[Federal Register Volume 72, Number 68 (Tuesday, April 10, 2007)]
[Notices]
[Pages 17944-17959]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-6632]
-----------------------------------------------------------------------
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 March 16, 2007 to March 29, 2007. The last
biweekly notice was published on March 27, 2007 (72 FR 14303).
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, Rulemaking,
Directives and Editing 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 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
[[Page 17945]]
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 e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely 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 e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of amendment request: November 13, 2006.
Description of amendment request: The proposed amendment changes
the technical specification (TS) testing frequency for the surveillance
requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The proposed
change revises the test frequency of 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.''
AmerGen has reviewed the proposed no significant hazards
consideration determination published in the Federal Register on August
23, 2004 (69 FR 51864), as part of the consolidated line item
improvement process (CLIIP) and has concluded that the proposed
determination presented in the notice is applicable to Clinton Power
Station, Unit No. 1.
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.
Based on the above, the proposed change presents no significant
hazards consideration under the standards set forth in 10 CFR 50.92(c),
and accordingly, a finding of ``no significant hazards consideration''
is justified.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Russell Gibbs.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: January 26, 2007.
Description of amendment request: The proposed amendment would
revise technical specifications (TS) requirements for unavailable
barriers by
[[Page 17946]]
adding limiting condition for operation (LCO) 3.0.9. This would
establish conditions under which TS systems would remain operable when
required physical barriers are not capable of providing their related
support function. Also, the proposed amendment would make editorial
changes to LCO 3.0.8 to be consistent with the terminology in LCO
3.0.9.
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 by a reference to a generic analysis published in the
Federal Register on October 3, 2006 (71 FR 58444), which is presented
below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an unavailable barrier if risk is assessed and managed.
The postulated initiating events which may require a functional
barrier are limited to those with low frequencies of occurrence, and
the overall TS system safety function would still be available for
the majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on the
allowance provided by proposed LCO 3.0.9 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.9.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. The addition of a
requirement to assess and manage the risk introduced by this change
will further minimize possible concerns.
Therefore, this 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 involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to an unavailable barrier, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an unavailable
barrier, if risk is assessed and managed. The postulated initiating
events which may require a functional barrier are limited to those
with low frequencies of occurrence, and the overall TS system safety
function would still be available for the majority of anticipated
challenges. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in RG 1.177. A
bounding risk assessment was performed to justify the proposed TS
changes. This application of LCO 3.0.9 is predicated upon the
licensee's performance of a risk assessment and the management of
plant risk. The net change to the margin of safety is insignificant
as indicated by the anticipated low levels of associated risk (ICCDP
and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety
Evaluation.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: L. Raghavan.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: January 18, 2007.
Description of amendment request: The proposed amendment would
revise the expiration limit for the reactor coolant system Pressure/
Temperature (P/T) limit graphs in Technical Specifications (TS); revise
the adjusted reference temperature for the reactor vessel; and revise
the Low Temperature Overpressure Protection (LTOP) arming temperature
value specified in TSs. It would also make editorial changes in the use
of inequality signs in TSs associated with the LTOP arming temperature
in order to make them consistent.
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 does not affect the accident initiators or
mitigation assumptions associated with any of the accidents
previously evaluated. Operating restrictions on pressure-temperature
conditions for the reactor pressure vessel provide assurance that
reactor vessel integrity will be maintained under accident or
transient conditions. The proposed change uses approved criteria and
analysis methods to update the time period for which the current
operating limits remain valid.
The LTOP system performs an automatic function by opening relief
valves if reactor coolant system pressure reaches a temperature-
dependent limit. The proposed change includes establishing a more
restrictive temperature limit for when this system must be in
service, to reflect the material condition of the reactor vessel at
the new EFPY limit proposed for the pressure-temperature graphs. The
mitigation function and capability of the LTOP system is not being
changed by this request.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
There are no new accident initiators being introduced by this
proposed change. The proposed change does not involve installation
of new plant equipment, modification of existing equipment, or
changes in the way that plant equipment is operated. Pressure-
temperature operating limits depicted by graphs in the technical
specifications will not be changed and will continue to be used by
plant operators. A change in the LTOP system arming temperature will
assure that the graphs remain valid for the proposed new operating
period of 27.2 EFPY [effective full power years].
Therefore, the proposed changes do 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
Operating limits on pressure and temperature conditions for the
reactor coolant system (RCS) are important to assure that the RCS
pressure boundary stresses are within analyzed limits. Margins of
safety are inherent in the analysis methods, assumptions, and limits
specified in regulations and guidance documents. The proposed change
is based on NRC-accepted methods, assumptions and limits and
maintains the required margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 17947]]
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 Acting Branch Chief: Douglas V. Pickett.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), Westchester
County, New York
Date of amendment request: March 13, 2007.
Description of amendment request: The amendment would revise
License Condition 2.K for IP2 and License Condition 2.H for IP3, which
require the implementation and maintenance of an approved Fire
Protection Program for each unit.
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 changes are strictly an administrative relocation
of the specific fire protection SER [safety evaluation report]
references and do not modify any requirements of the fire protection
programs.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are strictly an administrative relocation
of the specific fire protection SER references and do not modify any
requirements of the fire protection programs.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are strictly an administrative relocation
of the specific fire protection SER references and do not modify any
requirements of the fire protection programs.
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 Acting Branch Chief: Douglas V. Pickett.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: March 22, 2007.
Description of amendment request: The proposed amendment will
revise the test acceptance criteria specified in Technical
Specification Surveillance Requirement (SR) 3.8.1.10 for the diesel
generator endurance test.
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?
No. The proposed change revises the acceptance criteria to be
applied to an existing surveillance test of the facility emergency
diesel generators (DGs). Performing a surveillance test is not an
accident initiator and does not increase the probability of an
accident occurring. The proposed new acceptance criteria will assure
that the DGs are capable of carrying the peak electrical loading
assumed in the various existing safety analyses which take credit
for the operation of the DGs. Establishing acceptance criteria that
bound existing analyses validates the related assumption used in
those analyses regarding the capability of equipment to mitigate
accident conditions. 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?
No. The proposed change revises the test acceptance criteria for
a specific performance test conducted on the existing DGs. The
proposed change does not involve installation of new equipment or
modification of existing equipment, so no new equipment failure
modes are introduced. The proposed revision to the DG surveillance
test acceptance criteria also is not a change to the way that the
equipment or facility is operated and no new accident initiators are
created. Therefore the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The conduct of performance tests on safety-related plant
equipment is a means of assuring that the equipment is capable of
maintaining the margin of safety established in the safety analyses
for the facility. The proposed change in the DG technical
specification surveillance test acceptance criteria is consistent
with values assumed in existing safety analyses and is consistent
with the design rating of the DGs. Therefore the proposed change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Acting Branch Chief: Douglas V. Pickett.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 15, 2007.
Description of amendment request: The proposed changes would revise
Technical Specification (TS) 3.10.1 to expand its scope to include
provisions for reactor coolant temperature excursions greater than 212
[deg]F as a consequence of inservice leak and hydrostatic testing, and
as a consequence of scram time testing initiated in conjunction with an
inservice leak or hydrostatic test, while considering operational
conditions to be in Mode 4, which is defined to be reactor coolant
temperature less than or equal to 212 [deg]F.
This change was proposed by the industry's TS Task Force (TSTF) and
is designated TSTF-484. The NRC staff issued a notice of opportunity
for comment in the Federal Register on August 21, 2006 (71 FR 48561),
on possible amendments concerning TSTF-484, including a model safety
evaluation and model no significant hazards (NSHC) determination, using
the consolidated line item improvement process (CLIIP). The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 27,
[[Page 17948]]
2006 (71 FR 63050). The licensee affirmed the applicability of the
following NSHC determination in its application dated February 15,
2007.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1: The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Technical Specifications currently allow for operation at
greater than 212 [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact the probability or consequences of an accident
previously evaluated. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2: The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Technical Specifications currently allow for operation at
greater than 212 [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. No new operational conditions beyond those currently allowed by
LCO 3.10.1 are introduced. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3: The proposed change does not involve a significant
reduction in a margin of safety.
Technical Specifications currently allow for operation at
greater than 212 [deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact any margin of safety. Allowing completion of
inspections and testing and supporting completion of scram time
testing initiated in conjunction with an inservice leak or
hydrostatic test prior to power operation results in enhanced safe
operations by eliminating unnecessary maneuvers to control reactor
temperature and pressure. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Based on the above, the NRC staff concludes that the proposed
change presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of
no significant hazards consideration is justified.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Acting Branch Chief: Douglas V. Pickett.
Exelon Generation Company, LLC (EGC), Docket Nos. 50-373 and 50-374,
LaSalle County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of amendment request: November 17, 2006.
Description of amendment request: The proposed amendments would
replace references to Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (Code) with a
reference to the ASME Code of Operation and Maintenance of Nuclear
Power Plants (OM Code) in Technical Specification (TS) 5.5.7,
``Inservice Testing Program [IST].'' These proposed changes are
consistent with the implementation of the LSCS, Units 1 and 2 third 10-
year IST program in accordance with the requirements of Title 10 of the
Code of Federal Regulations (10 CFR) Section 50.55a, ``Codes and
standards,'' paragraph (f), ``Inservice testing requirements.'' The
third 10-year interval for LSCS, Units 1 and 2 is scheduled to start on
October 12, 2007.
In addition to the replacement of the references, EGC is also
adding provisions in TS 5.5.7, item b, to only apply Surveillance
Requirement (SR) 3.0.2 to those inservice testing frequencies of two
years or less. These proposed changes are based on TS Task Force (TSTF)
Traveler No. 479-A (TSTF-479-A), Revision 0, ``Changes to Reflect
Revision of 10 CFR 50.55a,'' as modified by TSTF-497, Revision 0,
``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies
of 2 Years or Less'' and approved by the NRC in December 6, 2005, and
October 4, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to
conform to the requirements of 10 CFR 50.55a, ``Codes and
standards,'' paragraph (f) regarding the inservice testing of pumps
and calves for the Third 10-year Interval. The current TS reference
the [American Society of Mechanical Engineers] ASME Boiler and
Pressure Vessel Code, Section XI, requirements for the inservice
testing of ASME Code Class 1, 2, and 3 pumps and valves. The
proposed changes would reference the ASME OM Code, which is
consistent with 10 CFR 50.55a, paragraph (f), ``Inservice testing
requirements,'' and approved for use by the NRC. In addition,
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only
applied to those inservice testing frequencies of two years or less.
The definitions of the frequencies are not changed by this license
amendment request.
The proposed changes are administrative in nature, do not affect
any accident initiators, do not affect the ability of LSCS to
successfully respond to previously evaluated accidents and do not
affect radiological assumptions used in the evaluations. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to
conform to the requirements of 10 CFR 50.55a(f) regarding the
inservice testing of pumps and valves for the Third 10-year
Interval. The current TS reference the ASME Boiler and Pressure
Vessel Code, Section XI, requirements for the inservice testing of
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes
would reference the ASME OM Code, which is consistent with the 10
CFR 50.55a(f) and approved for use by the NRC. In addition,
provisions modifying TS 5.5.7, item b, clarify that SR 3.0.2 is only
applied to those inservice testing frequencies of two years or less.
The definitions of the frequencies are not changed by this license
amendment request.
The proposed changes to TS Section 5.5.7 do not affect the
performance of any LSCS structure, system, or component credited
with mitigating any accident previously evaluated and do not
introduce any new modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes revise TS 5.5.7 for LSCS Units 1 and 2 to
conform to the requirements of 10 CFR 50.55a(f) regarding the
inservice testing of pumps and valves for the Third 10-Year
Interval. The current TS reference the ASME Boiler and Pressure
Vessel Code, Section XI, requirements for the inservice testing of
ASME Code Class 1, 2,
[[Page 17949]]
and 3 pumps and valves. The proposed changes would reference the
ASME OM Code, which is consistent with the 10 CFR 50.55a(f) and
approved for use by the NRC. In addition, provisions modifying TS
5.5.7, item b, clarify that SR 3.0.2 is only applied to those
inservice testing frequencies of two years or less. The definitions
of the frequencies are not changed by this license amendment
request.
The proposed changes do not modify the safety limits setpoints
at which protective actions are initiated and do not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 2, 2006.
Description of amendment request: The proposed amendments
incorporates revised 10 CFR Part 20 requirements for Limerick
Generating Station Units 1 and 2 technical specifications (TSs).
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.
Updating the Technical Specifications (TS) to be consistent with
10 CFR Part 20 has no impact on plant structures, systems, or
components, does not affect any accident initiators, and does not
change any safety analysis. Therefore, the proposed changes do not
involve an 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.
Updating the TS to be consistent with 10 CFR Part 20 will not
change any equipment, require new equipment to be installed, or
change the way current equipment operates. No credible new failure
mechanisms, malfunctions, or accident initiators are created by the
proposed changes.
Therefore, the 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?
Response: No.
Updating the TS to be consistent with 10 CFR Part 20 does not
adversely affect existing plant safety margins or the reliability of
equipment assumed to operate in the safety analysis. As such, there
are no changes being made to safety analysis assumptions, safety
limits or limiting safety system settings that would adversely
affect plant safety as a result of the proposed changes. 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. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Harold K. Chernoff.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: February 12, 2007.
Description of amendment request: The proposed license amendment
would revise Technical Specification (TS) Limiting Condition for
Operation 3.9.4, ``Containment Penetrations'', to allow penetrations
included under TS 3.9.4(c) to be opened during core alterations or
movement of irradiated fuel, under administrative controls. This change
is based on the TS Task Force Traveler No. 312-A, Revision 1.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow containment penetrations
identified under Technical Specification 3.9.4(c) to remain open
during fuel movement and core alterations. These penetrations are
normally closed during this time period to prevent the release of
radioactive material in the event of a Fuel Handling Accident inside
containment. These penetrations are not initiators of any accident.
The probability of a Fuel Handling Accident is unaffected by the
status of these penetrations.
The Fuel Handling Accident analyses demonstrate that the maximum
offsite dose is well [within] the acceptance limits specified in SRP
[Standard Review Plan] 15.7.4, and the control room dose is within
the acceptance criteria specified in GDC [General Design Criterion]
19. Furthermore, the existing analysis results are independent of
the containment release path, and therefore are unaffected by the
proposed change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve the addition or
modification of any plant equipment. Also, the proposed change will
not alter the design, configuration, or method of operation of the
plant beyond the standard functional capabilities of the equipment.
The proposed change involves a Technical Specification change that
will allow containment penetrations identified under Technical
Specification 3.9.4(c) to remain open during fuel movement and core
alterations. Open penetrations are not accident initiators, and will
not create the possibility of a new kind of accident. Administrative
controls will be implemented to ensure the capability to close the
affected containment penetrations in the event of a Fuel Handling
Accident.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change has the potential to slightly increase the
post-Fuel Handling Accident dose at the site boundary and in the
control room. However, the existing analyses take no credit for
containment of the release, so that the existing analysis results
will remain bounding.
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Russell Gibbs.
[[Page 17950]]
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: January 19, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 5.5.9, ``Diesel Fuel Oil Testing
Program,'' by relocating a reference to a specific American Society for
Testing and Materials (ASTM) international standard for fuel oil
testing to licensee-controlled documents, and by adding an alternate
criteria to the ``clear and bright'' acceptance test for new fuel oil,
per the consolidated line item improvement process (CLIIP).
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on February 22, 2006
(71 FR 9179), on possible amendments concerning the CLIIP, including a
model safety evaluation and a model no significant hazards
consideration determination. The NRC staff subsequently issued a notice
of availability of the models for referencing in license amendment
applications in the Federal Register on April 21, 2006 (71 FR 20735),
as part of the CLIIP.
In its application dated January 19, 2007, the licensee affirmed
the applicability of the following determination.
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 proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Requirements to perform testing in
accordance with applicable ASTM standards are retained in the TS as
are requirements to perform surveillances of both new and stored
diesel fuel oil. Future changes to the licensee-controlled document
will be evaluated pursuant to the requirements of 10 CFR 50.59,
``Changes, tests and experiments,'' to ensure that such changes do
not result in more than a minimal increase in the probability or
consequences of an accident previously evaluated. In addition, the
``clear and bright'' test used to establish the acceptability of new
fuel oil for use prior to addition to storage tanks has been
expanded to recognize more rigorous testing of water and sediment
content. Relocating the specific ASTM standard references from the
TS to a licensee-controlled document and allowing a water and
sediment content test to be performed to establish the acceptability
of new fuel oil will not affect nor degrade the ability of the
emergency diesel generators (DGs) to perform their specified safety
function. Fuel oil quality will continue to meet ASTM requirements.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any 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 changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. In addition, the ``clear and bright''
test used to establish the acceptability of new fuel oil for use
prior to addition to storage tanks has been expanded to allow a
water and sediment content test to be performed to establish the
acceptability of new fuel oil. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. The requirements retained in the TS continue to require
testing of the diesel fuel oil to ensure the proper functioning of
the DGs.
Therefore, the 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?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Instituting the proposed changes will
continue to ensure the use of applicable ASTM standards to evaluate
the quality of both new and stored fuel oil designated for use in
the emergency DGs.
Changes to the licensee-controlled document are performed in
accordance with the provisions of 10 CFR 50.59. This approach
provides an effective level of regulatory control and ensures that
diesel fuel oil testing is conducted such that there is no
significant reduction in a margin of safety.
The ``clear and bright'' test used to establish the
acceptability of new fuel oil for use prior to addition to storage
tanks has been expanded to allow a water and sediment content test
to be performed to establish the acceptability of new fuel oil. The
margin of safety provided by the DGs is unaffected by the proposed
changes since there continue to be TS requirements to ensure fuel
oil is of the appropriate quality for emergency DG use. The proposed
changes provide the flexibility needed to improve fuel oil sampling
and analysis methodologies while maintaining sufficient controls to
preserve the current margins of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Russell A. Gibbs
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: December 14, 2006, as supplemented by
letter dated March 14, 2007.
Description of amendment request: The proposed amendment would
modify the technical specification (TS) requirements for inoperable
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The
changes are consistent with NRC approved Industry/Technical
Specification Task Force (TSTF) standard TS change TSTF-372, Revision
4.
The proposed amendment includes an administrative change to LCO
3.0.1 that will clarify that LCO 3.0.7 allows specified TS requirements
to be suspended during physics tests performed in accordance with TSs
3.1.8 and 3.1.9. This administrative change will make the CR-3 TSs more
consistent with the standard TSs and with TSTF-372, Revision 4.
The NRC staff issued a notice of availability of a model safety
evaluation and model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed
the applicability of the model NSHC determination in its application
dated April 26, 2006.
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:
[[Page 17951]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows a delay time before declaring
supported TS systems inoperable when the associated snubber(s)
cannot perform its required safety function. Entrance into Actions
or delaying entrance into Actions is not an initiator of any
accident previously evaluated. Consequently, the probability of an
accident previously evaluated is not significantly increased. The
consequences of an accident while relying on the delay time allowed
before declaring a TS supported system inoperable and taking its
Conditions and Required Actions are no different than the
consequences of an accident under the same plant conditions while
relying on the existing TS supported system Conditions and Required
Actions. Therefore, the consequences of an accident previously
evaluated are not significantly increased by this change. Therefore,
this 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 allows a delay time before declaring
supported TS systems inoperable when the associated snubber(s)
cannot perform its required safety function. The proposed change
does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows a delay time before declaring
supported TS systems inoperable when the associated snubber(s)
cannot perform its required safety function. The proposed change
restores an allowance in the pre-ISTS [improved Standard Technical
Specifications] conversion TS that was unintentionally eliminated by
the conversion. The pre-ISTS TSs were considered to provide an
adequate margin of safety for plant operation, as does the post-ISTS
conversion TS. Therefore, this 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 T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: March 16, 2007.
Description of amendment request: The proposed amendment would add
new Technical Specification (TS) requirements for the response times
associated with a steam generator feedwater pump (SGFP) trip and
feedwater isolation valve (FIV) closure. The amendment would also
revise the TS requirements for the containment fan cooler unit (CFCU)
cooling water flow rate. These changes are associated with a revised
containment response analysis that credits a SGFP trip and FIV closure
(on a feedwater regulator valve failure) to reduce the mass/energy
release to the containment during a main steam line break (MSLB). The
containment analysis also credits a reduced heat removal capability for
the CFCUs, allowing a reduction in the required service water (SW) flow
to the CFCUs.
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 does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change establishes response time requirements for
feedwater isolation and reduced CFCU flow rates to support
containment analyses to accommodate reduced CFCU heat removal
capacity. The changes in analysis input assumptions affect plant
response to an accident and are not accident initiators; therefore,
they have no bearing on the probability of an accident. The Salem
FSAR [Final Safety Analysis Report] Chapter 15 accidents which are
impacted by a change in the CFCU modeling parameters are LOCA [loss-
of-coolant accident] and MSLB mass and energy release Containment
analyses. The consequences of these postulated accidents are shown
to be acceptable using assumptions consistent with the proposed
changes.
For the LOCA transients, the containment cooling systems are
considered for three aspects: core response, containment response
and dose. The core response is most limiting when the containment
conditions minimize back pressure since this increases the blowdown
and reduces the effectiveness of the ECCS [emergency core cooling
system]. The LOCA core response (10 CFR 50.46 [Section 50.46 of
Title 10 of the Code of Federal Regulations]--PCT [peak cladding
temperature]) is conservatively biased to minimize the containment
backpressure such that any safety injection effectiveness is
minimized (the core becomes the highest resistance flow path). Thus,
any reduction in the accident capability of the CFCUs has no bearing
on the LOCA core response.
The bounding containment integrity analyses are the LBLOCA
[large-break LOCA] and the MSLB Inside Containment events. The
containment integrity analysis relies on two heat removal paths to
maintain containment pressure and temperature conditions. The CFCU
air-to-water heat exchangers reject containment energy to the SW
System and the Containment Spray System removes containment energy
by using spray droplet direct contact heat exchange to transfer the
energy from the containment ambient to the containment sump, where
it is transferred out of containment via the RHR [residual heat
removal] heat exchanger and CCW [component cooling water]/SW
Systems. Containment integrity analyses for both LOCA and MSLB,
using input assumptions consistent with the proposed changes, show
that containment integrity is maintained with reduced CFCU heat
removal capacity.
The potential dose impacts due to reduced CFCU heat removal
capacity are bounded as the design basis assumptions concerning the
number of operating CFCUs (three of five), and the thermal-hydraulic
transient operation of the Containment Spray System are unchanged.
The Salem design basis only credits Containment Spray iodine removal
effectiveness during the LOCA injection and recirculation phases
based on a single failure of an entire ESF [engineered safety
features] train. This assumption results in 3 of 5 CFCUs being
available to ensure adequate mixing of the containment ambient air
as well as operation of a single Containment Spray Train, which
controls containment spray droplet size and pH, as described in
UFSAR [updated FSAR] Section 6.2.3. As a further conservatism, the
current LOCA Alternate Source Term (AST) analysis (Calculation S-C-
ZZ-MDC-1945, an interim revision of which was sent to the NRC
[Nuclear Regulatory Commission] staff for review via letter dated
September 16, 2004) only credits two CFCUs for mixing. The
Containment Building and Auxiliary Building leakage rates are
unaffected by the revised containment analysis as the peak
containment pressure and temperatures are less than the design basis
values described in the Salem UFSAR. Therefore, there is no impact
on offsite dose rates due to the reduced CFCU heat removal capacity.
One other high energy line break for consideration is the
rupture of a feedwater line break. From a containment response
aspect, this event is bounded by the MSLB event, so it is not
explicitly analyzed (or even discussed in the Salem UFSAR).
A review of the Salem design basis for AST dose calculations
shows that the revised Containment Integrity Analysis, WCAP-16503,
does not challenge any of the assumptions that are part of the AST
design basis.
Section 6.2 of the UFSAR indicates that the Appendix J Type A
containment leak rate test
[[Page 17952]]
pressure is based on the containment design pressure of 47.0 psig,
not the calculated accident pressure. Since the design pressure
value bounds the peak pressure calculated in WCAP-16503 and is not
being changed, the Appendix J testing requirements are not impacted.
Thus, in conclusion, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The proposed change modifies response
time requirements for feedwater isolation, and reduces CFCU flow
rates and heat removal requirements consistent with the new
containment analysis.
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 create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes support revised containment analysis to
accommodate the reduced CFCU heat removal capacity.
The response time-related changes impose new surveillance
acceptance criteria to existing plant equipment that actuates to
isolate feedwater following a safety injection signal. There is no
change in actuation logic associated with the addition of response
time criteria; therefore no new accident sequences would result from
the imposition of response time test criteria to existing plant
equipment.
The reduction in minimum service water system flow to the CFCUs
is supported by analyses demonstrating acceptable system performance
and containment integrity following a demand for system operation.
The post-accident conditions resulting from the proposed reduction
in flow do not adversely impact the environmental qualification of
equipment, such that no new consequential failures are introduced to
any design basis accident scenario. CFCU operation with the proposed
reduction in minimum required accident flow would not result in the
progression of any design basis event into a previously unanalyzed
accident. Therefore, no new accident scenarios are created from the
CFCU flowrate reduction.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The proposed change does not involve a significant reduction in
the margin of safety. The revised containment analyses accommodate
reduced CFCU heat removal capacity using input assumptions
consistent with the proposed changes.
The proposed change involves the addition of feedwater isolation
response time surveillance criteria and reduction in minimum service
water system flows to CFCUs. These changes affect input to the
analyses of mass/energy releases and containment response to a
design basis main steam line break or loss of coolant accident. The
analyses, consistent with the proposed changes, demonstrate that the
acceptance criteria continue to be met, and the post-accident
conditions do not adversely affect containment integrity or
otherwise challenge any safety limit. The margin of safety with
respect to containment pressure is preserved by demonstrating that
the calculated pressures do not exceed the design limit of 47 psig.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 19, 2006.
Brief description of amendments: The amendments requested would
revise Technical Specifications (TS) requirement 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' TS 3.8.1, ``AC Sources--Operating,'' TS
3.8.9, ``Distribution Systems--Operating,'' and TS Example 1.3-3.
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. D[o] the proposed change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes eliminate certain Completion Times from the
Technical Specifications. Completion Times are not an initiator to
any accident previously evaluated. As a result, the probability of
an accident previously evaluated is not affected. The consequences
of an accident during the revised Completion Time are no different
than the consequences of the same accident during the existing
Completion Times. As a result, the consequences of an accident
previously evaluated are not affected by this change. The proposed
changes do not alter or prevent the ability of structures, systems,
and components from performing their intended function to mitigate
the consequences of an initiating event within the assumed
acceptance limits. The proposed changes do not affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed changes do not increase
the types or amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposures. The proposed changes are
consistent with the safety analysis assumptions and resultant
consequences.
Therefore, the proposed change[s] d[o] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. D[o] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. The changes
do not alter any assumptions made in the safety analysis.
Therefore, the proposed change[s] d[o] not create the
possibility of a new or different accident from any accident
previously evaluated.
3. D[o] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to delete the second Completion Time does
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed changes will not result in plant operation in a
configuration outside of the design basis.
Therefore, the proposed change[s] d[o] 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: January 18, 2007.
Brief description of amendments: The amendments requested would
revise Technical Specifications (TS) requirement 3.8.1, ``AC Sources--
Operating,'' Extension of Completion Times for Diesel Generators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 17953]]
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) changes do not
significantly increase the probability of occurrence of a previously
evaluated accident because the Diesel Generators (DGs) are not
initiators of previously evaluated accidents involving a loss of
offsite power (LOOP). The proposed changes to the TS Required
Actions and Completion Times (CT) do not affect any of the
assumptions used in the deterministic or the Probabilistic Safety
Assessment (PSA) analysis. Implementation of the proposed changes
does not result in a risk significant impact. The onsite AC
[alternating current] power sources will remain highly reliable and
the proposed changes will not result in a significant increase in
the risk of plant operation. This is demonstrated by showing that
the impact on plant safety as measured by the increase in core
damage frequency (CDF) is less than 1E-06 per year and the increase
in large early release frequency (LERF) is less than 1E-07 per year.
In addition, for the CT changes, the incremental conditional core
damage probabilities (ICCDP) and incremental conditional large early
release probabilities (ICLERP) are less than 5E-07 and 5E-08,
respectively. These changes meet the acceptance criteria in
Regulatory Guides 1.174 and 1.177. Therefore, since the onsite AC
power sources will continue to perform their functions with high
reliability as originally assumed and the increase in risk as
measured by [Delta]CDF, [Delta]LERF, ICCDP, and ICLERP risk metrics
is within the acceptance criteria of existing regulatory guidance,
there will not be a significant increase in the consequences of any
accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained.
The proposed changes do not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits. The proposed changes do
not affect the source term, containment isolation, or radiological
release assumptions used in evaluating the radiological consequences
of an accident previously evaluated. The proposed changes are
consistent with safety analysis assumptions and resultant
consequences.
The proposed TS changes will continue to ensure the DGs perform
their function when called upon. Extending the TS CT to 14 days,
when an AACPS [alternate AC power source] is available, does not
affect the design, the operational characteristics, the function, or
the reliability of the DGs. Additionally, the CT extension to 14
days does not affect the interfaces between the DGs and other plant
systems. Conversely, in the absence of an AACPS, the DG 72-hour CT
will be applied. The availability of the onsite AC power system to
perform its accident mitigation function is not affected by the
proposed activity and thus there is no impact to the radiological
consequences of any accident analysis.
To fully evaluate the effect of the changes to the CT, PSA
methods were utilized. The results of this analysis show no
significant increase in the CDF and LERF.
The Configuration Risk Management Program (CRMP) in TS 5.5.18 is
an administrative program that assesses risk based on plant status.
The risk-informed CT will be implemented consistent with the CRMP
and approved plant procedures. When utilizing the 14-day extension,
requirements of the CRMP per TS 5.5.18 call for the consideration of
other measures to mitigate the consequences of an accident occurring
while a DG is inoperable. Furthermore, administrative controls will
be applied when exercising the 14-day CT extension and are adequate
to maintain defense-in-depth and sufficient safety margins.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not result in a change in the manner in
which the electrical distribution subsystems provide plant
protection. The changes to the CT do not change any existing
accident scenarios, nor create any new or different accident
scenarios.
In addition, the changes do not impose any new or different
accident mitigation requirements or eliminate any existing
requirements.
The proposed changes are consistent with the safety analysis
assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. Neither the safety analyses nor the safety
analysis acceptance criteria are impacted by these changes. The
proposed changes will not result in plant operation in a
configuration outside the current design basis. The proposed
activities only involve changes to certain TS CTs.
Therefore the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: March 8, 2007.
Brief description of amendment request: The proposed amendments
would revise the McGuire Nuclear Station, Units 1 and 2, Technical
Specification 3.5.2.8, and the associated Bases and authorizes changes
to the Updated Final Safety Analysis Reports concerning modifications
to the emergency core cooling system sump.
Date of publication of individual notice in Federal Register: March
19, 2007.
Expiration date of individual notice: Comments April 18, 2007;
Hearing May 18, 2007.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination,
[[Page 17954]]
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 51.22(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].
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: May 23, 2006, as supplemented by
letters dated October 3, 2006, and October 24, 2006.
Brief description of amendment: This amendment revises Technical
Specification by modifying the steam generator tube surveillance
program at Shearon Harris Nuclear Power Plant, Unit 1.
Date of issuance: March 16, 2007.
Effective date: This amendment is effective as of the date of
issuance and shall be implemented within 90 days of issuance.
Amendment No. 124.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75991). The supplemental letters provided additional 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 the Safety Evaluation dated: March 16, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3 New London County, Connecticut
Date of application for amendment: July 19, 2006.
Brief description of amendment: The proposed amendment changed the
Millstone Power Station, Unit No. 3 (MPS3) reactor core safety limits
Technical Specification (TS) and relocated the reactor core safety
limit figure to the Core Operating Limits Report in the MPS3 Technical
Requirements Manual.
Date of issuance: March 14, 2007
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 236
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51227). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 14, 2007.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: April 11, 2006.
Brief description of amendments: (TSTF-372, Rev. 4) The amendments
added Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.0.8 to allow a delay time for entering a supported system TS
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed with an approved Bases Control Program that is
consistent with the TS Bases Control Program described in Section 5.5
of the applicable vendor's Standard Technical Specifications. The
amendment also made an administrative change, renumbering existing LCO
3.0.8 to LCO 3.0.9.
Date of issuance: March 19, 2007
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 235, 231
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70555). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: April 11, 2006.
Brief description of amendments: (TSTF-372, Rev. 4) The amendments
added Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.0.8 to allow a delay time for entering a supported system TS
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed with an approved Bases Control Program that is
consistent with the TS Bases Control Program described in Section 5.5
of the applicable vendor's Standard Technical Specifications. The
amendment also made an administrative change, renumbering existing LCO
3.0.8 to LCO 3.0.9.
Date of issuance: March 29, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 238, 220.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70556). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 29, 2007.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: May 22, 2006, as supplemented by
letter dated February 5, 2007.
Brief description of amendment: The amendment revised Technical
Specification Surveillance Requirements 3.8.1.11, 3.8.1.12, 3.8.1.16,
and 3.8.1.19 to eliminate the specific test-performance mode
[[Page 17955]]
restrictions for the High-Pressure Core Spray Division 3 diesel
generator.
Date of issuance: March 23, 2007.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment No.: 203.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40745). The supplemental letter dated February 5, 2007, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 23, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts.
Date of application for amendment: December 27, 2006.
Brief description of amendment: The amendment revised the Technical
Specification Limiting Condition for Operation 3.14.A to adopt the
Technical Specification Task Force 484, Revision 0, ``Use of Technical
Specification 3.10.1 for Scram Time Testing Activities.''
Date of issuance: March 26, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 226.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 20, 2007 (72
FR 7776). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 26, 2007.
No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts.
Date of application for amendment: January 15, 2007.
Brief description of amendment: The amendment revised the Technical
Specifications (TS) to extend the use of the current pressure-
temperature limits as specified in TS Figures 3.6.1, 3.6.2, and 3.6.3
through the end of operating cycle 18.
Date of issuance: March 26, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 227.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 12, 2007 (72
FR 6609). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 26, 2007.
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: April 22, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) requirements for inoperable snubbers by relocating
the current TS requirements Limiting Condition for Operation (LCO)
3.6.I and Surveillance Requirement (SR) 4.6.I to the Technical
Requirements Manual and adding LCO 3.0.8 to the TSs. The associated TS
Bases section has also been relocated.
Date of Issuance: March 26, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 230.
Facility Operating License No. DPR-28: The amendment revised the
License and TSs.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32604). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 26, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: January 18, 2007.
Brief description of amendment: The amendment revised the
description of the control rod assemblies in Grand Gulf Nuclear
Station, Unit 1, Technical Specification 4.2.2, ``Control Rod
Assemblies,'' to allow the use of hafnium as an additional type of
control material.
Date of issuance: March 16, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No: 174.
Facility Operating License No. NPF-29: The amendment revises the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6782). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 16, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 26, 2006.
Brief description of amendment: The amendment deleted reference to
the containment fan cooler condensate flow switch from Technical
Specification 3.4.5.1, ``Reactor Coolant System Leakage--Leakage
Detection Instrumentation,'' and modified or deleted associated
actions. The Nuclear Regulatory Commission staff had determined that
the remaining leak detection methods provided adequate means for
detecting, and to the extent practical, identifying the location of the
source of potential reactor coolant leakage.
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 212.
Facility Operating License No. NPF-38: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6782). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: May 26, 2006, as supplemented
on December 26, 2006, and March 14, 2007.
Brief description of amendments: The amendments revised the
existing steam
[[Page 17956]]
generator (SG) tube surveillance program. The changes are modeled after
Technical Specifications Task Force (TSTF) traveler TSTF-449, Revision
4, ``Steam Generator Tube Integrity,'' and the model safety evaluation
prepared by the Nuclear Regulatory Commission staff and published in
the Federal Register on March 2, 2005 (70 FR 10298). In this regard,
the scope of the amendments includes changes to the definition of
leakage, changes to the primary-to-secondary leakage requirements,
changes to the SG tube surveillance program (SG tube integrity), and
changes to the SG reporting requirements.
Date of issuance: March 14, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 298 and 279.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: July 5, 2006 (71 FR
38183).
The supplemental letters 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 amendment is contained
in a Safety Evaluation dated March 14, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: March 7, 2006, as supplemented
by letters dated May 30, September 7, December 15, 2006, and January 2,
2007.
Brief description of amendment: The amendment revised Section 4.3,
``Fuel Storage,'' of the Monticello Nuclear Generating Plant, technical
specifications to allow for installation of an additional temporary 8x8
(64-cell) high-density spent fuel storage rack in the spent fuel pool
to maintain full core off-load capability.
Date of issuance: March 9, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 150.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 3, 2006 (71 FR
16599).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: March 23, 2006, as supplemented
on December 19, 2006.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.3.4, ``Loss of Power (LOP) Diesel Generator (DG)
Start and Load Sequence Instrumentation,'' and surveillance requirement
3.3.4.3.b to modify the TS title and correct nonconservatisms in the
allowable values for the degraded voltage time delay.
Date of issuance: March 21, 2007.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 225 & 231.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23958).
The December 19, 2006, supplement, contained clarifying information
and did not change the staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: February 16, 2006, supplemented
by letters dated July 21, and December 27, 2006.
Brief description of amendments: The amendments consist of changes
to the Technical Specifications (TSs) related to steam generator tube
integrity. The amendments are modeled after the U.S. Nuclear Regulatory
Commission approved Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-449, ``Steam Generator
Tube Integrity,'' Revision 4 (ML0510902003).
Date of issuance: March 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 177 and 167.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 11, 2006 (71 FR
18376)
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 20, 2007.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: May 30, 2006, as supplemented
by letters dated November 22, 2006, and January 11, 2007.
Brief description of amendments: The amendments revised the
existing steam generator (SG) tube surveillance program. The changes
were modeled after Technical Specification Task Force (TSTF) traveler
TSTF-449, Revision 4, ``Steam Generator Tube Integrity,'' and the model
safety evaluation prepared by the U.S. Nuclear Regulatory Commission
and published in the Federal Register on March 2, 2005 (70 FR 10298).
The scope of the application included changes to the definition of
leakage, changes to the primary-to-secondary leakage requirements,
changes to the SG tube surveillance program (SG tube integrity),
changes to the SG reporting requirements, and associated changes to the
Technical Specification Bases.
Date of issuance: March 21, 2007.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--194; Unit 2--195.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40751). The supplemental letters dated November 22, 2006, and January
11, 2007, provided additional information that clarified the
application, did not expand the scope of the application as
[[Page 17957]]
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 March 21, 2007.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: December 14, 2006.
Brief description of amendments: The amendments deleted Section 2.G
of Facility Operating License Nos. DPR-80 and DPR-82, which require
reporting of violations of the requirements of Sections 2.C, 2.E, and
2.F of the operating license. This operating license improvement was
made available by the U.S. Nuclear Regulatory Commission on November 4,
2005, as part of the consolidated line item improvement process
(CLIIP).
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-193; Unit 2-194.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
154).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: November 18, 2005, as
supplemented on November 29, 2006, December 1, 2006, December 15, 2006,
January 9, 2007, and March 12, 2007 (PLA-6168 and PLA-6169).
Brief description of amendments: The amendments change the SSES 1
and 2 Technical Specifications (TSs) to implement the Average Power
Range Monitor/Rod Block Monitor/TSs/Maximum Load Line Limit Analysis by
revising TS 1.1, ``Definitions,'' TS 5.6.5, ``Core Operating Limits
Report,'' and the surveillance requirement sections of TS 3.3.1.1,
``Reactor Protection System Instrumentation,'' and TS 3.3.2.1,
``Control Rod Block Instrumentation.'' The amendments also delete TS
3.2.4, ``Average Power Range Monitor Gain and Setpoints,'' and its
associated references in the TSs. Additionally, the amendments change
the method of evaluation for the postulated recirculation line break in
the reactor pressure vessel shield annulus region.
Date of issuance: March 23, 2007.
Effective date: As of the date of issuance and to be implemented
prior to the startup following the SSES 1 spring 2008 15th refueling
outage for Unit 1 and prior to the startup following the SSES 2 spring
2007 13th refueling outage for Unit 2.
Amendment Nos.: 242 and 220.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7810).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 23, 2007.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: September 7, 2006.
Brief description of amendments: The amendments revise the SSES 1
and 2 Technical Specifications (TSs) Section 5.5.6, ``Inservice Testing
Program,'' and TS 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' to be consistent with the requirements of Title 10 of the
Code of Federal Regulations (10 CFR) Section 50.55a(f)(4) and 10 CFR
50.55a(g)(4), respectively. The amendments implement TS Task Force
(TSTF)-343, Revision 1 and TSTF-479, Revision 0.
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment Nos.: 241 and 219.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75997).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania
Date of application for amendment: November 16, 2006, as
supplemented on February 15, 2007.
Brief description of amendment: The amendment changes the SSES 2
Technical Specification (TS) Section 2.1.1.2 by revising the Unit 2
Cycle 14 Minimum Critical Power Ratio Safety Limit for two-loop and
single-loop operation and the references listed in TS 5.6.5.b.
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment No.: 218.
Facility Operating License No. NPF-22: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75998).
The supplement dated February 15, 2007, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: May 1, 2006.
Brief description of amendments: The amendments eliminate the
requirement for a power range neutron flux high negative rate trip and
delete the references to this trip in Salem Unit Nos. 1 and 2 Technical
Specification (TS) Table 2.2-1, ``Reactor Trip System Instrumentation
Trip Setpoints,'' TS Table 3.3-1, ``Reactor Trip System
Instrumentation,'' TS Table 3.3-2, ``Reactor Trip System
Instrumentation Response Times,'' and TS Table 4.3-1, ``Reactor Trip
System Instrumentation Surveillance Requirements.'' The amendments also
incorporate administrative and editorial changes to correct
miscellaneous errors in the TSs for Salem Units Nos. 1 and 2.
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
[[Page 17958]]
Amendment Nos.: 278 and 261
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40752).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 2007.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: August 4, 2006, as supplemented
by letter dated February 20, 2007.
Brief description of amendments: The amendments allow the use of
blind flanges for containment isolation in the containment purge system
supply and exhaust lines, and make corresponding changes to the
Technical Specifications (TSs). The amendments also consolidate the
containment isolation requirements by moving the requirements of TS 3/4
6.1.7, ``Containment Ventilation System,'' to TS 3/4 6.3.1 (TS 3/4 6.3
for Unit No. 2), ``Containment Isolation Valves.''
Date of issuance: March 19, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 277 and 260.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the License and the TSs.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit No. 1, Salem County, New Jersey
Date of application for amendment: January 18, 2007, as
supplemented on February 23, March 9, and March 22, 2007.
Brief description of amendment: The amendment approves a one-time
change to the Technical Specifications (TSs) regarding the steam
generator (SG) tube inspection and repair required for the portion of
the SG tubes passing through the tubesheet region. Specifically, for
Salem Unit No. 1 refueling outage 18 (planned for spring 2007) and the
subsequent operating cycle, the TS changes limit the required
inspection (and repair if degradation is found) to the portions of the
SG tubes passing through the upper 17 inches of the approximate 21-inch
tubesheet region.
Date of issuance: March 27, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 279.
Facility Operating License No. DPR-70: The amendment revised the
TSs and the License.
Date of initial notice in Federal Register: January 25, 2007 (72 FR
3427).
The letters dated February 23, March 9, and March 22, 2007,
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 amendment is contained
in a Safety Evaluation dated March 27, 2007.
No significant hazards consideration comments received: No.
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: March 28, 2006, as supplemented
by letter dated October 24, 2006.
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirement 3.5.1.4 to change the method and
frequency for verifying emergency core cooling system accumulator boric
acid concentration.
Date of issuance: March 28, 2007.
Effective date: As of the date of issuance to be implemented within
45 days.
Amendment No.: 101.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23960) The October 24, 2006, letter 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 amendment is contained
in a Safety Evaluation dated March 28, 2007.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: August 22, 2005, as supplemented by
letters dated September 18, 2006, October 23, 2006, and February 16,
2007.
Brief description of amendments: These amendments modified
Technical Specification (TS) requirements related to control room
envelope habitability in TS 3.7.10, ``Control Room Emergency
Filtration/Pressurization System (CREFS)'' and TS Section 5.5,
``Administrative Controls--Programs and Manuals.''
Date of issuance: March 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 136/136.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67754). The supplemental letters dated September 18 and October 23,
2006, and February 16, 2007, 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 March 26, 2007.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 31, 2006.
Brief description of amendments: The amendments revised Technical
Specification 5.0 entitled, ``ADMINISTRATIVE CONTROLS.'' Specifically,
the change deleted the Vice President, Nuclear Operations, as an
alternative to the Plant Manager for certain functions.
Date of Issuance: March 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1-134; Unit 2-134.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: September 12, 2006 (71
FR 53722).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 20, 2007.
No significant hazards consideration comments received: No.
[[Page 17959]]
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station (CPSES), Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: February 21, 2006, as supplemented by
letter dated March 19, 2007.
Brief description of amendments: The amendments revise TS 5.6.5
entitled, ``Core Operating Limits Report (COLR),'' by adding two
reports providing Loss-of-Coolant Accident (LOCA) and non-LOCA analysis
methodologies for CPSES Unit 1.
Date of issuance: March 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance, but no later than the entry
into Mode 5 in the restart of Unit 1 from its spring 2007 refueling
outage.
Amendment Nos.: 135/135.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32609). The supplemental letter dated March 19, 2007, 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 March 26, 2007.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 25, 2006, as supplemented by
letter dated March 12, 2007.
Brief description of amendment: The amendment revised Technical
Specifications 3.1.7, ``Rod Position Indication,'' 3.2.1, ``Heat Flux
Hot Channel Factor (FQ(Z)) (FQ Methodology),''
3.2.4, ``Quadrant Power Tilt Ratio (QPTR),'' and 3.3.1, ``Reactor Trip
System (RTS) Instrumentation,'' to allow use of the Westinghouse
proprietary computer code, the Best Estimate Analyzer for Core
Operations--Nuclear (BEACON). Certain required actions, for when a
limiting condition for operation is not met, and certain surveillance
requirements are being changed to refer to power distribution
measurements or measurement information of the core.
Date of issuance: March 21, 2007.
Effective date: As of its date of issuance and shall be implemented
before entry into Mode 2 in the plant restart from the refueling outage
scheduled for the spring of 2007. This includes the incorporation of
the identified changes to the Final Safety Analysis Report (FSAR) in
Attachment 6 of the licensee's application dated May 25, 2006, into the
FSAR.
Amendment No.: 182.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40756) The supplemental letter dated March 12, 2007, 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 published in the Federal Register on July 18, 2006.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 3rd day of April 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-6632 Filed 4-9-07; 8:45 am]
BILLING CODE 7590-01-P