Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 41442-41449 [E5-3793]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 137 (Tuesday, July 19, 2005)]
[Notices]
[Pages 41442-41449]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-3793]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and 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 June 24 to July 7, 2005. The last biweekly 
notice was published on July 5, 2005 (70 FR 38712).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and

[[Page 41443]]

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 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by email to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by email to pdr@nrc.gov.

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

    Date of amendment request: May 27, 2005.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) testing frequency for the 
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
Specifically, the proposed change would revise the frequency for SR 
3.1.4.2, ``Control Rod Scram Time Testing,'' from ``120 days cumulative 
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC

[[Page 41444]]

determination in its application dated May 27, 2005. 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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: February 25, 2005.
    Description of amendment request: The proposed change would modify 
the Millstone Power Station, Unit No. 2 Technical Specification (TS) 
Surveillance Requirement for trisodium phosphate (TSP) to remove the 
granularity term and chemical detail. In addition, the proposed change 
will increase the allowed outage time from 48 to 72 hours. Basis for 
proposed no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. Does the proposed [license] amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The TSP stored in containment is designed to buffer the acids 
expected to be produced after a loss of coolant accident and is 
credited in the radiological analysis for iodine retention. The type 
and amount of TSP is not considered to be an initiator of any 
analyzed accident. The proposed change does not modify any plant 
equipment and only clarifies language used in a TSP surveillance 
requirement which does not impact any failure modes that could lead 
to an accident. Removing the detail for TSP granularity and type 
from the surveillance and increasing the allowed outage time, does 
not change the solubility or buffering capability of the TSP. 
Therefore this change does not impact the consequences of any 
accident. Based on this discussion, the proposed amendment does not 
increase the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed [license] amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The TSP chemical in containment is not being modified in any way 
by this proposed amendment. There is no impact on the capability of 
the TSP to increase the sump water pH to 7 or greater after a loss 
of coolant accident. No parameters of the TSP baskets are being 
modified and no changes are being made to the method in which 
borated water is delivered to the sump. The proposed changes to 
remove the terms ``granular'' and ``dodecahydrate,'' and to increase 
the allowed outage time do not introduce any new failure modes for 
the containment sump system. Removing the detail from the 
surveillance requirement will clarify that the intended parameter to 
be measured is volume. The proposed amendment does not introduce 
accident initiators or malfunctions that would cause a new or 
different kind of accident. Therefore, the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed [license] amendment involve a significant 
reduction in a margin of safety?
    Response: No.
    There is no significant reduction in the established margin of 
safety posed by the proposed change to remove detail from the TSP 
surveillance requirement and increase the allowed outage time. The 
TSP in containment provides the necessary pH control following a 
loss of coolant accident to assure iodine retention. Consequently 
iodine concentrations in the containment atmosphere are maintained 
within the assumptions of the offsite dose calculations. The 
proposed change does not introduce any new requirements for the TSP 
chemical used in containment that would impact a margin of safety. 
The allowed outage time of 72 hours is consistent with other 
emergency core cooling components which are also required to perform 
during a loss of coolant accident. 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

 Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of amendment request: April 27, 2005
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) related to the safety-related 
battery systems. The revision is based on TS Task Force (TSTF) Change 
Traveler TSTF-360, Revision 1, ``Direct Current (DC) Electrical 
Rewrite,'' and would revise TSs for inoperable battery chargers, 
provide alternative testing criteria for battery charger testing, and 
revise TSs for battery cell monitoring. Basis for proposed no 
significant hazards consideration determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of no 
significant hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The DC Sources and Battery Cell Parameters are not initiators of 
any accident sequence analyzed in JAFNPP's Updated Final Safety 
Analysis Report (UFSAR). As such, the proposed changes do not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The initial conditions of the Design Basis Accident (DBA) and 
transient analyses in JAFNPP's UFSAR assume Engineered Safety 
Feature (ESF) systems are operable. The DC electrical power 
distribution system is designed to provide sufficient capacity, 
capability, redundancy, and reliability to

[[Page 41445]]

ensure the availability of necessary power to ESF systems so that 
the fuel, reactor coolant system, and containment design limits are 
not exceeded. The operability of the DC electrical power 
distribution system in accordance with the proposed TS is consistent 
with the initial assumptions of the accident analyses and is based 
upon meeting the design basis of the plant. Therefore, the proposed 
changes do not involve a significant increase in the consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical alteration of 
the JAFNPP. The temporary charger, when placed in service, will be 
powered from an emergency bus and have appropriate electrical 
isolation. Installed equipment is not being operated in a new or 
different manner. There are no setpoints at which protective or 
mitigative actions are initiated that are affected by the proposed 
changes. The operability of the DC electrical power distribution 
system in accordance with the proposed TS is consistent with the 
initial assumptions of the accident analyses and is based upon 
meeting the design basis of the plant. These proposed changes will 
not alter the manner in which equipment operation is initiated, nor 
will the functional demands on credited equipment be changed. No 
alteration in the procedures, which ensure the unit remains within 
analyzed limits, is proposed, and no change is being made to 
procedures relied upon to respond to an off-normal event. As such, 
no new failure modes are being introduced. The proposed changes do 
not alter assumptions made in the safety analyses. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes will not adversely affect operation of 
plant equipment. These changes will not result in a change to the 
setpoints at which protective actions are initiated. Sufficient DC 
capacity to support operation of mitigation equipment is ensured. 
The changes associated with the new administrative TS program will 
ensure that the station batteries are maintained in a highly 
reliable manner. The equipment fed by the DC electrical power 
distribution system will continue to provide adequate power to 
safety-related loads in accordance with analyses assumptions. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, and 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois

    Date of amendment request: June 15, 2005.
    Description of amendment request: Exelon Generation Company, LLC 
(EGC), plans to transition to Westinghouse SVEA-96 Optima2 fuel at 
Dresden Nuclear Power Station (DNPS) and Quad Cities Nuclear Power 
Station (QCNPS) beginning with the QCNPS Unit 2 refueling outage in 
March 2006. Specifically, EGC requests approval of revisions to 
Technical Specifications (TSs) Section 3.1.4, ``Control Rod Scram 
Times,'' TS Section 4.2.1, sbull I11``Fuel Assemblies,'' and TS Section 
5.6.5, ``Core Operating Limits Report (COLR),'' to support this 
transition. The core reload analyses using the new Westinghouse 
analytical methods for the affected units may result in the need for 
additional TS changes to support the transition to SVEA-96 Optima2 
fuel, such as a change to the safety limit minimum critical power 
ratio. These changes, if any, will be submitted to the NRC in a 
separate license amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change has no effect on any accident initiator or 
precursor previously evaluated and does not change the manner in 
which the core is operated. The type of fuel is not a precursor to 
any accident. The new methodologies for determining core operating 
limits have been validated to ensure that the output accurately 
models predicted core behavior, and use of the methodologies will be 
within the ranges previously approved. The new methodologies being 
referenced will have all been submitted to the NRC, and have either 
been approved or are currently under NRC review. Those methodologies 
that are currently under NRC review are scheduled to receive NRC 
approval prior to the first use of SVEA-96 Optima2 fuel in a reload 
core at either DNPS or QCNPS.
    There is no change in the consequences of an accident previously 
evaluated. The proposed change in the administratively controlled 
analytical methods does not affect the ability to successfully 
respond to previously evaluated accidents and does not affect 
radiological assumptions used in the evaluations. Source term from 
SVEA-96 Optima2 fuel will be bounded by the source term assumed in 
the accident analyses. There is no effect on the type or amount of 
radiation released, and there is no effect on predicted offsite 
doses in the event of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not affect the performance of any DNPS 
or QCNPS structure, system, or component credited with mitigating 
any accident previously evaluated. The use of new analytical 
methods, which have either been reviewed and approved by the NRC or 
are currently being reviewed by the NRC, for the design of a core 
reload will not affect the control parameters governing unit 
operation or response of plant equipment to transient conditions. 
The proposed change does not introduce any new modes of system 
operation or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to TS 3.1.4 clarifies that analyses for 
design basis accidents and transients will continue to support the 
scram times listed in TS Table 3.1.4-1, independent of whether 
General Electric analyzes the core. The proposed change does not 
alter the acceptance criteria for control rod scram times. Future 
core reloads will be analyzed using the NRC-approved methodology for 
modeling control rod insertion during a scram. The proposed change 
to TS Section 4.2.1 revises the description of fuel assemblies to 
envelope the SVEA-96 Optima2 fuel characteristics. The proposed 
change to TS Section 5.6.5 adds new analytical methods for design an 
analysis of core reloads to the list of methods currently used to 
determine the core operating limits. The NRC has either previously 
approved the analytical methods being added, or is currently 
reviewing the methods.
    The proposed change does not modify the safety limits or 
setpoints at which protective actions are initiated, and does not 
change the requirements governing operation or availability of 
safety equipment assumed to operate to preserve the margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 41446]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: April 4, 2005.
    Description of amendment request: In order to support the steam 
generator replacement project (SGRP), the proposed amendment would 
temporarily revise the Operating License to allow the licensee to 
operate with one of the two recently installed 18-inch diameter 
penetrations through the Shield Building dome to be opened while the 
unit is in Modes 1-4. Either of the Shield Building penetrations will 
be allowed to be opened for a combined total of up to 5 hours a day, 6 
days a week while in Modes 1-4 during the portion of the ongoing Cycle 
7 operation between receipt of NRC approval and Mode 5 at the start of 
the Cycle 7 refueling outage. The technical specifications will revert 
to the pre-amendment requirements prior to entering Mode 4 during 
startup from the Cycle 7 outage, since work activities related to the 
SGRP will permanently eliminate these penetrations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The bounding transients and accidents (i.e., loss-of-coolant-
accident (LOCA), tornado, and earthquake) that are potentially 
affected by the assumptions associated with the use of one of the 
Shield Building dome penetrations have been evaluated/analyzed. 
Weather and seismic related events are determined by regional 
conditions. Therefore, the probability of a tornado or earthquake is 
not affected by the use of one of the Shield Building dome 
penetrations. Failure of the Shield Building or emergency gas 
treatment system (EGTS) is not an initiator of any of the accidents 
and transients described in the Updated Final Safety Analysis Report 
(UFSAR). Therefore, since no initiating event mechanisms are being 
changed, the use of one of the Shield Building dome penetrations 
will not result in an increase in the probability of any previously 
evaluated accident.
    The use of one of the Shield Building dome penetrations affects 
the integrity of the Shield Building and the ability of the EGTS to 
maintain the annulus at a negative pressure relative to the outside 
atmosphere such that the function in mitigating the radiological 
consequences of an accident is affected. TVA's evaluation documents 
the radiological consequences of a LOCA assuming the open 
penetration is closed within fifteen minutes and the mission dose an 
individual may receive during ingress from the Auxiliary Building 
roof to the Shield Building dome, closure of the steel hatch 
assembly, and egress from the Shield Building dome. The LOCA 
radiological consequences with the penetration open for fifteen 
minutes are higher than those described in the UFSAR, however, the 
offsite and Control Room doses remain within the limits of 10 CFR 
[Title 10, Code of Federal Regulations] 100, Reactor Site Criteria, 
and 10 CFR 50, Appendix A, General Design Criteria (GDC) 19, Control 
Room, respectively. The calculated mission doses are also less than 
the limits of GDC 19. Therefore, since the increase in radiological 
consequences of the previously evaluated LOCA remains bounded by the 
applicable regulatory limits, the increased consequences are not 
considered significant.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Loss of Shield Building integrity or EGTS failure is not an 
initiator of any of the accidents and transients described in the 
UFSAR. A loss of Shield Building integrity during Modes 1-4 puts the 
plant into a Limiting Condition for Operation (LCO) situation and 
requires that the plant initiate shutdown within a specified 
timeframe if Shield Building integrity cannot be restored within the 
specified timeframe. The steel hatch assembly over each Shield 
Building dome penetration performs the same function as the concrete 
it replaces. Similar to a failure of the Shield Building, a failure 
of the steel hatch assembly will not initiate any of the accidents 
and transients described in the UFSAR. Postulated failures of the 
steel hatch assembly are degradation/damage to the seal or damage to 
the hatch hinges. Like any other Shield Building failure, these 
postulated steel hatch assembly failures result in a loss of Shield 
Building integrity and require that the failed component be repaired 
or replaced within a specified timeframe or that plant shutdown be 
initiated.
    Therefore, a failure of a steel hatch assembly during use of the 
Shield Building dome penetration will not initiate an accident nor 
create any new failure mechanisms. The changes do not result in any 
event previously deemed incredible being made credible. The use of 
the Shield Building dome penetration is not expected to result in 
more adverse conditions in the annulus and is not expected to result 
in any increase in the challenges to safety systems.
    Manual action is required to close an open Shield Building dome 
penetration and to configure the EGTS control loops following the 
opening and closing of a Shield Building dome penetration such that 
the EGTS will respond as designed. NRC Information Notice (IN) 97-
78, Crediting of Operator Actions in Place of Automatic Actions and 
Modifications of Operator Actions, Including Response Times, and 
ANSI/ANS [American Nuclear Standard Institute/American Nuclear 
Society]-58.8, Time Response Design Criteria for Safety-related 
Operator Actions, provide guidance for consideration of safety-
related operator actions.
    The manual actions implemented as a result of this change can be 
completed within the guidance and criteria provided in IN 97-78 and 
ANSI/ANS-58.8. Consequently, the manual actions can be credited in 
the mitigation of events that require Shield Building integrity. 
With credit for the manual actions to close an open Shield Building 
dome penetration and configure the EGTS control loops subsequent to 
an event, the types of accidents currently evaluated in the UFSAR 
remains the same.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The manual actions to close an open Shield Building dome 
penetration and to configure the EGTS control loops following the 
opening and closing of a Shield Building dome penetration ensure 
that the EGTS will respond as designed. Safety-related 
instrumentation is available to inform operators that a reactor trip 
has occurred, and dedicated trained individuals will be positioned 
to close an open Shield Building dome penetration, should an 
accident occur. The manual actions meet the criteria for safety-
related operator actions contained in NRC IN 97-78 and ANSI/ANS-
58.8. The use of manual actions maintains the margin of safety by 
assuring compliance with acceptance limits reviewed and approved by 
the NRC. The appropriate acceptance criteria for the various 
analyses and evaluation have been met; therefore, there has not been 
a reduction in any margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the

[[Page 41447]]

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 (PDR), located at One White Flint 
North, Public File Area 01F21, 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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of application for amendments: February 4, 2004.
    Brief description of amendments: The amendments revise Technical 
Specification 3.7.1, ``Main Steam Safety Valves (MSSVs),'' to permit 
operation in Mode 3 with five to eight inoperable MSSVs (two to five 
operable MSSVs) per steam generator, increase the Completion Time to 
reduce the variable overpower trip setpoint when one to four MSSVs per 
steam generator are inoperable, and make associated editorial changes.
    Date of issuance: July 7, 2005.
    Effective date: July 7, 2005, and shall be implemented within 90 
days of the date of issuance.
    Amendment Nos.: Unit 1-155, Unit 2-155, Unit 3 -155.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revise the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40671).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 7, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 20, 2004.
    Brief description of amendments: The amendments correct references 
in TS 5.6.7 and TS Table 3.3.10-1, and delete reference to hydrogen 
analyzers in TS 3.8.1, which were removed from the TSs by Amendment 
Nos. 262 and 239, for Unit Nos. 1 and 2, respectively, on March 2, 
2004.
    Date of issuance: July 5, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 274 and 251.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 2005 (70 FR 
400).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated July 5, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 17, 2005.
    Description of amendment request: The amendments replace the 
existing requirement of Technical Specification 3.4.5, ``RCS [Reactor 
Coolant System] Leakage Detection Instrumentation,'' Required Action 
D.1, to enter Limiting Condition for Operation (LCO) 3.0.3 if required 
leakage detection systems are inoperable with the requirement to be in 
Mode 3 within 12 hours and Mode 4 within 36 hours.
    Date of issuance: June 28, 2005.
    Effective date: June 28, 2005.
    Amendment Nos.: 237 and 265.
    Facility Operating License Nos. 50-325 and 50-324: Amendments 
revise the technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (70 FR 34161 dated June 13, 2005). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by August 12, 2005, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated June 28, 2005.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall.

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

    Date of application for amendment: October 15, 2004.
    Brief description of amendment: This amendment revises Technical 
Specifications by extending the inspection interval for reactor coolant 
pump flywheels to 20 years.
    Date of issuance: June 21, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 119.
    Facility Operating License No. NPF-63.: Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9988).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 21, 2005.
    No significant hazards consideration comments received: No.

[[Page 41448]]

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

    Date of application for amendments: September 8, 2004, as 
supplemented May 23, 2005.
    Brief description of amendments: These amendments delete the 
Technical Specifications associated with hydrogen recombiners and 
hydrogen monitors.
    Date of issuance: June 29, 2005.
    Effective date: As of the date of issuance and shall be implemented 
by December 31, 2005.
    Amendment Nos.: 287 and 224.
    Facility Operating License Nos. DPR-65 and NPF-49: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5238). The May 23, 2005 supplement provided clarifying information that 
did not change the scope of the proposed amendments as described in the 
original notice of proposed action published in the Federal Register, 
and did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 21, 2003, as supplemented on 
July 23, 2003, and March 31, 2005.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) to extend the surveillance test interval for the 
reactor protection system (RPS) intermediate range monitor (IRM) 
functional tests from weekly to 31 days. In addition, the amendment 
adds instrument check and calibration requirements for the RPS IRM--
High Flux function.
    Date of Issuance: July 7, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 225.
    Facility Operating License No. DPR-28: Amendment revised the TSs.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40713). The supplements contained clarifying information only, and did 
not change the initial no significant hazards consideration 
determination or expand the scope of the initial Federal Register 
notice.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated July 7, 2005.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: September 26, 2003, as 
supplemented December 8, 2004.
    Brief description of amendments: These amendments approve 
modifications to the Fire Protection Program. Specifically, the 
modifications involve converting the existing automatic carbon dioxide 
fire suppression systems installed in each of the four emergency diesel 
generator rooms and the cable spreading room to manual actuation.
    Date of issuance: June 24, 2005.
    Effective date: As of the date of issuance, to be implemented 
following completion of fire protection system modifications.
    Amendments Nos.: 255 and 258.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments approve modifications to the Fire Protection Program.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68669). The December 8, 2004, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 24, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 29, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain 
Valves,'' for the condition of having one or more SDV vent or drain 
lines with one valve inoperable.
    Date of issuance: June 23, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 259.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5247).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 23, 2005.
    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: December 19, 2003, as 
supplemented February 18, and March 17, 2004.
    Brief description of amendment: The amendment conforms the license 
to reflect the transfer of Operating License No. DPR-43 to Dominion 
Energy Kewaunee, Inc., as approved by order of the Commission dated 
June 10, 2004.
    Date of issuance: July 5, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 185.
    Facility Operating License No. DPR-43: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2734). The supplements dated February 18, and March 17, 2004, were 
within the scope of the initial application as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 2004.

 R. E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: December 20, 2004.
    Brief description of amendment: The amendment revises the sampling 
and testing requirements in Technical Specification 5.5.12, ``Diesel 
Fuel Oil Testing Program,'' which verify the acceptability of new 
diesel fuel oil for use, prior to addition to the storage tanks, and to 
stored fuel oil.
    Date of issuance: July 7, 2005.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 91.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19117).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 7, 2005.
    No significant hazards consideration comments received: No.

[[Page 41449]]

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

    Date of application for amendments: May 27, 2005, as supplemented 
by letters dated June 7, June 24, and July 1, 2005.
    Brief description of amendments: The amendments revise Technical 
Specification 3.3.7, ``DG-Undervoltage Start,'' by changing 
Surveillance Requirement 3.3.7.3.a to lower the allowable values for 
dropout and pickup of the degraded voltage function.
    Date of issuance: July 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 196 and 187
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2005 (70 FR 
34506). The supplemental letters dated June 7, June 24, and July 1, 
2005, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 1, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: May 17, 2005, as supplemented June 13, 
2005.
    Brief Description of amendments: The amendments revise the 
Technical Specification Section 3.7, ``Plant Systems,'' and Section 
4.0, ``Design Features,'' to establish cask storage area boron 
concentration limits and to restrict the minimum burnup of spent fuel 
assemblies associated with spent fuel cask loading operations.
    Date of issuance: June 29, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 169 and 161.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 2005 (70 FR 
30148). The supplement dated June 13, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 2005.
    No significant hazards consideration comments received: No. The NRC 
staff made a final determination that the amendment involves no 
significant hazards considerations.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: October 26, 2004
    Brief description of amendments: The amendments modify TS 
requirements to adopt the provisions of Industry/TS Task Force (TSTF) 
change TSTF-359, ``Increased Flexibility in Mode Restraints.''
    Date of issuance: June 24, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 137 and 116.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2898).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 24, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: July 8, 2004, as supplemented on 
April 15, 2005.
    Brief description of amendment: This amendment removes the 
requirement to maintain an automatic transfer capability for the power 
supply to the Low Pressure Coolant Injection inboard injection and 
recirculation pump discharge valves. The amendment also deletes 
references to Reactor Motor Operator Valve Boards D and E from the 
Technical Specifications.
    Date of issuance: June 20, 2005.
    Effective date: The amendment is effective as of the date of 
issuance.
    Amendment No.: 254.
    Facility Operating License No. DPR-33: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 2004 (69 FR 
64990). The April 15, 2005, letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 20, 2005.
    No significant hazards consideration comments received: No.

    Dated in Rockville, Maryland, this 11th day of July 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. E5-3793 Filed 7-18-05; 8:45 am]
BILLING CODE 7590-01-P