitigative actions for this accident. Therefore, the consequences of an 
RCCA are unaffected.
    For three inoperable main steam safety valves in one or more steam 
generators, the addition of a requirement for a lower Power Range 
Neutron Flux High trip setpoint ensures the proposed change does not 
increase the consequences of this postulated accident.
    The current Salem licensing basis for the Spurious Activation of 
the Safety Injection System credits operator action to unblock a 
pressurizer Power Operated Relief Valve prior to the water solid 
pressurizer reaching the safety valve lift setpoint. The analyses that 
determined the time at which the safety valve would reach its pressure 
setpoint covered the -3% tolerance. Since this would conservatively 
result in the earliest opening time, there was no need to consider the 
positive side of the tolerance. The results of the analyses indicate 
that the allowable operator action time has not changed, such that 
water relief continues to occur through the Power Operated Relief 
Valves and not through the PSVs. As such the consequences of this event 
have not changed as a result of the proposed change.
    Increasing the MSSV lift setting tolerance may result in increased 
secondary side backpressure for the Auxiliary Feedwater Pumps. However, 
analyses have demonstrated that with the elevated backpressures that 
could result from increasing the MSSV setpoint upper tolerance to +3%, 
the Auxiliary Feedwater Pumps would still provide greater than the 
minimum flow required to mitigate events in which normal feedwater is 
not available, a Loss of Normal Feedwater and a Loss of Offsite Power 
to Station Auxiliaries.
    In terms of radiological consequences, the current design and 
licensing basis analyses that include steaming through the MSSV bound 
the proposed lift setpoint tolerance change.

[[Page 41625]]

    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposal will result in a change in the allowed Pressurizer 
Safety Valve and Main Steam Safety Valve lift setpoint tolerance range. 
No physical changes to these valves or to their nominal lift setpoint 
is required. These valves are assumed to malfunction only as the 
initiator for the accidental depressurization of the existing Reactor 
Coolant System or Main Steam System accident analyses. An increased 
lift setpoint tolerance range does not change the assumption of these 
depressurization events nor create a new type of event.
    Requiring a reduction in reactor thermal power or a reduction in 
reactor thermal power in conjunction with a reduction in Power Range 
Neutron Flux High Trip setpoint in the event of inoperable MSSV is 
consistent with the existing analysis assumptions. Initiation of any 
Salem Updated Final Safety Analysis Report (UFSAR) analyzed event at a 
power level less than full power is bounded by those events analyzed at 
full power, or specifically analyzed at the limiting power level, and 
does not constitute a new or different kind of accident. Also, no 
changes are being made to the power range high flux trip setpoint that 
will make it inconsistent with any analytical assumption.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    Analyses performed demonstrate that the proposed increase in the 
Pressurizer Safety Valve and MSSV lift pressure setpoint tolerance from 
1% to 3% will provide primary and secondary 
side pressure responses to the anticipated operational occurrences and 
design basis accidents that are within the existing margins of safety. 
The limiting overpressurization transients, Loss of External Electrical 
Load and/or Turbine Trip, and Single Reactor Coolant Pump Locked Rotor, 
stay well within the acceptance criteria of 110% of the design 
pressure.
    For operation with one or two inoperable MSSVs in one or more steam 
generators, the proposed reduction in reactor thermal power will ensure 
that current margins are maintained. The current requirement to reduce 
the power range high neutron flux trip setpoint does not affect the 
margin of safety since this trip does not provide any mitigation for 
the limiting secondary system pressurization event, Loss of External 
Electrical Load and/or Turbine Trip with one or two inoperable MSSVs.
    Specific accident analyses for RCCA Bank Withdrawal at Power 
scenarios demonstrate that a reactor trip on OTDT, in conjunction with 
the available relief capacity that exists with up to two inoperable 
safety relief valves on each steam generator, results in a secondary 
side pressurization within existing margins.
    For three inoperable MSSVs in one or more steam generators, thermal 
reactor power must be reduced in conjunction with a reduction in the 
Power Range Neutron Flux High trip setpoint to ensure pressurization of 
the main steam system remains within current analysis margins.
    The current licensing basis for the Spurious Activation of the 
Safety Injection System credits operator action to unblock a 
pressurizer Power Operated Relief Valve prior to the water solid 
pressurizer reaching the Pressurizer Safety Valve lift setpoint. As the 
PSVs are not designed for water relief, failure to unblock a Power 
Operated Relief Valve before reaching the Pressurizer Safety Valve lift 
setpoint would result in water relief and likely failure of the 
Pressurizer Safety Valve to reseat. This condition would escalate the 
Spurious Activation of the Safety Injection System (Condition II event) 
into a small break Loss Of Coolant Accident (Condition III event). The 
analyses that determined the time at which primary system pressure 
would reach the Pressurizer Safety Valve setpoint bound the -3% 
tolerance. Since the Pressurizer Safety Valve would not fail due to 
water relief, there is no reduction in the margin of safety for this 
event.
    Increasing the Main Steam Safety Valve lift setpoint tolerance may 
result in increased secondary side backpressure for the Auxiliary 
Feedwater System. However, analyses have demonstrated that under 
degraded Auxiliary Feedwater Pump performance, and with secondary side 
backpressure corresponding to 103% of the lowest MSSV setpoint, the 
Auxiliary Feedwater System can provide greater than the minimum flow 
required to mitigate those events where normal feedwater is not 
available, a Loss of Normal Feedwater and a Loss of Offsite Power to 
Station Auxiliaries.
    Therefore the proposed changes to the Technical Specifications do 
not involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 24, 2001.
    Description of amendment request: The Virgil C. Summer Nuclear 
Station (VCSNS) Technical Specifications (TS) Surveillance Requirement 
(SR) 4.7.1.2 would be revised to include the emergency feedwater system 
automatic isolation valves into the SRs. SR 4.7.1.2.b would include 
verification of the functional capability of the check valves in the 
instrument air system supplying the six new automatic isolation valves. 
SR 4.7.1.2.c.2 would include the six new automatic isolation valves 
into the requirement that assures critical valves can be closed and 
held closed when normal instrument air is unavailable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change addresses necessary changes to the VCSNS 
Technical Specification[s] (TS) 4.7.1.2.b and 4.7.1.2.c.2 associated 
with the installation of six new automatic isolation valves in the 
EF [emergency feedwater] system. The TS [need] to be changed to 
assure the same level of operability for the EF system as exists 
with the present day configuration.
    The only Final Safety Analysis Report (FSAR) analyzed accident 
for which the EF system could contribute as an initiator would be 
minor secondary line break, as described in Section 15.3.2. The 
addition of isolation valves in the EF piping to the steam 
generators will not increase the likelihood of a pipe break, since 
the addition will be in accordance with the same codes and standards 
as the corresponding, existing portions of the system. Piping stress 
analyses have demonstrated the addition of these valves does not 
result in the need to postulate any additional pipe breaks.

[[Page 41626]]

    The accidents analyzed in the FSAR, which rely on EF to mitigate 
consequences, are loss of normal feedwater, loss of off-site power, 
and major secondary system pipe ruptures. The addition of these 
automatic isolation valves will eliminate the need for operator 
action to manually close a flow control valve in response to a major 
secondary system line break. The elimination of operator manual 
action is accomplished by the addition of a new pneumatically 
operated isolation valve in series with each of the six existing 
flow control valves. Therefore, the change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change does not result in changes to actual 
operating pressures, flow rates, flow paths, or system interfaces. 
There are no alterations to system operability requirements. The 
existing system alarm setpoints are not affected, as is the 
information available to the operators. The addition of six new 
isolation valves will not change system design criteria and the 
surveillance testing will be the same as for the existing flow 
control valves.
    This change does not introduce any new or different kind of 
failure mechanisms or limiting single failures. Piping analysis has 
concluded that no new pipe break locations or break sizes will 
result from this change. Equipment protection features are not 
impacted, the frequency of pump and valve operation remains the 
same. Independence and redundancy are actually improved. Therefore, 
this proposed change would not create the possibility of an accident 
of a different type.
    3. Does this change involve a significant reduction in margin of 
safety?
    The design basis for the EF system is to assure the required 
flow and pressure to remove decay heat from the core under the worst 
postulated conditions. An additional function of the system is to 
isolate flow to a faulted SG [steam generator] within the time 
assumed in the safety analysis. The proposed change eliminates the 
need for operators to take actions to manually close the flow 
control valves in the event of a single failure.
    The proposed change will create a surveillance requirement for 
the new isolation valves that is the same as the existing flow 
control valves. The acceptance criteria will assure the operability 
of these valves. The design and installation of these isolation 
valves will maintain the requirements for independence, redundancy, 
separation and testability. The margins assumed in the safety 
analysis will be enhanced by this proposed change. Due to the 
automatic isolation capability, additional water will be available 
for the intact SGs and a reduced mass will be available to be 
released into the containment building. No credible single failure 
will be capable of preventing isolation of a faulted SG upon a high 
flow signal.

    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: Richard L. Emch, Jr.

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

    Date of amendment request: June 19, 2001.
    Description of amendment request: This request proposes to change 
Technical Specification (TS) Section 3.4.6.2, including its Bases, to 
increase the allowed operational leakage for Reactor Coolant System 
(RCS) Pressure Isolation Valves (PIV). The present criteria of 1 gallon 
per minute for all size valves would be changed to the industry 
standard of 0.5 gallons per minute per nominal inch of valve size, up 
to a maximum of 5 gallons per minute per valve, consistent with NUREG-
1431. This request also proposes to revise Table 3.4-1 to reflect the 
allowable leakage rates for each PIV.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. This proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change provides a more appropriate Pressure 
Isolation Valve (PIV) allowable leakage criteria in consideration of 
the safety significance and design capabilities of the plant as 
determined by the improved standard technical specification industry 
effort.
    The TS leakage limit for PIVs is 0.5 gallon per minute per 
nominal inch of valve size with a maximum limit of 5 gallons per 
minute. The previous criteria of 1 gallon per minute for all valve 
sizes imposed an unjustified penalty on the larger valves without 
providing information on potential valve degradation and can result 
in higher personnel radiation exposures due to unwarranted rework 
and retesting. An NRC sponsored study concluded a leakage rate limit 
based on the valve size was superior to a single allowable value.
    The revision to a leakage criterion related to valve size is 
acceptable because associated systems that have larger valves also 
have greater pressure relief capability. The new criteria allows for 
leakage above 1 gallon per minute, although limited to a maximum of 
5 gallons per minute, because the isolated low pressure system will 
not be overpressurized based on [its] relief capacity being greater 
than [its] allowed leakage limit. Therefore, the proposed change to 
the Limiting Condition for Operation will result in lower radiation 
exposures to personnel and a superior leak rate limit based on valve 
size as compared to a single allowable value.
    Since this proposed revision would continue to support the 
required safety functions, without modification to the plant 
features, neither the probability nor the consequences of an 
accident are increased.
    2. This proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed revision is not a result of changes to plant 
equipment, system design, testing methods, or operating practices. 
The modified LCO [limiting condition for operation] requirement will 
allow some relaxation of the current operability criteria for the 
PIVs, consistent with NUREG-1431. This change provides a more 
appropriate requirement in consideration of the safety significance 
and design capabilities of the plant as determined by the improved 
standard technical specification industry effort. Since the 
functions of the associated systems will continue to perform without 
change, the proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. This proposed change does not involve a significant reduction 
in a margin [of] safety.
    The proposed revision to the PIV leakage acceptance criteria 
will not result in changes to system design or setpoints that are 
intended to ensure timely identification of plant conditions that 
could be precursors to accidents or potential degradation of 
accident mitigation systems. These systems will continue to operate 
without change and only the associated allowable leakage criteria 
has been altered.
    Since the setpoints and design features that support the margin 
of safety are unchanged and actions for inoperable systems continue 
to provide appropriate time limits and compensatory measures, the 
proposed changes will not significantly reduce 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 G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

[[Page 41627]]

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: July 25, 2001 (TS-415).
    Description of amendment request: The proposed amendment would 
delete Technical Specification Action Statement 3.3.1.1.I.2, which 
limits plant operation to 120 days in the event of the inoperability of 
the Oscillation Power Range Monitor (OPRM) trip system at Browns Ferry 
Nuclear Plant, Unit 3 (BFN). For this situation, the proposed change 
would allow plant operation to continue if the existing TS Required 
Action 3.3.1.1.I.1, to implement an alternate means to detect and 
suppress thermal hydraulic instability oscillations, were taken.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The OPRM function is not considered as an initiator of any 
previously analyzed accident. Therefore, this proposed change does 
not significantly increase the probability of such accidents. This 
proposed change would allow the use of existing well-established 
alternate methods to detect and suppress thermal hydraulic 
instability oscillations. Considering that multiple Boiling Water 
Reactors plants, including BFN, have satisfactorily operated using 
alternate stability monitoring methods for extended periods of 
operation prior to the installation of OPRM systems, it is concluded 
these measures are adequate. Therefore, the consequences of a 
previously analyzed accident would not be significantly increased.
    B. The proposed amendment 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, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    This proposed change would allow the use of existing alternate 
methods to detect and suppress thermal hydraulic instability 
oscillations to continue to operate the reactor in the event of the 
inoperability of the OPRM system. Considering that multiple Boiling 
Water Reactors plants, including BFN, have satisfactorily operated 
using alternate stability monitoring methods for extended periods of 
operation, it is concluded these measures are adequate, and that the 
proposed change does not significantly reduce 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Patrick M. Madden (Acting).

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/NRC/ADAMS/index.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, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: March 1, 2001, as supplemented 
on June 27, 2001.
    Brief description of amendment: The proposed amendment revised the 
Technical Specifications (TSs) to change the frequency of closure time 
testing of the main steam isolation valves (MSIVs). These tests may now 
be conducted during each cold shutdown unless this test has been 
performed within the past 92 days.
    Date of Issuance: July 17, 2001.
    Effective date: 7/17/01 and shall be implemented within 30 days of 
issuance.
    Amendment No.: 221.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
19999).
    The June 27, 2001, letter provided ``camera-ready'' TS pages and 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 17, 2001.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1

    Date of amendment request: October 5, 2000, as supplemented by 
letters dated June 27, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification Sections 3.2.1.a, 3.2.1.e, and 3.2.1.f to relocate 
administrative controls to the Quality Assurance Program Description.
    Date of issuance: July 23, 2001.
    Effective date: 30 days from the date of issuance.
    Amendment No: 49.

[[Page 41628]]

    Facility Operating License No. DPR-5: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71134) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 23, 2001.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: May 8, 2001.
    Brief description of amendment: The amendment revises the frequency 
of the Technical Specification (TS) surveillance requirement to check 
the movement of the control rods. Specifically, the frequency listed 
for this requirement in TS Table 4.1-3, ``Frequencies for Equipment 
Tests,'' is changed from ``every 31 days'' to ``quarterly during 
reactor critical operations.''
    Date of issuance: July 18, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 217.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31704).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendments: December 27, 2000, as 
supplemented on March 28, April 12, June 9, June 13, and June 29 (3), 
2001. The addition of a Technical Specification (TS) Bases control 
program was requested on March 28, 2001.
    Brief description of amendments: These amendments allow: (1) 
Revisions to reactor trip and engineered safety feature actuation 
setpoints and allowable values, (2) implementation of the revised 
thermal design procedure, (3) relocations of TS requirements to the 
core operating limits report, (4) relocation of TS requirements to the 
licensee requirements manual, (5) miscellaneous editorial changes. In 
addition, License Condition 2.(C).(3) regarding less than 3-loop 
operation was deleted.
    Date of issuance: July 20, 2001.
    Effective date: Immediately and to be implemented within 120 days.
    Amendment Nos.: 239 and 120.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20002) for the December 27, 2000, amendment request. A portion of a 
March 28, 2001, amendment request was also issued in this amendment. 
The date of the initial notice for the March 28, 2001, amendment 
request was June 20, 2001 (66 FR 33111).
    The March 28, April 12, June 9, June 13, and June 29 (3), 2001, 
letters provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination and did not 
expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 20, 2001.
    No significant hazards consideration comments received: No.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: January 4, 2001, as supplemented by 
letters dated March 12 and April 4, 2001.
    Brief description of amendment: The amendment revises License 
Condition 2.B.(6)(d) to reference revisions to the Physical Security 
Plan, Guard Training and Qualification Plan, and Safeguards Contingency 
Plan.
    Date of issuance: July 25, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. DPR-36: The amendment revised the 
License.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13805).
    The March 12, 2001, supplemental letter superseded certain aspects 
of the January 4, 2001, amendment request, as described in the original 
Federal Register notice (FRN), but did not change the initial no 
significant hazards consideration determination (NSHCD). The April 4, 
2001, supplemental letter provided clarifying information that did not 
change the scope of the original FRN or the initial NSHCD.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 25, 2001.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of application for amendment: February 5, 2001; as 
supplemented on April 19, 2001.
    Brief description of amendment: The amendment revises Section 
3.6.1.3, ``Primary Containment Isolation Valves,'' those portions 
regarding requirements for excess flow check valve surveillance 
testing.
    Date of issuance: July 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from Refueling Outage 8, currently scheduled for 
approximately spring 2002.
    Amendment No.: 96.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15927).
    The April 19, 2001, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated July 12, 2001.
    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: December 5, 2000, as 
supplemented June 28, 2001.
    Brief description of amendment: The amendment implements 
programmatic controls for radiological effluent technical 
specifications (RETS) in the administrative section of the Technical 
Specifications (TSs) and relocates the procedural details of the RETS 
to the offsite dose calculation manual, the process control program, or 
other new programs, consistent with the guidance of Standard TSs 
(NUREG-1433) and Nuclear Regulatory Commission Generic Letter 89-01.
    Date of issuance: July 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 120.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9385).

[[Page 41629]]

    The June 28, 2001, supplement provided corrected TS pages to 
reflect the inclusion of amendments approved subsequent to the December 
5, 2000, application, to correct a typographical error on one TS page, 
and to make a terminology change from ``site'' to ``unit'' on one TS 
page. The supplemental information did not change the initial no 
significant hazards consideration determination and did not expand the 
scope of the original Federal Register notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 24, 2001.
    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: December 13, 2000, as 
supplemented July 3, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification 3.8/4.8 to clarify the air ejector offgas activity sample 
point and operability requirements.
    Date of issuance: July 25, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 121.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7685).
    The supplemental information did not change the initial no 
significant hazards consideration determination and did not expand the 
scope of the original Federal Register notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 25, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 4, 2001.
    Brief description of amendments: The amendments deletes Technical 
Specifications (TS) Section 5.5.3, ``Post Accident Sampling,'' for 
Diablo Canyon Nuclear Power Plant, Units 1 and 2, and thereby eliminate 
the requirements to have and maintain the post-accident sampling 
systems (PASS). The amendment for Unit 1 also deletes PASS-related 
License Condition 2.C(6).e from Facility Operating License DPR-80.
    Date of issuance: July 13, 2001.
    Effective date: July 13, 2001, to be implemented within 90 days 
from the date of issuance.
    Amendment Nos.: Unit 1--149; Unit 2--149.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications and Facility Operating License 
DPR-80.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31712).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 13, 2001.
    No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: October 23, 2000, and 
supplemental letters dated January 11 and April 16, 2001.
    Brief description of amendment: The amendment deletes the 
definitions of Site Boundary and Unrestricted Area from the technical 
specifications and makes related conforming changes.
    Date of issuance: July 13, 2001.
    Effective date: July 13, 2001, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 128.
    Facility Operating License No. DPR-54: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13806).
    The January 11 and April 16, 2001, supplemental letters provided 
additional clarifying information, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 2001.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: May 14, 2001 (TS 01-02).
    Brief description of amendments: This amendment revised License 
Condition 2.C.(9)(d) in Operating License DPR-77 for the Sequoyah 
Nuclear Plant. The revised license condition now references a licensee 
letter that specifies a minimum voltage threshold for steam generator 
tube eddy current inspections.
    Date of issuance: July 18, 2001.
    Effective date: July 18, 2001.
    Amendment Nos.: 270.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the Operating License.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29362).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 2001.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a

[[Page 41630]]

nuclear power plant or in prevention of either resumption of operation 
or of increase in power output up to the plant's licensed power level, 
the Commission may not have had an opportunity to provide for public 
comment on its no significant hazards consideration determination. In 
such case, the license amendment has been issued without opportunity 
for comment. If there has been some time for public comment but less 
than 30 days, the Commission may provide an opportunity for public 
comment. If comments have been requested, it is so stated. In either 
event, the State has been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 Assess and Management Systems 
(ADAMS) Public Electronic Reading Room on the internet at the NRC Web 
site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-415-4737 or by email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By September 7, 2001, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). 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.
    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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington,

[[Page 41631]]

DC 20555-001, and 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: July 25, 2001.
    Brief description of amendment: The proposed amendment deletes TS 
Required Action 3.3.1.1.I.2, which limits plant operation to 120 days 
in the event of the inoperability of the Oscillation Power Range 
Monitor trip system. For this situation, the proposed change would 
allow plant operation to continue if the existing TS Required Action 
3.3.1.1.I.1, to implement an alternate means to detect and suppress 
thermal hydraulic instability oscillations, were taken.
    Date of issuance: July 25, 2001.
    Effective date: July 25, 2001.
    Amendment No.: 273.
    Facility Operating License No. NPF-90: Amendment revises the TS. 
The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration, are contained in a Safety Evaluation dated July 
25, 2001.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Section Chief: Patrick M. Madden (Acting).

    Dated at Rockville, Maryland, this 31st day of July 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-19746 Filed 8-7-01; 8:45 am]
BILLING CODE 7590-01-P