[Federal Register Volume 66, Number 182 (Wednesday, September 19, 2001)]
[Notices]
[Pages 48283-48295]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-23209]


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NUCLEAR REGULATORY COMMISSION


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

    (Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.)

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 August 27, 2001 through September 7, 2001. 
The last biweekly notice was published on September 5, 2001 (66 FR 
46473).

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)

[[Page 48284]]

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 October 19, 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, 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

[[Page 48285]]

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].

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

    Date of amendment request: July 23, 2001.
    Description of amendment request: As a follow-up response to a 
commitment identified in the Nuclear Regulatory Commission (NRC) staff 
letter dated December 22, 2000, ``Completion of Licensing Action for 
Generic Letter (GL) 96-06, Assurance of Equipment Operability and 
Containment Integrity During Design-Basis Accident Conditions,'' 
Entergy Operations Inc., (Entergy, the licensee) has proposed to revise 
their Waterford Steam Electric Station, Unit 3 (Waterford 3) Final 
Safety Analysis Report (FSAR) to resolve the ten containment 
penetrations susceptible to thermally induces overpressurization 
through an evaluation, detailed analysis, or installation of physical 
modifications prior to startup from the spring 2002 refueling outage. 
Entergy determined a change to Waterford 3's license basis, through 
procedural controls, risk analysis, and engineering analysis, for seven 
penetrations, as discussed in this license basis change request. 
Permanent resolution to the GL 96-06 issues for the remaining three 
penetrations could be satisfied through the installation of physical 
modifications.
    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 FSAR change reflects the use of administrative 
procedural controls to ensure these seven containment penetrations 
(two 4-inch diameter Steam Generator Blowdown penetrations and five-
\1/2\ inch diameter Process Sampling penetrations) contain fluid at 
temperatures representative of Reactor Coolant, and the very low 
probability for overpressurization failure of containment 
penetrations during Mode 4 plant operation as a permanent solution 
to the GL 96-06 issue. The engineering analysis determined these 
seven containment penetrations met the acceptance criteria for 
allowed stresses contained in ASME [American Society of Mechanical 
Engineers] Section III Code, [Boiler and Pressure Vessel Code] 
Appendix F 1995. The result of the risk analysis is such that the 
very small change in LERF (Large Early Release Frequency], on the 
order of 1 x 10-9 per reactor year, remained 
well below the 1 x 10-7 LERF 
guideline for a small change given in Regulatory Guide 1.174. The 
negligible reduction in LERF that would be achieved by adding 
thermal relief valve overpressure protection is not risk significant 
and is too small to justify the addition of the relief valves.
    With respect to the probability or the consequences of an 
accident previously evaluated in the FSAR, the proposed deviation to 
the existing ASME Section III Code, Class 2 design provisions and 
operating requirements for the seven containment penetrations would 
not significantly increase the probability of an accident since the 
administrative procedural controls are being provided to: (1) 
minimize penetration heat-up and over-pressurization during a small 
window of vulnerability, approximately 1% per year of Mode 4 plant 
operation; and (2) minimize process fluid cooldown during normal 
plant operation by closing the containment isolation valves for the 
five sample penetrations when process fluid samples are obtained and 
the laboratory sample valves downstream of the CIV [containment 
isolation valves] are closed or flow through the penetration is 
stopped. Also the results of engineering analyses showed that the 
containment penetrations may exceed ASME Section III, Subsection NC 
3500 Code required yield stresses and experience plastic 
deformation, but would not catastrophically fail; therefore, the 
penetrations would retain their ability to perform their safety 
function and maintain containment integrity.
    On this basis, the proposed changes are not considered to 
constitute a significant increase in the probability or consequences 
of an accident due to:
     Administrative controls to minimize penetration heat-up 
and over-pressurization during the small window of vulnerability
     The seven containment penetrations retaining their 
ability to perform their safety function and maintaining containment 
integrity in accordance with engineering analyses performed that met 
acceptance criteria for allowed stresses contained in ASME Section 
III Code, Appendix F 1995, and
     The low risk significance of overpressurization failure 
of the seven containment penetrations during a DBA [Design Basis 
Accident] while the plant is in Mode 4.
    The proposed changes will not significantly affect the results 
of any accident previously evaluated. The accident mitigation 
features of the plant are not significantly affected by these 
proposed changes. The proposed changes do not add or modify any 
existing equipment.
    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 change proposes a deviation to the existing ASME, Section 
III, Class 2 license basis requirements for portions of the Steam 
Generator Blowdown System, Primary Sampling System, and Secondary 
Sampling System that penetrate the containment as a permanent 
solution to the GL 96-06 issues. This change involves recognition of 
the acceptability of administrative procedural controls to minimize 
penetration heat-up and over-pressurization during the small window 
of vulnerability, approximately 1% per year for Mode 4 plant 
operation. Added assurance is provided through the engineering 
analysis performed on these penetrations that determined allowable 
stresses did not exceed the ASME Section III Code, Appendix F 1995 
pipe stress values. Therefore, the change would not contribute to 
the possibility of, or be the initiator for any new or different 
kind of accident.
    The proposed change does not alter the configuration of the 
plant. There has been no physical change to plant systems, 
structures, or components.
    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 change does not involve a significant reduction in 
margin of safety. The existing licensing basis for Waterford 3, with 
respect to the ASME Section III, Subsection NC-3621.2 provisions for 
portions of the Steam Generator Blowdown System, Primary Sampling 
System, and Secondary Sampling System that penetrate the 
containment, is to ensure piping that has the potential to 
experience pressurization due to trapped fluid expansion shall be 
designed to withstand the increased pressure or have provisions for 
relieving the excess pressure piping. With the acceptance of this 
proposed deviation to the license basis, it will be recognized that 
the seven containment penetrations have administrative procedural 
controls to minimize penetration heat-up and over-pressurization 
during the small window of vulnerability, approximately 1% per year 
for Mode 4 plant operation. Added assurance is also provided through 
the engineering analysis performed on these penetrations that

[[Page 48286]]

determined stresses did not exceed the ASME Section III Code, 
Appendix F 1995 pipe stress values and predicted the penetration 
piping would experience plastic deformation, but would not 
catastrophically fail. Therefore, the penetrations would retain 
their ability to perform their safety function and maintain 
containment integrity. This deviation to license basis requirements 
for these seven containment penetrations is not considered to 
constitute a significant decrease in the margin of safety.
    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.

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 amendment request: August 13, 2001.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications (TSs) to maintain a 
Post-Accident Sampling System (PASS). Licensees were generally required 
to implement PASS upgrades as described in NUREG-0737, ``Clarification 
of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. 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 August 13, 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 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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Timothy G. Colburn, Acting.

[[Page 48287]]

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

    Date of amendment request: August 22, 2001.
    Description of amendment request: Florida Power and Light Company 
(FPL) requests to amend Facility Operating Licenses DPR-67 for St. 
Lucie Unit I and NPF-16 for St. Lucie Unit 2 by revising Technical 
Specifications (TS) relating to positive reactivity additions while in 
shutdown modes. The proposed changes clarify TS involving positive 
reactivity additions to the shutdown reactor, and would allow small, 
controlled, safe insertions of positive reactivity while in shutdown 
modes. The proposed changes conform closely to an NRC approved generic 
change for Standard Technical Specifications, known as TSTF-286 Rev. 2, 
which revises most actions requiring ``Suspend operations involving 
positive reactivity additions'' to allow minimum reactivity additions 
due to temperature fluctuations or operations, which are necessary to 
maintain fluid inventory within the required shutdown margin or 
refueling boron concentration, as applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed TS changes revise actions that either require 
suspension of operations involving positive reactivity additions or 
preclude reduction in boron concentration less than the reactor 
coolant system (RCS). Reactivity excursions are analyzed events. The 
proposed changes limit positive reactivity additions into the RCS 
such that the required shutdown margin (SDM) or refueling boron 
concentration continue to be met. Reactivity changes performed 
during shutdown modes are currently governed by strict 
administrative controls. Although the proposed changes will allow 
procedural flexibility with regards to RCS temperature and boron 
concentration, these operations will still be under administrative 
control. The changes proposed by these amendments are within the 
scope and assumptions of the existing analyses. 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 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 revisions relate to positive reactivity 
additions while in shutdown modes of operation. Reactivity 
excursions are analyzed events. The operational flexibility allowed 
in these proposed license amendments will be performed under strict 
administrative controls in order to limit the potential for excess 
positive reactivity addition. Although the existing procedural 
controls will need modification, no new or different operational 
failure modes would be introduced by these changes.
    Additionally, implementation of these proposed changes do not 
require any physical plant modifications, so no new or different 
hardware related failure modes are introduced. The changes proposed 
by these amendments are within the scope and assumptions of the 
existing analyses. Therefore, 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.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed changes conform closely to the industry and NRC 
approved TSTF-286, Rev. 2 and relate to small, controlled, safe 
insertions of positive reactivity additions while in shutdown modes. 
These changes revise actions that either require suspension of 
operations involving positive reactivity additions, or prohibit RCS 
boron concentration reduction. The proposed changes provide 
operational flexibility while controlling positive reactivity 
additions in order to preserve the required SDM or refueling boron 
concentration. The proposed changes to provide for continued safe 
reactor operations, while also limiting any potential for excess 
positive reactivity addition. 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: Richard P. Correia.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of amendment request: June 22, 2001, as supplemented August 
24, 2001.
    Description of amendment request: The proposed amendment would 
revise the St. Lucie Unit 2 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. TS 3.9.4.b. requires a minimum of 
one door in each airlock to be closed during core alterations or 
movement of irradiated fuel within containment. The proposed change 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 within 30 minutes, (b) the plant is in MODE 6 with at least 23 
feet of water above the reactor pressure vessel flange, and (c) a 
designated crew is available at the equipment door to close the door. 
The capability to close the containment equipment door includes the 
requirements that the door is capable of being closed and that any 
cables or hoses across the equipment door have quick-disconnects to 
ensure the door is capable of being closed in a timely manner. The 
proposed change to TS 3.9.4.b would allow both doors of each 
containment airlock to be open during core alterations and movement of 
irradiated fuel in containment provided: (a) At least one door of each 
open containment airlock is capable of being closed, (b) the plant is 
in MODE 6 with at least 23 feet of water above the reactor pressure 
vessel flange, and (c) a designated individual is available outside 
each open containment airlock to close the door. The capability to 
close the containment airlock door includes the requirement that the 
door is capable of being closed and that any cables or hoses across the 
airlock door have quick-disconnects to ensure the door is capable of 
being closed in a timely manner.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change 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 and a single door on each 
containment airlock is closed during fuel movement or core 
alterations to prevent the escape of radioactive material in the

[[Page 48288]]

event of an in-containment fuel handling accident. Neither the 
containment equipment door nor either of the containment airlock 
doors is an initiator of an accident. Whether the containment 
equipment door or both doors of the containment air locks are open 
or closed during fuel movement and core alterations has no affect on 
the probability of any accident previously evaluated. Allowing the 
containment equipment door and the containment airlock doors 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 
100. In addition, the calculated doses are larger than the expected 
doses because the calculation does not incorporate the closing of 
the containment equipment door or the containment airlock doors 
after the containment is evacuated, which would be much less than 
the two hours assumed in the analysis. The proposed change would 
significantly reduce the dose to workers in containment in the event 
of a fuel handling accident by reducing the time required to 
evacuate the containment. 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 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change to Technical Specification 3.9.4, 
``Containment Building Penetrations,'' affects a previously 
evaluated fuel handling accident. The new Fuel Handling Accident 
Analysis assumes that all of the iodine and noble gases that become 
airborne escape and reach the exclusion boundary and low population 
zone with no credit taken for filtration, the containment building 
barrier or for decay or deposition. Since the proposed change does 
not involve the addition or modification of equipment nor does it 
alter the design of plant systems and the revised analysis is 
consistent with the Fuel Handling Accident Analysis, the proposed 
change does 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 a margin of 
safety.
    The margin of safety as defined by 10 CFR part 100 has not been 
significantly reduced. The calculated dose is well within the limits 
given in 10 CFR part 100 or NUREG 0800. The proposed change does 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 amendment would not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: February 28, 2001.
    Description of amendment request: The proposed amendment to the 
Cooper Nuclear Station (CNS) Operating License DPR-46 would revise the 
design basis accidents (DBA) radiological assessment methodology for 
offsite and control room radiological doses, and the associated 
supporting Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revisions to the CNS DBA radiological assessment 
methodology for offsite and control room doses, and the associated 
supporting TS changes, do not involve initiators or precursors of 
accidents previously evaluated. Furthermore, these changes do not 
affect the design, function, or modes of operation of systems, 
structures, or components within the facility. Therefore, the 
proposed radiological assessment calculational methodology revisions 
and TS changes do not involve a significant increase in the 
probability of an accident previously evaluated in the Updated 
Safety Analysis Report (USAR).
    The proposed revisions to the CNS DBA radiological assessment 
methodology for offsite and control room doses, and the associated 
supporting TS changes, do not affect the design, function or modes 
of operation of systems, structures or components in the facility. 
The calculation revisions utilize conservatively lower accident 
mitigation system filter efficiency assumptions and incorporate 
plant specific accident mitigation system operating parameter and 
design assumptions. Due to the changes in the calculational 
methodology and assumptions, and an increase in the postulated 
accident source term, the calculated radiological dose consequences 
of each DBA have changed and in some cases increased. In each case, 
however, the calculated radiological dose consequences are within 
the exclusion area boundary (EAB) and low population zone (LPZ) 
radiological dose acceptance criteria specified in 10 CFR part 100 
and the control room dose acceptance criteria discussed in General 
Design Criterion (GDC) 19 of 10 CFR part 50, Appendix A. Therefore, 
the proposed revisions to the radiological assessment methodology, 
and associated TS changes, do not involve a significant increase in 
the consequences of an accident previously evaluated in the USAR.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The proposed revisions to the CNS DBA radiological assessment 
methodology for offsite and control room doses, and the associated 
supporting TS changes, do not affect the design, function or mode of 
operation of systems, structures or components in the facility such 
that new equipment failure modes are created. No new or different 
type of plant equipment is installed by the revised radiological 
assessment calculational methodology or changes to the TS. Neither 
the calculations nor the TS changes introduce changes to existing 
design parameters governing normal plant operation or new plant 
operating modes. No new types of accident initiators or precursors 
are created by the proposed revisions. Therefore, the proposed 
revisions to radiological assessment methodology and the proposed 
changes to the TS do not create the possibility of a new or 
different kind of accident previously evaluated in the USAR.
    3. Does not create a significant reduction in the margin of 
safety.
    The proposed revisions to the CNS DBA radiological assessment 
methodology for offsite and control room doses, and the associated 
supporting TS changes, do not affect the design, function or mode of 
operation of systems, structures or components in the facility. 
These proposed TS changes are consistent with the criteria of 10 CFR 
50.36(c)(2)(ii) for TS content.
    The proposed revisions will not result in any challenges to 
plant equipment, fuel integrity, or the reactor coolant system 
pressure boundary. Due to the changes in the calculational 
methodology and assumptions, and an increase in the postulated 
accident source term, the calculated radiological dose consequences 
of each design basis accident have changed and in some cases 
increased. In each case, however, the calculated radiological dose 
consequences are within the EAB and LPZ radiological dose acceptance 
criteria specified in 10 CFR part 100 and the control room dose 
acceptance criteria discussed in GDC 19 of 10 CFR part 50, Appendix 
A. Therefore, the proposed revisions to the radiological assessment 
methodology, and associated TS 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

[[Page 48289]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 12, 2001.
    Description of amendment request: The proposed amendment would 
change the Cooper Nuclear Station (CNS) Technical Specification (TS) 
5.5.10.b.2 to replace the phrase, ``A change to the updated FSAR or 
Bases that involves an unreviewed safety question as defined in 10 CFR 
50.59'' with the phrase ``A change to the updated FSAR or Bases that 
requires NRC approval pursuant 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. Do the proposed changes 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 
revisions to 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. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed change will not reduce the 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 revised 10 
CFR 50.59 (c)(2) will still require NRC approval pursuant to 10 CFR 
50.59. This change is administrative in nature as discussed by the 
NRC in FR (Volume 64, Number 191, Pages 53582-53617) dated October 
4, 1999, docketing the change 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April, 12, 2001.
    Description of amendment request: The amendment request would 
modify the Cooper Nuclear Station (CNS) Technical Specifications 
Surveillance Requirement (SR) 3.6.1.3.8 to relax the SR frequency by 
allowing a representative sample of Excess Flow Check Valves (EFCVs) to 
be tested every 18 months, such that each EFCV will be tested once 
every 10 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The current SR frequency requires each reactor instrumentation 
line EFCV to be tested every 18 months. The EFCVs at CNS are 
designed to close automatically in the event of a line break 
downstream of the valve. This proposed change allows a reduced 
number of EFCVs to be tested every 18 months. Industry operating 
experience, documented in BWR [Boiling Water Reactor] Owners' Group 
Topical Report NEDO-32977-A [``Excess Flow Check Valve Testing 
Relaxation,'' dated June 2000], concludes that a change in 
surveillance test frequency has a minimal impact on the reliability 
for these valves. A failure of an EFCV to isolate cannot initiate 
previously evaluated accidents. Furthermore, neither the EFCV 
actuation test, nor the frequency of testing is considered an 
initiator of any analyzed event. Therefore, there is no increase in 
the probability of occurrence of an accident as a result of this 
proposed change.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed for the analysis, and the 
availability and successful functioning of the equipment assumed to 
operate in response to the analyzed event, and the setpoints at 
which these actions are initiated. This change does not affect the 
performance of any credited equipment. The installed restricting 
orifice on each associated instrument line provides assurance that 
any instrument line break will limit offsite doses to substantially 
below 10 CFR part 100 values. Neither the EFCV actuation test, nor 
the frequency of testing is an analysis assumption. Therefore, there 
is no increase in the previously evaluated consequences of the 
rupture of an instrument line and there is no potential increase in 
the radiological consequences of an accident previously evaluated as 
a result of this change.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change allows a reduced number of EFCVs to be 
tested each operating cycle. No other changes in requirements are 
being proposed. Industry operating experience as documented in [BWR 
Owners' Group Topical Report NEDO-32977-A] provides supporting 
evidence that the reduced testing frequency will not affect the high 
reliability of these valves. The potential failure of an EFCV to 
isolate as a result of the proposed reduction in test frequency is 
bounded by the previous evaluation of an instrument line pipe break. 
This change will not physically alter the plant (no new or different 
type of equipment will be installed). This change will not alter the 
operation of process variables, structures, systems, or components 
as described in the safety analysis. Thus, a new or different kind 
of accident will not be created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. EFCV design, operation, and flow actuation criteria 
remain unaffected by this change. Restricting orifices for each 
associated instrument line remains available to mitigate an 
instrument line break. The proposed change, which impacts the 
frequency of testing EFCVs is acceptable because the tests continue 
to require appropriate confirmation of the assumed function of the 
system (and thereby assure continued operability), and has been 
shown to reflect an acceptable frequency for detecting failures. 
There is no detrimental impact on any other equipment design 
parameter, and the plant will still be required to operate within 
prescribed limits. Therefore, the 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.

[[Page 48290]]

    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: August 9, 2001.
    Description of amendment request: The amendment would change the 
Seabrook Station Technical Specifications (TSs) Index, TS 3/4.9.3 
(``Decay Time''), TS 3/4.9.4 (``Containment Building Penetrations''), 
and TS 3/4.9.9 (``Containment Purge And Exhaust Isolation System''). 
The amendment would also change Bases 3/4.9.3, Bases 3/4.9.4, and Bases 
3/4.9.9 for consistency with the proposed TS changes. These changes are 
consistent with the improved Standard Technical Specifications (STS) 
for Westinghouse plants.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 
3/4.9.9 do not adversely affect accident initiators or precursors 
nor do they adversely alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. In addition, the proposed changes do not 
adversely affect the manner in which the plant responds in normal 
operation, transient or accident conditions nor do they change any 
of the procedures related to operation of the plant. Though a 
portion of the proposed change to TS 3/4.9.4 appears to be a 
relaxation to the current licensing basis, North Atlantic has 
incorporated administrative conservatism into TS 3/4.9.4 to assure 
the proposed changes, in conjunction with other TS required 
surveillance testing, do not alter or prevent the ability of 
structures, systems and components (SSCs), in particular the 
Containment Purge and Exhaust Isolation System, to perform its 
intended function to mitigate the consequences of an initiating 
event within the acceptance limits assumed in the Updated Final 
Safety Analysis Report (UFSAR).
    The proposed changes do not adversely affect the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated in the Seabrook Station UFSAR. Further, the proposed 
changes do not increase the types and amounts of radioactive 
effluent that may be released offsite, nor significantly increase 
individual or cumulative occupational/public radiation exposures.
    Therefore, it is concluded that these proposed revisions to TS 
Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not involve a 
significant increase in the probability or consequence of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    This proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and 
TS 3/4.9.9 do not adversely affect the operation nor do they change 
the design basis of any plant system or component during normal or 
accident conditions. The proposed changes do not include any 
physical changes to the plant. In addition, the proposed changes do 
not adversely affect the function or operation of plant equipment or 
introduce any new failure mechanisms such that the design basis is 
adversely affected. The current licensing basis allows penetration 
isolation by manual or automatic means. The plant equipment will 
continue to respond per the design and analyses and there will not 
be a malfunction of a new or different type introduced by the 
proposed changes that creates the possibility of a new or different 
kind of accident.
    The proposed changes do not modify the facility nor do they 
adversely affect the plant's response to normal, transient or 
accident conditions. The changes do not introduce a new mode of 
plant operation. While these changes may afford North Atlantic 
operational flexibility, the changes are an enhancement and do not 
affect plant safety. The plant's design and design basis are not 
revised and the current safety analyses remains in effect.
    Thus, these proposed revisions to TS Index, TS 3/4.9.3, TS 3/
4.9.4, and TS 3/4.9.9 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.
    The proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS 
3/4.9.9 do not adversely affect the safety margins established 
through Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety Limits as specified in the Technical 
Specifications nor is the plant design revised by the proposed 
changes. The current licensing basis allows penetration isolation by 
manual or automatic means.
    Though a portion of the proposed change to TS 3/4.9.4 appears to 
be a relaxation to the current licensing basis, North Atlantic has 
incorporated administrative conservatism into TS 3/4.9.4 to ensure 
the proposed changes, in conjunction with other TS required 
surveillance testing, offset any potential minimal reduction in the 
margin of safety. North Atlantic believes that the proposed change 
to TS 3/4.9.4 is more conservative than that currently allowed in 
the improved STS, NUREG-1431, Revision 2.
    Thus, it is concluded that these proposed revisions to TS Index, 
TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: August 15, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to (1) reflect the 
replacement of Monticello's licensed operator initial and 
requalification training programs with an accredited systems approach 
to training program and (2) relocate the existing TS requirements for 
procedures, records, and reviews to the operational quality assurance 
plan.
    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 changes are administrative in nature and compliance 
with applicable regulatory requirements will continue to be 
maintained. The proposed changes do not involve any change to the 
configuration or alter existing system relationships. In addition, 
the proposed changes do not alter the conditions or assumptions in 
any of the previous accident analyses thus, the radiological 
consequences previously evaluated are not adversely affected by the 
proposed changes.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by the proposed amendment.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed changes are administrative in nature and compliance 
with applicable regulatory requirements will continue to be 
maintained. The proposed changes do not involve any change to the 
configuration or method of operation of any plant equipment. 
Accordingly, no new failure modes have been introduced for any plant 
system or component important to safety nor has any new limiting 
single failure been identified as a result of the proposed changes. 
Also, there

[[Page 48291]]

will be no changes in types or increases in the amounts of any 
effluents released offsite.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated will not be created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes are administrative in nature and do not 
involve any change in the methodology or method of operation of any 
plant equipment. The proposed changes do not involve any change to 
the configuration or alter existing system relationships. The 
appropriate controls to provide continued assurance of compliance to 
applicable regulatory requirements has been maintained.
    Therefore, the proposed amendment will 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: August 31, 2001.
    Description of amendment request: The proposed amendment would 
amend the licenses to change the required implementation date for 
previously issued Amendment No. 184 to Facility Operating License NPF-
14 and Amendment No. 158 to Facility Operating License NPF-22. The 
proposed amendment would not alter any of the requirements of the 
Susquehanna Steam Electric Station (SSES) Unit 1 and 2 Technical 
Specifications (TSs). The previously issued amendments incorporate 
long-term power stability solution instrumentation into the SSES Unit 1 
and 2 TSs. When implemented, these amendments will incorporate into the 
TSs the licensee's final response to GL 94-02, ``Long Term Solutions 
and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic 
Instabilities in Boiling Water Reactors.'' Specifically, these 
amendments will, in part, add TS requirements related to the operating 
power range monitoring (OPRM) system. The licensee stated that recently 
identified deficiencies in the OPRM trip setpoint methodology, as 
documented in a General Electric 10 CFR part 21 report issued on June 
29, 2001, have adversely affected its ability to implement the subject 
amendments. Therefore, the licensee requested that the required 
implementation date for Amendment No. 184 to License No. NPF-14 and 
Amendment No. 158 to License No. NPF-22 be revised to become effective 
no later than November 1, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment implementation date extension is 
administrative in nature and does not require any physical plant 
modifications, physically affect any plant systems or components, or 
entail changes in plant operation. The resulting consequences of 
transients and accidents will remain within the NRC approved 
criteria. Therefore, the proposed action does not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment implementation date extension is 
administrative in nature and does not require any physical plant 
modifications, physically affect any plant systems or components, or 
entail changes in plant operation. 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 change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment implementation date extension is 
administrative in nature and does not require any physical plant 
modifications, physically affect any plant systems or components, 
nor entail changes in plant operation. Since the proposed changes do 
not affect the physical plant or have any impact on plant operation, 
the proposed changes will not jeopardize or degrade the function or 
operation of any plant system or component. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Peter Tam, Acting.

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

    Date of amendment request: August 17, 2001.
    Description of amendment request: The proposed amendments would 
revise the reactor vessel pressure-temperature (P-T) limits depicted in 
Technical Specification Figure 3.4.9-1 for each unit. In addition, 
pursuant to 10 CFR 50.12, TVA is requesting an exemption from the 
requirements of 10 CFR part 50, Appendix G, to allow the use of 
American Society of Mechanical Engineers (ASME) Code Case N-640 as a 
basis for these revised curves. Code Case N-640, ``Alternative 
Requirement Fracture Toughness for Development of P-T Limit Curves for 
ASME Boiler and Pressure Vessel Code Section XI, Division 1,'' permits 
the use of the plane strain fracture toughness (KIc) curve 
instead of the crack arrest fracture toughness (KIa) curve 
for reactor pressure vessel materials in determining the P-T limits. 
The exemption request is being reviewed separately.
    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 proposed Units 2 and 3 change deals exclusively with the 
reactor vessel pressure-temperature (P-T) curves which define the 
permissible regions for operation and testing. Failure of the 
reactor vessel is not considered as a design basis accident. Through 
the design conservatisms used to calculate the P-T curves, reactor 
vessel failure has a low probability of occurrence and is not 
considered in the safety analyses. The proposed changes adjust the 
reference temperature for the limiting material to account for 
irradiation effects and provide the same level of protection as 
previously evaluated and approved. The adjusted reference 
temperature calculations were performed using the guidance contained 
in Regulatory Guide 1.99, Revision 2, and ASME Section XI Code Case 
N-640 to reflect use of the operating limits to 19.5 Effective Full 
Power Years (EFPY). These changes do not alter or prevent the 
operation of equipment required to mitigate any accident analyzed in 
the BFN Final Safety Analysis Report. Therefore, this change does 
not increase the probability or consequences of any previously 
evaluated accident.

[[Page 48292]]

    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 to the Units 2 and 3 reactor vessel P-T 
curves does not involve a modification to plant equipment. No new 
failure modes are introduced. There is no effect on the function of 
any plant system, and no new system interactions are introduced by 
this change. 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.
    The proposed curves conform to the guidance contained in 
Regulatory Guide 1.99, Revision 2, and maintain the safety margins 
specified in 10 CFR 50, Appendix G. Therefore, the proposed 
amendment does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Richard P. Correia.

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

    Date of amendment request: August 7, 2001 (TS-01-04).
    Description of amendment request: The proposed amendment would add 
a new condition and associated actions to the Technical Specification 
Limiting Condition for Operation (LCO) 3.8.1, ``AC Sources Operating,'' 
to allow one Diesel Generator (DG) be out of service for 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The emergency DGs are designed as backup AC power sources in the 
event of loss of offsite power. The proposed AOT [allowed outage 
time] does not change the conditions, operating configurations, or 
minimum amount of operating equipment assumed in the safety analysis 
for accident mitigation. No changes are proposed in the manner in 
which the DGs provide plant protection or which create new modes of 
plant operation. In addition, a Probabilistic Safety Analysis (PSA) 
evaluation concluded that the risk contribution of the AOT extension 
is non-risk significant. Therefore, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    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 introduce any new modes of plant 
operation or make physical changes to plant systems. Therefore, 
extension of the allowable AOT for DGs does not create the 
possibility of a new or different accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The DGs are designed as backup AC power sources in the event of 
loss of offsite power. The proposed AOT does not change the 
conditions, operating configurations, or minimum amount of operating 
equipment assumed in the safety analysis for accident mitigation. No 
changes are proposed in the manner in which the DGs provide plant 
protection or which create new modes of plant operation. In 
addition, a PSA evaluation concluded that the risk contribution of 
the AOT extension is non-risk significant. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

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

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: August 20, 2001.
    Description of amendment request: The proposed change to the 
Technical Specifications (TSs) would revise certain requirements 
associated with demonstrating the operability of alternate trains when 
redundant equipment is made or found to be inoperable. The TSs revised 
include: 4.4.B, 4.5.A.2, 4.5.A.3, 4.5.A.4, 4.5.B.2, 4.5.C.2, 4.5.C.3, 
4.5.D.2, 4.5.D.3, 4.5.E.2, 4.5.F.2, 4.5.H.1, 4.7.B.3.c, 4.10.B.1, and 
4.10.B.3.b.2. Some format and typographical errors are also being 
corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Because changing surveillance test requirements does not change 
the probability of accident precursors, this proposed change does 
not affect the probability of an accident previously evaluated. 
Since other periodic and post-maintenance surveillance requirements 
ensure that the operability of systems and components is maintained, 
there is no significant increase in the consequences of accidents 
previously evaluated.
    Furthermore, the removal of the additional surveillance testing 
from the Technical Specifications would result in a decrease in the 
probability of equipment failure because the excessive testing 
causes unnecessary wear on the safety-related equipment and 
unnecessary challenges to safety systems. Reduced testing may also 
eliminate the potential for human error associated with system 
alignments and misdirection of attention from monitoring and 
directing plant operations.
    Administrative changes to the Technical Specifications do not 
alter any technical requirements, and as such, do not increase the 
probability or consequences of accidents.
    Therefore, the proposed change will not increase the probability 
or consequences of any accident previously evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Reduced surveillance testing does not create new or different 
kinds of accidents since modes of operation are unchanged and 
additional accident precursors are not introduced. System 
operability requirements and design bases remain the same, and 
reactor operations are unchanged. Since system and component testing 
only involves the assurance of operability, reduced testing does not 
introduce mechanisms that may contribute to the possibility of new 
or different kinds of accidents.
    Administrative changes to the Technical Specifications do not 
alter any technical requirements, and as such, do not create the 
possibility of new or different kinds of accidents.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    The proposed change will not decrease operability requirements, 
nor reduce the equipment required during various plant conditions. 
An acceptable level of testing exists in other Technical 
Specification requirements to demonstrate system and component 
operability. There are no changes to system or component operability 
requirements; therefore, systems and

[[Page 48293]]

components will be available to provide existing margins of safety. 
The same systems and components with the same performance levels 
assumed in safety analyses will still be available to mitigate 
consequences of postulated accidents.
    Administrative changes to the Technical Specifications do not 
alter any technical requirements, and as such, have no effect on 
margins of safety.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

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

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, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: March 29, 2001, as supplemented 
by letters dated June 27, 2001, and July 24, 2001.
    Brief description of amendment: The amendment revised the reactor 
coolant system heatup, cooldown, and inservice leak hydrostatic test 
limitations for the reactor coolant system to a maximum of 29 effective 
full power years in accordance with Title 10 of the Code of Federal 
Regulations, Part 50, Appendix G. These pressure-temperature (P-T) 
limits are contained in TMI Unit 1 Technical Specification (TS) 3.1.2. 
In addition, the amendment revised the low-temperature overpressure 
protection (LTOP) requirements in TSs 3.1.12 and 4.5.2 to reflect the 
revised P-T limits. These changes will allow operation of two reactor 
coolant pumps in a single loop during LTOP conditions.
    Date of issuance: September 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38758).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2001.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Docket No. 
72-8, Calvert Cliffs Independent Spent Fuel Storage Installation, 
Calvert County, Maryland

    Date of application for amendments: November 22, 1999, as 
supplemented by letters dated October 4 and November 10, 2000, and May 
18, 2001.
    Brief description of amendments: The amendments authorize revisions 
to the Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis 
Report and Independent Spent Fuel Storage Installation Updated Safety 
Analysis Report to incorporate changes associated with the aircraft 
hazards analysis due to increased ``random'' military flights in the 
vicinity of these facilities. These changes constitute an unreviewed 
safety question as defined in 10 CFR 50.59 and 10 CFR 72.48.
    Date of issuance: August 29, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 246 and 221.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69 and 
Materials License No. SNM-2502: Amendments revised licenses.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73085).
    The supplemental letters dated October 4 and November 10, 2000, and 
May 18, 2001, provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of these amendments is contained in 
a Safety Evaluation dated August 29, 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: December 11, 2000.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to incorporate editorial revisions, 
clarifications, and corrections. Specifically, the amendment: (1) 
Provides updated information and corrections to the TS cover page, 
table of contents, and list of figures, (2) revises TS 4.5.E, ``Control 
Room Air Filtration System,'' to remove an incorrect system test 
description and provide consistent test values for system flow rate and 
filter efficiency, (3) revises TS 6.2.1.a, ``Facility Management and

[[Page 48294]]

Technical Support,'' to reference the Quality Assurance Program 
Description as the location of the documentation rather than the 
Updated Final Safety Analysis Report, (4) revises TS 6.9.1.7, ``Monthly 
Operating Report,'' to change the recipient of the Monthly Operating 
Report, and (5) corrects the periodicity of the Radioactive Effluent 
Release Report from semi-annual to annual in TS 6.15, ``Offsite Dose 
Calculation Manual'' and TS 6.16, ``Major Changes to Radioactive 
Liquid, Gaseous and Solid Waste Systems.'' In addition, the amendment 
revises TS Figure 5.1-1B concerning the indicated vent location 
associated with Indian Point Unit 3 (IP3). The labels for the IP3 plant 
vent and the machine shop were reversed.
    Date of issuance: August 29, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 219.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11057).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 23, 2001, as supplemented 
June 25, June 29, and July 19, 2001.
    Brief description of amendment: The amendment revises pressure-
temperature limit curves and cold overpressure protection limits.
    Date of issuance: August 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance. August 27, 2001.
    Amendment No.: 197.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36340).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 27, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, PSEG Nuclear LLC, and Atlantic City Electric 
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
Station (PBAPS), Units 2 and 3, York County, Pennsylvania

    Date of application for amendments: April 3, 2001.
    Brief description of amendments: The amendments revised the PBAPS 
Units 2 and 3 Technical Specifications (TSs) to incorporate Technical 
Specification Task Force (TSTF) Item 258, Revision 4. TSTFs are changes 
to the improved standard TS that were initiated by the nuclear power 
industry and submitted to the NRC staff. TSTF-258, Revision 4, revises 
TS Section 5.0, Administrative Controls, to delete specific TS staffing 
requirements for licensed Reactor Operators (ROs) and Senior Reactor 
Operators (SROs), relocate the working hour limits to a plant 
procedure, clarify requirements for the Shift Technical Advisor 
position, add regulatory definitions for ROs and SROs, revise the 
Radioactive Effluent Controls Program to be consistent with the intent 
of 10 CFR Part 20, and revises radiological area control requirements 
for high radiation areas to be consistent with 10 CFR 20.1601(c).
    Date of issuance: August 30, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendments Nos.: 240 and 243.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31708).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 7, 2001, as supplemented 
April 25, June 20, and July 16, 2001.
    Brief description of amendment: The amendment revised the Improved 
Technical Specifications (ITS) 5.6.2.20, ``Containment Leakage Rate 
Testing Program'' to allow a one-time interval increase for the Type A 
Integrated Leakage Rate Test for no more than 5 years.
    Date of issuance: August 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 197.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2001 (66 FR 
17967). The supplemental letters provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 1, 2000.
    Brief description of amendments: The amendment approves changes to 
the Updated Final Safety Analysis Report (UFSAR) regarding the modeling 
of the pressurizer heater operation and spray effectiveness as they 
relate to certain transients that are analyzed for pressurizer 
overfill. Specifically, the amendment approves a change to the 
moderator temperature coefficient currently in the UFSAR assumed as an 
initial condition for the loss of all nonemergency alternating current 
power and loss of normal feedwater transients.
    Date of issuance: August 23, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 237.
    Facility Operating License No. DPR-74: Amendment revised the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56953).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 23, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 18, 2001, as 
supplemented April 20, 2001.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.10.m to 
increase the minimum reactor coolant flow from

[[Page 48295]]

85,500 gallons per minute (gpm) flow per loop to 93,000 gpm flow per 
loop.
    Date of issuance: September 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 157.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11062).
    The April 20, 2001, supplemental information contained clarifying 
information and 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 September 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 18, 2001.
    Brief description of amendment: The amendment deleted items 3 and 4 
from Section 5.15, ``Post-Accident Radiological Sampling and 
Monitoring,'' of the Fort Calhoun Station, Unit No. 1 Technical 
Specifications, and thereby eliminates the requirements to have and 
maintain the post-accident sampling system (PASS).
    Date of issuance: August 29, 2001.
    Effective date: August 29, 2001, and shall be implemented within 
120 days from the date of issuance.
    Amendment No.: 200.
    Facility Operating License No. DPR-40. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38765).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 11, 2001, as supplemented 
June 13, 2001.
    Brief description of amendment: The amendment revises the Hope 
Creek Technical Specifications (TSs) to relax the frequency for testing 
of excess flow check valves (EFCVs). Specifically, TS surveillance 
requirement 4.6.3.4 has been changed to revise required testing of 
EFCVs from once per 18 months for all valves to a test of a 
representative sample each 18 months such that all valves are tested 
once in 10 years.
    Date of issuance: August 28, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented during Refueling Outage 10, currently scheduled to commence 
in October 2001.
    Amendment No.: 132.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in  Federal Register: May 30, 2001 (66 FR 
29361).
    The June 13, 2001, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 22, 2000.
    Brief Description of amendments: These amendments revise the 
Facility Operating Licenses ( FOLs) and the Technical Specifications 
(TS) to remove obsolete license conditions, make editorial changes in 
the FOLs, and implement associated changes to the TS and Bases.
    Date of issuance: August 30, 2001.
    Effective date: August 30, 2001.
    Amendment Nos.: 227 and 227.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the License and Technical Specifications.
    Date of initial notice in  Federal Register: November 1, 2000 (65 
FR 65351).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 30, 2001.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: March 22, 2001.
    Brief description of amendment: The amendment (1) decreases the 
allowable values for Function 8, pressurizer pressure-low and 
pressurizer pressure-high, in Table 3.3.1-1, ``Reactor Trip System 
Instrumentation,'' and (2) increases the allowable value for Function 
1.d, pressurizer pressure-low for safety injection, in Table 3.3.2-1, 
``Engineered Safety Feature Actuation System Instrumentation.''
    Date of issuance: August 30, 2001.
    Effective date: August 30, 2001, and shall be implemented prior to 
entry into Mode 3 in the restart from refueling outage 12 scheduled for 
the Spring 2002.
    Amendment No.: 140.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: May 2, 2001 (66 FR 
22035).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 30, 2001.
    No significant hazards consideration comments received: No.


    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.


    Dated at Rockville, Maryland, this 10th day of September, 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-23209 Filed 9-18-01; 8:45 am]
BILLING CODE 7590-01-P