[Federal Register Volume 70, Number 137 (Tuesday, July 19, 2005)]
[Notices]
[Pages 41442-41449]
From the Federal Register Online via the Government Publishing 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, http://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, [email protected];
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
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by email to [email protected].
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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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