[Federal Register Volume 72, Number 48 (Tuesday, March 13, 2007)]
[Notices]
[Pages 11383-11403]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-4251]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from February 15, 2007 through March 1, 2007.
The last biweekly notice was published on February 27, 2007 (72 FR
8800).
[[Page 11384]]
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 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
[[Page 11385]]
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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendment request: February 1, 2007.
Description of amendment request: The proposed license amendment
would revise Surveillance Requirement (SR) 3.5.2.8 in Technical
Specification 3.5.2, ``ECCS [Emergency Core Cooling System]--
Operating,'' to reflect the replacement of the containment
recirculation sump suction inlet trash racks and screens with
strainers, in response to Nuclear Regulatory Commission (NRC) Generic
Letter 2004-02, ``Potential Impact of Debris Blockage on Emergency
Recirculation during Design Basis Accidents at Pressurized-Water
Reactors.'' The proposed license amendment would replace ``trash racks
and screens'' with ``strainers'' in SR 3.5.2.8.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The consequences of accidents evaluated in the Updated Final
Safety Analysis Report [UFSAR] that could be affected by the
proposed change are those involving the pressurization of
Containment and associated flooding of the Containment and
recirculation of this fluid within the Emergency Core Cooling System
(ECCS) or the Containment Spray System (CSS) (e.g., loss-of-coolant
accidents [LOCAs]). The proposed change does not impact the
initiation or probability of occurrence of any accident. Although
the configurations of the existing containment recirculation sump
trash racks and screen and the replacement sump strainer cassettes
are different, they serve the same fundamental purpose of passively
removing debris from the sump's suction supply of the supported
system pumps. Removal of trash racks does not impact the adequacy of
the pump net positive suction head assumed in the safety analysis.
Likewise, the change does not reduce the reliability of any
supported systems or introduce any new system interactions. The
greatly increased surface area of the new strainer is designed to
reduce head loss and reduce the approach velocity at the strainer
face significantly, decreasing the risk of impact from large debris
entrained in the sump flow stream.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The containment recirculation sump strainers are a passive
system used for accident mitigation. As such, they cannot be
accident initiators. Therefore, there is no possibility that this
change could create any new or different kind of accident. No new
accident scenarios, transient precursors, or limiting single
failures are introduced as a result of the proposed change. There
will be no adverse effect or challenges imposed on any safety-
related system as a result of the change. Therefore, the possibility
of a new or different [kind] of accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be OPERABLE in the accident
analyses, as a result of the proposed Technical Specification
change. No new equipment performance burdens are imposed. The
possibility of a malfunction of safety-related equipment with a
different result is not 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?
The proposed change does not affect the acceptance criteria for
any analyzed event nor is there a change to any safety analysis
limit. There will be no effect on the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. The proposed change does not adversely affect the fuel,
fuel cladding, Reactor Coolant System, or containment integrity. The
radiological dose consequence acceptance criteria listed in the
Updated Final Safety Analysis Report will continue to be met.
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: Carey Fleming, Esquire, Senior Counsel--
Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt
Street, 17th floor, Baltimore, MD 21202.
NRC Acting Branch Chief: John P. Boska.
Carolina Power & Light Company, Docket Nos. 50-325 and 5-324 Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendments request: December 21, 2006.
Description of amendment request: The proposed amendment would
[[Page 11386]]
modify technical specification (TS) requirements of TS 3.4.1,
``Recirculation Loops Operating,'' to require the recirculation loops
be operated with matched flows versus recirculation pump speeds as
currently required. This change affects the Limiting Condition for
Operation (LCO) requirements and Surveillance Requirements (SRs) of TS
3.4.1.
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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment implements more conservative requirements
associated with recirculation loop operation. Specifically, the LCO
requirements of TS 3.4.1 and SR 3.4.1.1 are being revised to
directly monitor recirculation loop jet pump flows versus
recirculation pump speed, eliminating potential non-conservatism
associated with relating recirculation loop jet pump flow to
recirculation pump speed. These requirements assure that the
mismatch between recirculation loop jet pump flows are bounded by
the existing design bases analyses. As a result, the proposed change
ensures that the consequences of a design bases LOCA [loss-of-
coolant accident] remain within the existing evaluation.
The proposed change does not involve a physical change to the
Reactor Recirculation system, nor does it alter the assumptions of
the accident analyses. Therefore the probability of an accident
previously evaluated is not affected.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical change to the
Reactor Recirculation system, nor does it alter the assumptions of
the accident analyses.
The implementation of more conservative requirements associated
with recirculation loop operation does not introduce any new failure
modes. 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?
Response: No.
The proposed amendment implements more conservative requirements
associated with recirculation loop operation. These requirements
ensure that the Reactor Recirculation system is operated consistent
with the initial conditions of the existing design bases analyses.
Since the design bases analyses assumptions are unchanged, 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: 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: L. Raghavan.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: December 15, 2006.
Description of amendment request: The amendment would incorporate
changes to the Technical Specifications (TS) associated with previously
approved industry initiatives. The first change would relocate the
Safety Limit Violation specifications from the administrative controls
TS section to the safety limit TS sections as approved by TSTF-05-A,
``Deletion of Safety Limit Violation Requirements.'' The second change
would incorporate generic position titles, as approved by TSTF-65-A,
``Use of Generic Titles for Utility Positions,'' and incorporates
changes approved by NRC Administrative Letter (AL) 95-06, ``Relocation
of Technical Specification Administrative Controls Related to Quality
Assurance.''
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?
No. The proposed amendment consists of changes to and relocation
of administrative TS requirements that were previously generically
approved by the NRC. The proposed amendment would not change any of
the previously evaluated accidents in the updated safety analysis
report (USAR). The administrative controls that are affected by the
proposed amendment do not have any function related to preventing or
mitigating any of these previously evaluated accidents. The proposed
amendment does not affect any systems, structures, or components
(SSCs) that have the function of preventing or mitigating any of
these previously evaluated accidents. The proposed amendment does
not increase the likelihood of the malfunction of an SSC, thus the
potential impact on analyzed accidents need not be considered.
Because the proposed amendment is a relocation of administrative
requirements that are not associated with preventing or mitigating
the consequences of any previously evaluated accidents, there is no
affect on the probability or consequences of an accident previously
evaluated.
Therefore, the proposed amendment 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?
No. The proposed amendment consists of changes to and relocation
of administrative TS requirements previously generically approved by
the NRC. This amendment will not change the design function of any
SSC or the manner that any SSC is operated. Because this amendment
does not change the design function or operation of any SSC, the
amendment would not create the possibility of a new or different
kind of accident due to credible new failure mechanisms,
malfunctions, or accident initiators not considered in the design
and licensing bases.
Therefore, the proposed amendment 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?
No. The proposed amendment consists of changes to and relocation
of administrative TS requirements previously generically approved by
the NRC. The amendment does not alter any design basis safety limit
and no safety margins are affected.
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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Acting Branch Chief: P. Milano.
Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear
Station, Unit 1 (Catawba), York County, South Carolina
Date of amendment request: November 22, 2006.
Description of amendment request: The amendment would revise the
Catawba Unit 1 Facility Operating License (FOL) to provide for an
extension of the time limit to complete the required modification to
the Emergency Core Cooling System (ECCS) sump.
[[Page 11387]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed license amendment delineates a new Unit 1 FOL
condition to implement a completion date associated with the ECCS
sump strainer modification. The proposed license amendment is
administrative in nature and is being submitted to fulfill a
commitment made in previous Duke licensing correspondence.
Therefore, the proposed license amendment has no effect upon either
the probability or consequences of an accident previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
As stated above, the proposed license amendment is
administrative in nature and does not change the manner in which
Unit 1 is designed or operated. Therefore, the proposed license
amendment cannot create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant
system, and the containment. The performance of these barriers will
not be affected by the addition of the proposed FOL condition. Being
administrative in nature, the proposed license amendment therefore
does not involve a significant reduction in any safety margin.
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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
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: April 11, 2006.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) related to the organizational
description in TS 5.2.1
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided it's analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change revises an organizational description in TS
5.2.1 to reflect the change of the title of the Vice President Nuclear
Generation. The change is solely administrative in nature and has no
impact on any accident probabilities or consequences. The change does
not affect structures or components in the plant. The change has no
affect on any 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
There are no new accident causal mechanisms created as a result of
this proposed change. No changes are being made to the plant that will
introduce any new accident causal mechanisms. The change is solely
administrative in nature and does not impact any plant systems that are
accident initiators. Therefore, no new accidents or a different
accident than previously evaluated is being created.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
Margin of safety is related to confidence in the ability of the
fission product barriers to perform their design functions during and
following an accident situation. The proposed change is solely
administrative in nature and does not affect the performance of the
barriers. Consequently, no safety margins will be impacted. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied, therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN
50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.
1 and 2, Will County, Illinois.
Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units
2 and 3, Grundy County, Illinois.
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and
2, LaSalle County, Illinois.
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,
Units 2 and 3, York and Lancaster Counties, Pennsylvania.
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois.
Date of amendment request: December 15, 2006.
Description of amendment request: The proposed amendment would
modify the technical specifications (TSs) by replacing the term
``plant-specific titles'' with ``generic titles'' in TS Section
5.2.1.a, ensuring the TS description is consistent with the EGC Quality
Assurance Topical Report (QATR). The proposed amendment will also
revise the Peach Bottom TS Section 5.2.1.a, to replace the reference to
the Updated Final Safety Analysis Report with reference to the EGC
QATR. This will align the Peach Bottom TS wording with the rest of the
EGC fleet.
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 is a word replacement in TS 5.2.1, ``Onsite
and Offsite Organizations.'' The proposed change involves no changes
to plant systems or accident analyses. The proposed change is
administrative in nature and, as such, does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve any increase in
the probability or
[[Page 11388]]
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.
Creation of the possibility of a new or different kind of
accident would require creating one or more new accident precursors.
New accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The proposed change does not involve a physical alteration of the
plant, add any new equipment, or allow any existing equipment to be
operated in a manner different from the present method of operation.
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?
Response: No.
The proposed change is administrative in nature and has no
impact on equipment design or method of operation. There are no
changes being made to safety limits or safety system allowable
values that would adversely affect plant safety as a result of the
proposed change.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Michael L. Marshall, Jr.
Exelon Generation Company, LLC (EGC) Docket Nos. 50-254 and 50-265,
Quad Cities Nuclear Power Station, Unit 1, Rock Island County, Illinois
Date of amendment request: January 16, 2007.
Description of amendment request: The proposed amendment revises
the values of the safety limit minimum critical power ratio (SLMCPR) in
the Quad Cities Nuclear Power Station (QCNPS), Unit 1, Technical
Specification (TS) Section 2.1.1, ``Reactor Core SLs [Safety Limits].''
Specifically, the proposed change would require that for QCNPS, Unit 1,
the minimum critical power ratio shall be greater than 1.11 for two
recirculation loop operation, or greater than 1.13 for single
recirculation loop operation. This change is needed to support the next
cycle of operation for QCNPS, Unit 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the SLMCPR for QCNPS, Unit 1,
Cycle 20 such that the fuel is protected during normal operation and
during plant transients or anticipated operational occurrences
(AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during plant transients or AOOs.
Operational limits will be established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated. This will ensure that the
fuel design safety criterion (i.e., that at least 99.9% of the fuel
rods do not experience transition boiling during normal operation
and AOOs) is met. Since the proposed change does not affect
operability of plant systems designed to mitigate any consequences
of accidents, the consequences of an accident previously evaluated
are not expected to increase.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident requires creating one or more new accident precursors. New
accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The proposed change does not involve any plant configuration
modifications or changes to allowable modes of operation. The
proposed change to the SLMCPR assures that safety criteria are
maintained for QCNPS, Unit 1, Cycle 20.
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 SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the MCPR limit is not violated. The
proposed change will ensure the current level of fuel protection is
maintained by continuing to ensure that at least 99.9% of the fuel
rods do not experience transition boiling during normal operation
and AOOs if the MCPR limit is not violated. The proposed SLMCPR
values were developed using NRC-approved methods. Additionally,
operational limits will be established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated. This will ensure that the
fuel design safety criterion (i.e., that no more than 0.1% of the
rods are expected to be in boiling transition if the MCPR limit is
not violated) is met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment
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.
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, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Michael L. Marshall, Jr.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: December 29, 2006.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 3.1.8, ``Scram Discharge Volume (SDV) Vent
and Drain Valves,'' to allow a vent or drain line with one inoperable
valve to be isolated instead of requiring the valve to be restored to
operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments concerning the consolidated line item implement process
(CLIIP), including a model safety evaluation and a model no significant
hazards consideration
[[Page 11389]]
(NSHC) 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 15, 2003 (68 FR 18294),
as part of the CLIIP. In its application dated December 29, 2006, 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead or requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDVs is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDVs is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of an SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin 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: Michael L. Marshall, Jr.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio, and Docket Nos. 50-334
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver
County, Pennsylvania
Date of amendment request: January 11, 2007.
Description of amendment request: The proposed license amendments
would modify technical specification (TS) requirements for inoperable
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8. The
proposed license amendments also modify LCO 3.0.1 to incorporate the
addition of LCO 3.0.8. This change is based on the TS Task Force (TSTF)
Traveler, TSTF-372, Revision 4. A notice of availability for this TS
improvement using the consolidated line item improvement process was
published in the Federal Register on May 4, 2005.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model no significant hazards consideration (NSHC)
determination for referencing license amendment applications in the
Federal Register on November 24, 2004 (69 FR 68412), and May 4, 2005
(70 FR 23252). The licensee affirmed the applicability of the model
NSHC determination in its application dated January 11, 2007.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast 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 allowance provided by proposed LCO 3.0.8
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 3.0.8. 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. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
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 inoperable snubbers, 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. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The proposed change allows 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. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast 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 3.0.8 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. 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 W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Michael L. Marshall, Jr.
[[Page 11390]]
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: December 14, 2006.
Description of amendment request: The proposed license amendment
would revise the accident source term used in the NMP1 design basis
radiological consequence analyses in accordance with 10 CFR 50.67. The
revised accident source term replaces the current methodology that is
based on TID-14844, ``Calculation of Distance Factors for Power and
Test Reactor Sites,'' with the alternative source term (AST)
methodology described in Regulatory Guide (RG) 1.183, ``Alternative
Source Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors.'' The amendment request is for full implementation of the AST
as described in RG 1.183, with the exception that TID-14844 will
continue to be used as the radiation dose basis for equipment
qualification and vital area access.
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.
Adoption of the AST and those plant systems affected by
implementing AST do not initiate DBAs [design-basis accidents]. The
AST does not affect the design or manner in which the facility is
operated; rather, for postulated accidents, the AST is an input to
calculations that evaluate the radiological consequences. The AST
does not by itself affect the post-accident plant response or the
actual pathway of the radiation released from the fuel. It does,
however, better represent the physical characteristics of the
release, so that appropriate mitigation techniques may be applied.
Implementation of the AST has been incorporated in the analyses for
the limiting DBAs at NMP1.
The structures, systems and components affected by the proposed
change mitigate the consequences of accidents after the accident has
been initiated. Application of the AST does result in changes to
NMP1 Updated Final Safety Analysis Report (UFSAR) functions (e.g.,
Liquid Poison system). As a condition of the application of AST,
NMPNS is proposing to use the Liquid Poison system to control the
suppression pool pH following a LOCA [loss-of-coolant accident]. The
proposed changes also revise operability requirements for the
secondary containment and certain post-accident filtration systems
while handling irradiated fuel that has decayed for greater than 24
hours and during core alterations. These changes have been included
within the AST evaluations. These changes do not require any
physical changes to the plant. As a result, the proposed changes do
not involve a revision to the parameters or conditions that could
contribute to the initiation of a DBA discussed in Chapter XV of the
NMP1 UFSAR. Since design basis accident initiators are not being
altered by adoption of the AST, the probability of an accident
previously evaluated is not affected.
Plant-specific AST radiological analyses have been performed
and, based on the results of these analyses, it has been
demonstrated that the dose consequences of the limiting events
considered in the analyses are within the acceptance criteria
provided by the NRC for use with the AST. These criteria are
presented in 10 CFR 50.67 and Regulatory Guide 1.183. Even though
the AST dose limits are not directly comparable to the previously
specified whole body and thyroid dose guidelines of General Design
Criterion 19 and 10 CFR 100.11, the results of the AST analyses have
demonstrated that the 10 CFR 50.67 limits are satisfied. Therefore,
it is concluded that adoption of the AST does not involve a
significant increase in the consequences of an accident previously
evaluated.
Based on the above discussion, it is concluded that 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.
Implementation of AST and the proposed changes do not alter or
involve any design basis accident initiators. These changes do not
involve any physical changes to the plant and do not affect the
design function or mode of operations of systems, structures, or
components in the facility prior to a postulated accident. Since
systems, structures, and components are operated essentially no
differently after the AST implementation, no new failure modes are
created by this proposed change.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes proposed are associated with a new licensing basis
for analysis of NMP1 DBAs. Approval of the licensing basis change
from the original source term to the AST is being requested. The
results of the accident analyses performed in support of the
proposed changes are subject to revised acceptance criteria. The
limiting DBAs have been analyzed using conservative methodologies,
in accordance with the guidance contained in Regulatory Guide 1.183,
to ensure that analyzed events are bounding and that safety margin
has not been reduced. The dose consequences of these limiting events
are within the acceptance criteria presented in 10 CFR 50.67 and
Regulatory Guide 1.183. Thus, the proposed changes continue to
ensure that the doses at the exclusion area boundary and low
population zone boundary, as well as in the control room, are within
corresponding regulatory criteria.
Therefore, by meeting the applicable regulatory criteria for
AST, it is concluded that the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Acting Branch Chief: John P. Boska.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: January 4, 2007.
Description of amendment request: The proposed license amendment
would revise Technical Specification (TS) 3.7.1, ``Service Water (SW)
System and Ultimate Heat Sink (UHS),'' as follows: Revise the existing
Limiting Condition for Operation (LCO) statement to require four
operable SW pumps to be in operation when SW subsystem supply header
water temperature is <=82 [deg]F; add a requirement that five operable
SW pumps be in operation when SW subsystem supply header water
temperature is >82 [deg]F and <=84 [deg]F; delete Condition G and the
associated Required Actions and Completion Times; revise Surveillance
Requirement 3.7.1.3 to increase the maximum allowed SW subsystem supply
header water temperature from 82 [deg]F to 84 [deg]F; and modify the
requirements for increasing the surveillance frequency as the
temperature approaches the limit.
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.
[[Page 11391]]
The proposed change eliminates the requirement to perform
temperature averaging when the UHS temperature is >82 [deg]F,
establishes 84 [deg]F as the design limit for UHS water temperature
for operation on a continuous basis, and revises the frequency for
verifying that the UHS temperature is within the prescribed limit.
The TS currently allow operation with the UHS water temperature
temporarily exceeding 82 [deg]F, up to a maximum of 84 [deg]F. The
UHS temperature itself is not an initiator of accidents analyzed in
the Updated Safety Analysis Report (USAR). Raising the maximum
temperature limit and revising the associated surveillance
requirement frequency do not involve any plant hardware changes or
new operator actions that could serve to initiate an accident.
Continuous operation with the elevated UHS temperature may result in
a few balance-of-plant equipment high temperature alarms. Operator
response to these alarms would be in accordance with established
alarm response procedures. In all cases, trip setpoints leading to a
reactor scram or a power runback will not be reached, and the
likelihood of component failures that could initiate an accident
will not be significantly increased.
The potential impact of the proposed change on the ability of
the plant to mitigate postulated accidents has been evaluated. These
evaluations demonstrate that safety-related systems and components
that rely on the UHS as the cooling medium or as a pump suction
source are capable of performing their intended safety functions at
the higher UHS temperature, and that containment integrity and
equipment qualification are maintained. The calculated post-accident
dose consequences reflected in the USAR do not directly utilize UHS
temperature as an input and thus are not impacted by the proposed
change.
Based on the above, the proposed change will have no adverse
effect on plant operation or the availability or operation of any
accident mitigation equipment. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not alter the current plant
configuration (no new or different type of equipment will be
installed) or require any new or unusual operator actions. The
proposed change will not alter the way any structure, system, or
component functions and will not cause an adverse effect on plant
operation or accident mitigation equipment. The response of the
plant and the operators following a design-basis accident is
unaffected by the change. The proposed change does not introduce any
credible new failure mechanisms, malfunctions, or accident
initiators not considered in the design and licensing bases.
Analyses have shown that the design basis heat removal capability of
the affected safety-related components is maintained at the
increased UHS water temperature limit.
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?
Response: No.
The margin of safety is determined by the design and
qualification of the plant equipment, the operation of the plant
within analyzed limits, and the point at which protective or
mitigative actions are initiated. The proposed change does not
impact these factors. An evaluation of the safety systems has been
performed to ensure their safety functions can be met for operation
with a UHS water temperature of 84 [deg]F on a continuous basis.
Operation with the UHS water temperature temporarily exceeding 82
[deg]F, up to a maximum of 84 [deg]F, is currently allowed.
Operating on a continuous basis at the higher UHS temperature
represents a slight reduction in design margins in terms of the
ability of affected systems to remove accident heat loads. However,
the evaluation has demonstrated that the proposed change does not
have a significant impact on the capability of the affected systems
to perform their safety-related post-accident functions and to
mitigate accident consequences. The design limits for the
containment and fuel cladding will not be exceeded, and equipment
qualification will be maintained. No protection setpoints are
affected by the proposed change. The revised frequency for
performing the TS surveillance to verify that the UHS temperature is
within the prescribed limit will continue to assure that plant
operators are aware of and are monitoring increasing UHS temperature
trends prior to reaching a value of 82 [deg]F, when a fifth SW pump
must be placed in operation. This action is no different than that
required by the current TS.
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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Acting Branch Chief: John P. Boska.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: January 29, 2007.
Description of amendment request: The proposed amendment would
revise Table 3.3.5.1-1, ``Emergency Core cooling System
Instrumentation,'' of the Technical Specifications (TSs) to extend the
quarterly surveillance interval from quarterly to a nominal 24-month
interval for three low pressure coolant injection loop select logic
functions. Consistent with the extended test interval, the licensee
also proposed to change the allowable values associated with each of
the three logic functions (i.e., response time in seconds). The
licensee stated that the quarterly surveillance requirement was
inappropriately introduced when the TSs was converted from its previous
custom format to the current Improved Technical Specification format by
Amendment No. 146. Before the conversion, there was no such quarterly
surveillance requirement. Furthermore, the plant was not designed to
have these three logic functions tested while on-line.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC).
The NRC staff reviewed the licensee's analysis, and has performed its
own analysis as follows:
(1) Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed amendment would extend the performance interval
from quarterly to a 24-month interval, and change the associated
allowable values for the three logic functions. The performance of
these surveillances, or the failure to perform, as well as the
surveillance finding (i.e., response time in seconds) are not
precursors to, and do not affect the probability of, an accident. There
is no design or operation change associated with the proposed
amendment. Therefore, the proposed amendment does not increase the
probability of an accident previously evaluated.
A delay in performing these surveillances would not result in a
system being unable to perform its required function. The extended
surveillance and associated changed allowable values will not affect
the three logic functions to operate as designed. Therefore, the plant
systems required to mitigate accidents will remain capable of
performing their design function. As a result, the proposed amendment
will not lead to any significant change in the consequences of any
accident.
(2) Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
[[Page 11392]]
No. The proposed amendment does not involve a physical alteration
of any system, structure, or component (SSC) or a change in the way any
SSC is operated. The proposed amendment does not involve operation of
any SSCs in a manner or configuration different from those previously
recognized or evaluated. No new failure mechanisms will be introduced
by the extended surveillance interval and associated allowable values.
Thus, the proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction in
a margin of safety?
No. The proposed amendment only changes the surveillance interval
and associated allowable values for the three logic functions. There
will be no modification of any TSs limiting condition for operation, no
change to any limit on previously analyzed accidents, no change to how
previously analyzed accidents or transients would be mitigated, no
change in any methodology used to evaluate consequences of accidents,
and no change in any operating procedure or process. The
instrumentation and components involved in this proposed amendment have
exhibited reliable operation based on the results of their performance
during past periodic emergency core cooling system functional testing.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
the NRC staff's own analysis above, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
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 amendment request: January 29, 2007.
Description of amendment request: The proposed amendments would
revise technical specification (TS) 3.5.3, ``ECCS (Emergency Core
Cooling Systems)--Shutdown'' operability requirements for the Safety
Injection (SI) subsystem. These revisions will allow the required SI
pump to be rendered incapable of injecting into the Reactor Coolant
System (RCS) during low temperature (MODE 4) operations due to a single
action or automatic signal. The capability of the plant operators to
initiate SI flow on a timely basis will be maintained.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to add a new Note to
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be
considered operable within the Limiting Condition for Operation
requirements while the system is not capable of automatic injection
provided it is capable of being manually aligned for injection.
This Emergency Core Cooling System is not an accident initiator,
thus the proposed changes do not increase the probability of an
accident. The current licensing basis, Technical Specifications and
Bases do not require automatic initiation instrumentation for the
Emergency Core Cooling System in Mode 4, but rather assume operator
action to mitigate an accident. With the proposed Technical
Specification and Bases changes, the Emergency Core Cooling System
will continue to be operable for manual initiation. Since the
changes proposed in this license amendment request do not impact the
performance of the system, these changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
The changes proposed in this license amendment 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.
This license amendment request proposes to add a new Note to
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be
considered operable within the Limiting Condition for Operation
requirements while the system is not capable of automatic injection
provided it is capable of being manually aligned for injection.
The changes proposed for the Emergency Core Cooling System
Technical Specifications do not change any system operations,
maintenance activities or testing requirements. The Limiting
Condition for Operation will continue to be met, no new failure
modes or mechanisms are created and no new accident precursors are
generated by this change. The Technical Specification changes
proposed in this license amendment do 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.
This license amendment request proposes to add a new Note to
Technical Specification 3.5.3, ``Emergency Core Cooling System--
Shutdown''. This Note will allow the Safety Injection system to be
considered operable within the Limiting Condition for Operation
requirements while the system is not capable of automatic injection
provided it is capable of being manually aligned for injection.
The current licensing basis, Technical Specifications and Bases
rely upon operator actions to initiate safety injection to mitigate
an accident in Mode 4 and do not require operability of any process
instrumentation capable of automatically initiating the Emergency
Core Cooling System. With the changes proposed in this license
amendment request, the safety injection system will continue to be
operable and the plant will continue to rely on operator actions for
safety injection initiation. Thus, the Technical Specification
changes proposed in this license amendment request 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 requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: P. Milano.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: October 11, 2006, as supplemented on
October 25, November 21, and December 4, 2006.
Description of amendment request: The proposed amendments would
increase the SSES 1 and 2 licensed thermal power to 3952 Mega-watts
thermal (MWt), which is 20% above the original rated thermal power
(RTP) of 3293 MWt, and approximately 13% above the current RTP of 3489
MWt. The proposed amendments would revise the SSES 1 and 2 Operating
License and Technical Specifications necessary to implement the
increased power level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 11393]]
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?
Extended Power Uprate
Response: No.
The probability (frequency of occurrence) of Design Basis
Accidents occurring is not affected by the increased power level,
because Susquehanna continues to comply with the regulatory and
design basis criteria established for plant equipment. A
probabilistic risk assessment demonstrates that the calculated core
damage frequencies do not significantly change due to Constant
Pressure Power Uprate (CPPU). Scram setpoints (equipment settings
that initiate automatic plant shutdowns) are established such that
there is no significant increase in scram frequency due to CPPU. No
new challenges to safety-related equipment result from CPPU.
The changes in consequences of postulated accidents, which would
occur from 102% of the CPPU (rated thermal power) RTP compared to
those previously evaluated, are acceptable. The results of CPPU
accident evaluations do not exceed the NRC-approved acceptance
limits. The spectrum of postulated accidents and transients has been
investigated, and are shown to meet the plant's currently licensed
regulatory criteria. In the area of fuel and core design, for
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and
other applicable Specified Acceptable Fuel Design Limits (SAFDLS)
are still met. Continued compliance with the SLMCPR and other SAFDLs
will be confirmed on a cycle specific basis consistent with the
criteria accepted by the NRC.
Challenges to the Reactor Coolant Pressure Boundary were
evaluated at CPPU conditions (pressure, temperature, flow, and
radiation) were found to meet their acceptance criteria for
allowable stresses and overpressure margin.
Challenges to the containment have been evaluated, and the
containment and its associated cooling systems continue to meet 10
CFR [Part] 50, Appendix A, Criterion 16, Containment Design;
Criterion 38, Containment Heat Removal; and Criterion 50,
Containment Design Basis. The increase in the calculated post LOCA
[loss-of-coolant accident] suppression pool temperature above the
currently assumed peak temperature was evaluated and determined to
be acceptable.
Radiological release events (accidents) have been evaluated, and
shown to meet the guidelines of 10 CFR 50.67.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
LPRM [Local Power Range Monitor] Calibration Interval Technical
Specification SR [Surveillance Requirement] Frequency Change
Response: No.
The revised surveillance interval continues to ensure that the
LPRM signal is adequately calibrated. This change will not alter the
basic operation of process variables, structures, systems, or
components as described in the SSES FSAR [final safety analysis
report], and no new equipment is introduced by the change in LPRM
surveillance interval. The performance of the APRM [average power
range monitor] and RBM [rod block monitor] systems is not
significantly affected by the proposed LPRM surveillance interval
increase. Therefore, the probability of accidents previously
evaluated is unchanged.
The proposed change results in no change in radiological
consequences of the design basis LOCA as currently analyzed for
SSES. The consequences of an accident can be affected by the thermal
limits existing at the time of the postulated accident, but LPRM
chamber exposure has no significant effect on the calculated thermal
limits because LPRM accuracy does not significantly deviate with
exposure. For the extended calibration interval, the assumption in
the safety limit analysis remains valid, maintaining the accuracy of
the thermal limit calculation. Therefore, the thermal limit
calculation is not significantly affected by LPRM calibration
frequency and the consequences of an accident previously evaluated
are unchanged.
The change does not affect the initiation of any event, nor does
it negatively impact the mitigation of any event. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
RHR [Residual Heat Removal] Service Water System and Ultimate Heat
Sink Technical Specification and Methods Change
Response: No.
The proposed changes do not involve any new initiators for any
accidents nor do they increase the likelihood of a malfunction of
any Structures, Systems or Components (SSCs). Implementation of the
subject changes reduces the probability of adverse consequences of
accidents previously evaluated, because inclusion of the manual
spray array bypass isolation valves and the small spray array
isolation valves in the Technical Specifications (TS) increases
their reliability to function for safe shutdown. The use of the ANS/
ANSI-5.1-1979 decay heat model in the UHS [ultimate heat sink]
performance analysis is not relevant to accident initiation, but
rather, pertains to the method used to evaluate currently postulated
accidents. Its use does not, in any way, alter existing fission
product boundaries, and provides a conservative prediction of decay
heat. Therefore, the change in decay heat calculational method does
not involve a significant increase in 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.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, and the ANS/ANSI-5.1-1979 decay
heat model are not relevant to accident initiation, but rather,
pertain to the method used to evaluate postulated accidents. The use
of these elements does not, in any way, alter existing fission
product boundaries, and provides a conservative prediction of the
containment response to DBA [design-basis accident]-LOCAs.
Therefore[,] the Containment Analysis Method Change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Feedwater Pump/Condensate Pump Trip Change
Response: No.
Feedwater pump trips and condensate pump trips rarely occur. A
low water level SCRAM on loss of one feedwater pump or one
condensate pump is bounded by the loss of all feedwater transient in
Final Safety Analysis Report (FSAR) Appendix 15E. A trip of one
feedwater pump or a trip of one condensate pump does not result in
the loss of all feedwater. The Feedwater Pump / Condensate Pump Trip
Change is included in the CPPU Probabilistic Risk Assessment (PRA).
The best estimate for the Susquehanna Steam Electric Station (SSES)
Core Damage Frequency (CDF) risk increase due to the CPPU is 6E-08
for Unit 1 and 7E-08 for Unit 2 which are in the lower left corner
of Region III of Regulatory Guide [sic] (Reference 15) (i.e., very
small risk changes). The best estimate for the Large Early Release
Frequency (LERF) increase is 1.0E-09/yr for both units which is also
in the lower left corner of the Region III range of Regulatory Guide
1.174. Therefore, the Feedwater Pump/Condensate Pump Trip Change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Main Turbine Pressure Regulation System
Response: No.
Technical Specification 3.7.8 does not directly or indirectly
affect any plant system, equipment, component, or change the process
used to operate the plant. Technical Specification 3.7.8 would
ensure acceptable performance, since it would establish requirements
for adhering to the appropriate thermal limits, depending on the
operability of the main turbine pressure regulation system. Use of
the appropriate limits assures that the appropriate safety limits
will not be exceeded during normal or anticipated operational
occurrences. Thus, Technical Specification 3.7.8 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?
Extended Power Uprate
Response: No.
Equipment that could be affected by EPU has been evaluated. No
new operating mode, safety-related equipment lineup, accident
scenario, or equipment failure mode was identified. The full
spectrum of accident
[[Page 11394]]
considerations has been evaluated and no new or different kind of
accident has been identified. CPPU uses developed technology and
applies it within capabilities of existing or modified plant safety
related equipment in accordance with the regulatory criteria
(including NRC approved codes, standards and methods). No new
accidents or event precursors have been identified.
The SSES TS require revision to implement EPU. The revisions
have been assessed and it was determined that the proposed change
will not introduce a different accident than that previously
evaluated. Therefore[,] the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
LPRM Calibration Interval Technical Specification SR Frequency
Change
Response: No.
The proposed change will not physically alter the plant or its
mode of operation. The performance of the APRM and RBM systems is
not significantly affected by the proposed LPRM surveillance
interval increase. As such, no new or different types of equipment
will be installed and the basic operation of installed equipment is
unchanged. The methods of governing plant operation and testing are
consistent with current safety analysis assumptions. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
RHR Service Water System and Ultimate Heat Sink Technical
Specification and Methods Change
Response: No.
The subject changes apply Technical Specification controls to
new UHS manual bypass isolation valves and the existing small spray
array isolation valves. The design functions of the systems are not
affected.
The addition of manually operated valves in the system,
operational changes and the Technical Specification changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
The use of the ANS/ANSI-5.1-1979 decay heat model is not
relevant to accident initiation, but rather pertains to the method
used to evaluate currently postulated accidents. The use of this
analytical tool does not involve any physical changes to plant
structures or systems, and does not create a new initiating event
for the spectrum of events currently postulated in the FSAR.
Further, it does not result in the need to postulate any new
accident scenarios. Therefore[,] the decay heat calculational method
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated[.]
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay
heat model are not relevant to accident initiation, but pertain to
the method used to evaluate currently postulated accidents. The use
of these analytical tools does not involve any physical changes to
plant structures or systems, and does not create a new initiating
event for the spectrum of events currently postulated in the FSAR.
Further, they do not result in the need to postulate any new
accident scenarios. Therefore, the Containment Analysis Method
Change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Feedwater Pump/Condensate Pump Trip Change
Response: No.
The occurrence of a reactor SCRAM is already considered in the
current licensing basis and is not an accident. A SCRAM resulting
from the trip of a feedwater pump or a condensate pump is bounded by
a loss of all feedwater event. The loss of all feedwater transient
is already considered in the plant licensing basis. The SCRAM due to
the feedwater or condensate pump trip does not change the results of
the loss of all feedwater transient in any way. Therefore, the
Feedwater Pump/Condensate Pump Trip Change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Main Turbine Pressure Regulation System
Response: No.
Technical Specification 3.7.8 will not directly or indirectly
affect any plant system, equipment, or component and therefore does
not affect the failure modes of any of these items. Thus, Technical
Specification 3.7.8 does not create the possibility of a previously
unevaluated operator error or a new single failure.
Therefore, Technical Specification 3.7.8 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?
Extended Power Uprate
Response: No.
The CPPU affects only design and operational margins. Challenges
to the fuel, reactor coolant pressure boundary, and containment were
evaluated for CPPU conditions. Fuel integrity is maintained by
meeting existing design and regulatory limits. The calculated loads
on affected structures, systems and components, including the
reactor coolant pressure boundary, will remain within their design
allowables for design basis event categories. No NRC acceptance
criterion is exceeded. Because the SSES configuration and responses
to transients and postulated accidents do not result in exceeding
the presently approved NRC acceptance limits, the proposed changes
do not involve a significant reduction in a margin of safety.
LPRM Calibration Interval Technical Specification Change
Response: No.
The proposed change has no impact on equipment design or
fundamental operation and there are no changes being made to safety
limits or safety system allowable values that would adversely affect
plant safety as a result of the proposed change. The performance of
the APRM and RBM systems is not significantly affected by the
proposed LPRM surveillance interval increase. The margin of safety
can be affected by the thermal limits existing prior to an accident;
however, uncertainties associated with LPRM chamber exposure have no
significant effect on the calculated thermal limits. For the
extended calibration interval, the assumption in the safety limit
analysis remains valid, maintaining the accuracy of the thermal
limit calculation.
Since the proposed change does not affect safety analysis
assumptions or initial conditions, the margin of safety in the
safety analyses are maintained. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
RHR Service Water System and Ultimate Heat Sink Technical
Specification and Methods Change
Response: No.
Implementation of the subject changes does not significantly
reduce the margin of safety since these changes add components and
Technical Specification controls for the components not currently
addressed in the Technical Specifications. These changes increase
the reliability of the affected components/systems to function for
safe shutdown.
Therefore[,] these changes do not involve a significant
reduction in margin of safety.
The ANS/ANSI-5.1-1979 model provides a conservative prediction
of decay heat. The use of this element is consistent with current
industry standards, and has been previously accepted by the staff
for use in containment analysis by other licensees, as described in
GE Nuclear Energy. ``Constant Pressure Power Uprate,'' Licensing
Topical Report NEDC-33004P-A, Revision 4, dated July 2003; and the
letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of
the SHEX Computer Program and ANSI/ANS 5.1-1979, ``Decay Heat Source
Term for Containment Long-Term Pressure and Temperature Analysis,''
July 13, 1993. Therefore, the decay heat calculational method change
does not involve a significant reduction in the margin of safety.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay
heat model are realistic phenomena, and provide a conservative
prediction of the plant response to DBA-LOCAs. The use of these
elements is consistent with current industry standards, and has been
previously accepted by the staff for other licensees, as described
in GE Nuclear Energy: ``Constant Pressure Power Uprate,'' Licensing
Topical Report NEDC-33004P-A, Revision 4, dated July 2003; the
letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of
the SHEX Computer Program; and ANSI/ANS 5.1-1979, ``Decay Heat
Source Term for Containment Long-Term Pressure and Temperature
Analysis,'' July 13, 1993. Therefore the Containment Analysis Method
Change does not involve a significant reduction in [a] margin of
safety.
[[Page 11395]]
Feedwater Pump/Condensate Pump Trip Change
Response: No.
A low water level SCRAM on loss of one feedwater pump or one
condensate pump is bounded by the loss of all feedwater transient in
FSAR Appendix 15E. The loss of all feedwater transient is a non-
limiting event that does not contribute to the setting of the fuel
safety limits. Consequently, a SCRAM resulting from a feedwater pump
or condensate pump trip does not reduce the margin to fuel safety
limits. Therefore, the potential for a SCRAM resulting from a
feedwater pump trip or a condensate pump trip does not involve a
significant reduction in [a] margin of safety.
Main Turbine Pressure Regulation System
Since Technical Specification 3.7.8 does not alter any plant
system, equipment, component, or processes used to operate the
plant, the proposed change will not jeopardize or degrade the
function or operation of any plant system or component governed by
Technical Specifications. Technical Specification 3.7.8 preserves
the margin of safety by establishing requirements for adhering to
the appropriate thermal limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC (Acting) Branch Chief: Douglas V. Pickett.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: February 2, 2007.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) LCO 3.10.1 to expand its scope to
include provisions for temperature excursions greater than 212 degrees
F as a consequence of scram time testing initiated in conjunction with
an inservice leak or hydrostatic test. During these tests and with
temperature greater than 212 degrees F, operational conditions are
considered to be in Mode 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 October 27, 2006 (71 FR 63050). The licensee
affirmed the applicability of the model NSHC determination in its
application dated February 2, 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.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of amendment request: January 12, 2007.
Description of amendment request: The proposed amendment would
revise the steam generator (SG) program requirements in the Sequoyah
(SQN) Unit 2 Technical Specifications (TSs) to allow use of an SG
voltage-based repair criteria probability of detection (POD) method
using plant-specific SG tube inspection results. The proposed POD
method is referred to as the probability of prior cycle detection
(POPCD) method.
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 use of a revised SG voltage-based repair
criteria POD method, the POPCD method, to determine the BOC
[beginning of cycle] indication voltage distribution for the SQN
Unit 2 operational assessments does not increase the probability of
an accident. Based on industry and plant-specific bobbin detection
data for ODSCC [outside diameter stress corrosion cracking] within
the SG tube support plate (TSP) region, large voltage bobbin
indications which individually can challenge structural or leakage
integrity can be detected with near 100 percent certainty. Since
large voltage outside diameter stress corrosion cracking ODSCC
bobbin indications within the SG TSP can be detected, they will not
be left in service, and therefore these indications should not be
included in the voltage distribution for the purpose of operational
assessments. The POPCD method improves the estimate of potentially
undetected indications for operational assessments, but does not
directly affect the inspection results. Since large voltage
indications are detected, they will not result in an increase in the
probability of SG tube rupture accident or an increase in the
consequences of a tube rupture or main steam line break (MSLB)
accident.
[[Page 11396]]
2. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The use of the POPCD method is associated with numerical
predictions of probabilities for the steam generator tube rupture
(SGTR) accident. Since the SGTR accident is considered in SQN's
Updated Final Safety Analysis Report, there is no possibility to
create a design basis accident that has not been previously
evaluated. Therefore, the proposed change does not create the
possibility of a new or different accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The use of the POPCD method to determine the BOC
voltage distribution for the SQN Unit 2 operational assessments does
not involve a significant reduction in a margin of safety. The
applicable margin of safety potentially impacted is the SG tube
structural and leakage criteria. Based on industry and plant-
specific bobbin detection data for ODSCC within the SG TSP region,
large voltage bobbin indications that can individually challenge
structural or leakage integrity can be detected with near 100
percent certainty and will not be left in service. Therefore, these
indications should not be included in the voltage distribution for
the purposes of operational assessments. Since these large voltage
indications are detected, they will not result in a significant
increase in the actual EOC [end of cycle] leakage for a MSLB
accident or the actual EOC probability of burst. The POPCD method
approach to POD considers the potential for missing indications that
might challenge structural or leakage integrity by applying the
POPCD data from successive inspections. If a large indication was
missed in one inspection, it would continue to grow until detected
in a later inspection. Accordingly, there is no significant increase
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Brenda Mozafari (Acting).
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.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: February 2, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.6.1.7, ``Suppression Chamber-to-
Drywell Vacuum Breakers,'' to allow a one-time extension to the current
closure verification surveillance requirement for one of two redundant
disks in one of nine vacuum breakers until reliable position indication
can be restored in the main control room during the next refueling
outage (R-18), which is scheduled to begin on May 12, 2007.
Date of publication of individual notice in Federal Register:
February 12, 2007 (72 FR 6606).
Expiration date of individual notice: February 26, 2007.
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 proposed amendment would revise
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 publication of individual notice in Federal Register:
February 20, 2007 (72 FR 7776).
Expiration date of individual notice: March 22, 2007 (public
comments) and April 23, 2007 (hearing requests).
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 request supercedes
the previously submitted license amendment request dated April 12,
2006, proposing new Pressure-Temperature (PT) curves and to extend the
applicability of current PT limits expressed in Technical Specification
Figures 3.6.1, 3.6.2, and 3.6.3 through the end of operating cycle 18.
Date of publication of individual notice in Federal Register:
February 12, 2007 (72 FR 6609).
Expiration date of individual notice: March 14, 2007 (public
comments) and April 13, 2007 (hearing requests).
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit No. 1, Salem County, New Jersey
Date of amendment request: January 18, 2007.
Brief description of amendment request: The amendment request
proposes 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 proposed TS
changes would 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 publication of individual notice in Federal Register:
January 25, 2007 (72 FR 3427).
Expiration date of individual notice: February 26, 2007 (public
comments) and March 26, 2007 (hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance
[[Page 11397]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: September 28, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) requirements for mode change limitations in Limiting
Condition for Operation (LCO) 3.0.4 and Surveillance Requirement 3.0.4
to adopt the provisions of Industry/TS Task Force (TSTF) Traveler
number TSTF-359, ``Increase Flexibility in Mode Restraints.'' The
amendments also revised TS Example 1.4-1 to reflect the changes made to
LCO 3.0.4 and to be consistent with TSTF-485, which has been
incorporated into the Standard Technical Specifications Revision 3.1.
Date of issuance: February 21, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1-165, Unit 2--165, Unit 3--165.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65140). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 21, 2007.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: February 14, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) requirements in the Limiting Condition for Operation
for TS 3.6.3, ``Containment Isolation Valves,'' and associated Actions
and Surveillance Requirements to allow for a blind flange to be used
for containment isolation in each of the two flow paths of the 42-inch
refueling purge valves in Modes 1 through 4, without remaining in TS
3.6.3 Condition D.
Date of issuance: February 22, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-166, Unit 2-166, Unit 3-166.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13171).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 2007.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: April 26, 2006.
Brief Description of amendments: Revised the Technical
Specification (TS) requirements for inoperable snubbers by adding
Limiting Condition for Operation 3.0.8.
Date of issuance: February 15, 2007.
Effective date: February 15, 2007, implement within 90 days.
Amendment Nos.: 241 and 269.
Renewed Facility Operating License Nos. DPR-71 and DPR-62:
Amendments change the TSs.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32603).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 15, 2007.
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, as supplemented
November 29, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) related to steam generator tube
integrity. The changes are consistent with the consolidated line-item
improvement process, Nuclear Regulatory Commission's approved Technical
Specification Task Force (TSTF) Standard Specification Change Traveler,
TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
Date of issuance: March 1, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 237, 218.
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
70557) The supplement dated November 29, 2006, 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 1, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO-2), Pope County, Arkansas
Date of application for amendment: March 20, 2006.
Brief description of amendment: The amendment removed ANO-2 reactor
coolant structural integrity requirements contained in TS 3.4.10.1. The
TS change is consistent with NUREG-1432, ``Standard Technical
Specifications Combustion Engineering Plants,'' Revision 3.1. The Bases
for TS 3.4.10.1 will be deleted and performed under the ANO-2 TS Bases
Control Program, and is not included with the submittal. The amendment
also renumbers TS pages 3/4 4-22a, 23, 23a, and 23b as TS pages 3/4 4-
23, 24, 25, and 26, respectively.
Date of issuance: March 1, 2007.
[[Page 11398]]
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 270.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
26999). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 1, 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: November 1, 2006.
Brief description of amendment: The amendment modified technical
specification requirements for inoperable snubbers by adding Limiting
Condition of Operation 3.0.8 using the Consolidated Line Item
Improvement Process.
Date of issuance: February 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 171.
Facility Operating License No. NPF-29: The amendment revises the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70558). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 20, 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: November 13, 2006.
Brief description of amendment: The amendment revised Grand Gulf
Nuclear Station, Unit 1, Technical Specification (TS) Limiting
Condition of Operation 3.10.1, and the associated TS Bases, to expand
its scope to include provisions for temperature excursions greater than
200 [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.
Date of issuance: February 21, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 172.
Facility Operating License No. NPF-29: The amendment revises the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75993). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 21, 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: May 8, 2006, as supplemented by
letter dated November 16, 2006.
Brief description of amendment: The change added an NRC-approved
topical report to the analytical methods referenced in Technical
Specification Section 5.6.5, ``Core Operating Limits Report (COLR).''
Date of issuance: February 22, 2007.
Effective date: As of the date of issuance and shall be implemented
prior to Cycle 16 operation.
Amendment No: 173.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 20, 2006 (71 FR
35458). The supplement dated November 16, 2006, 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 February 22, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of application for amendment: May 31, 2006, as supplemented by
letter dated August 30, 2006.
Brief description of amendment: The amendments revise the Technical
Specifications (TSs) associated with steam generator tube integrity
consistent with Revision 4 to the TS Task Force (TSTF) Standard
Technical Specification Change Document TSTF-449, ``Steam Generator
Tube Integrity.''
Date of issuance: February 20, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 251 and 233.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the TSs.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43531). The August 30, 2006, supplement 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 as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 20, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: April 4, 2006.
Brief description of amendments: The amendments add one NRC-
approved topical report reference to the list of analytical methods in
Technical Specification (TS) Section 5.6.5, ``Core Operating Limits
Report (COLR),'' that can be used to determine core operating limits
and delete seven obsolete references from the same TS section.
Date of issuance: February 15, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 181/168.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46933). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 15, 2007.
No significant hazards consideration comments received: No.
[[Page 11399]]
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: February 25, 2005, as
supplemented by letters dated November 11, 2005, April 19, July 10,
2006, September 1, October 24, December 7, 2006, and February 1, 2007.
Brief description of amendments: The amendment converts the current
Technical Specifications to the Improved Technical Specifications
(ITSs) format and relocates certain requirements to other licensee-
controlled documents. The ITSs are based on NUREG-1431, ``Standard
Technical Specifications--Westinghouse Plants,'' Revision 2, with the
Technical Specification Task Force changes to make the Beaver Valley
Power Station Unit Nos. 1 and 2 (BVPS-1 and 2) ITS more consistent with
Revision 3; the Commission's Final Policy Statement, ``NRC Final Policy
Statement on Technical Specification Improvements for Nuclear Power
Reactors,'' dated July 22, 1993 (58 FR 39132); and 10 CFR 50.36,
``Technical specifications.'' The purpose of the conversion is to
provide clearer and more readily understandable requirements in the TSs
for BVPS-1 and 2 to ensure safe operation. In addition, the amendment
includes a number of issues that were considered beyond the scope of
NUREG-1431.
Date of issuance: February 1, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 150 days.
Amendment Nos.: 278 and 161.
Facility Operating License Nos. DPR-66 and NPF-73: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: March 22, 2006 (71 FR
14554). The letters dated November 11, 2005, April 19, July 10, 2006,
September 1, October 24, December 7, 2006, and February 1, 2007,
supplement 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 as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 1, 2007.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: April 28, 2006.
Description of amendment request: The amendment revised the
Seabrook Technical Specifications (TSs) Limiting Condition for
Operation 3.0.4 and Surveillance Requirement (SR) 4.0.4 to adopt the
provisions of Industry/TS Task Force (TSTF) change TSTF-359, Revision
9, ``Increased Flexibility in Mode Restraints.'' TSTF-359 is part of
the consolidated line item improvement process. Specifically, the
proposed change allows, for systems and components, mode changes into a
TS condition that has a specific required action and completion time.
Date of issuance: February 9, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 114.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: July 5, 2006 (71 FR
38182). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 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: October 23, 2006.
Brief description of amendments: The amendments to the Technical
Specifications (TSs) eliminate the use of the defined term CORE
ALTERATIONS in the TSs.
Date of issuance: February 15, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 224 & 230.
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications/License.
Date of initial notice in Federal Register: December 5, 2006 (71 FR
70562). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 15, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: February 13, 2006.
Brief description of amendments: The amendments revise Prairie
Island Nuclear Generating Plant, Units 1 and 2, Technical
Specifications (TS) to change the wording in TS 3.0, ``Surveillance
Requirement (SR) Applicability'' and change format and titles in TS
5.0, ``Administrative Controls.'' The proposed changes improve the TS
usability, conformance with the industry standard, NUREG-1431,
``Standard Technical Specifications, Westinghouse Plants,'' Revision
3.0 and accuracy.
Date of issuance: February 13, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 176 and 166.
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
18375). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 13, 2007.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 13, 2006.
Brief description of amendment: The amendment relocated the
requirements of Technical Specification (TS) 2.22, ``Toxic Gas
Monitors,'' and TS Table 3-3, Item 29, to the Fort Calhoun Station,
Unit No. 1, Updated Safety Analysis Report.
Date of issuance: February 28, 2007.
Effective date: As of its date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 248.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75996). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated February 28, 2007.
No significant hazards consideration comments received: No.
[[Page 11400]]
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: April 28, 2006.
Brief description of amendments: The amendments revise the SSES 1
and 2 Technical Specifications 3.1.7, ``Standby Liquid Control (SLC)
System,'' to modify the SLC system for single loop pump operation and
the use of enriched sodium pentaborate solution.
Date of issuance: February 28, 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 and SSES 2 Spring 2007 13th refueling outage for Units 1 and 2,
respectively.
Amendment Nos.: 240 and 217.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and license.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46936). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 28, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259 Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of application for amendment: May 1, 2006 (TS-455), as
supplemented by letters dated September 1, and November 6, 2006.
Brief description of amendment: The amendment revises the numeric
values of the safety limit critical power ratio (SLMCPR) in the
Technical Specification (TS) Section 2.1.1.2 for one and two reactor
recirculation loop operation to incorporate the results of the Cycle 7
SLMCPR analysis.
Date of issuance: February 6, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 267.
Facility Operating License Nos. DPR-33: Amendment revised the TSs.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46937). The supplements dated September 1, and November 6, 2006,
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 February 6, 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 and October 23, 2006.
Brief description of amendments: The amendments revised the Final
Safety Evaluation Report Sections 1, 6, and 15. The changes reflect the
licensee's adoption of Nuclear Regulatory Commission's Regulatory Guide
1.195, ``Methods and Assumptions for Evaluating Radiological
Consequences of Design Basis Accidents at Light-Water Reactors,'' for
calculating radiological consequences and replacement of steam
generators for Comanche Peak Steam Electric Station, Unit 1, in the
spring of 2007.
Date of issuance: February 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 130/130.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Final Safety Analysis Report and Facility Operating
Licenses.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67754). The supplements dated September 18 and October 23, 2006,
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 February 20, 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: February 21, 2006, as supplemented by
letters dated September 12 and December 14, 2006.
Brief description of amendments: The amendments increased the
allowable values (AVs) for steam generator (SG) water level trip
setpoints and the required minimum SG secondary side water inventory in
shutdown modes for the replacement SGs in Comanche Peak Steam Electric
Station (CPSES), Unit 1. For CPSES Unit 2, the corresponding AVs and
the SG secondary water inventory in the current TSs remain unchanged
since the existing SGs in Unit 2 will continue to be used.
Date of issuance: February 20, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: NPF-87--131; NPF-89--131.
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 supplements dated September 12 and December 14, 2006,
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 February 20, 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: December 16, 2005, as supplemented by
letters dated August 31 and September 29, 2006.
Brief description of amendments: The amendments revised Technical
Specifications (TSs) 1.1 and 5.6.6 consistent with the Nuclear
Regulatory Commission (NRC)-approved Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-419,
``Revise PTLR [Pressure Temperature Limits Report] Definition and
References in ISTS [Improved Standard Technical Specification] 5.6.6.
Date of issuance: February 22, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: NPF-87-132 and NPF-89-132.
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: March 14, 2006 (71 FR
[[Page 11401]]
13182). The supplements dated August 31 and September 29, 2006,
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 February 22, 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: December 12, 2005.
Brief description of amendments: The amendments revise the
Technical Specification (TS) Surveillance Requirements (SRs) 3.3.1.2
and 3.3.1.3, ``Reactor Trip System (RTS) Instrumentation.'' The license
amendment request is based on Technical Specification Task Force (TSTF)
Traveler, TSTF-371-A, Revision 1, ``NIS [Nuclear Instrumentation
System] Power Range Channel Daily SR TS Change to Address Low Power
Decalibration.'' TSTF-371-A, Revision 1, revised the requirements for
performing a daily surveillance adjustment of the power range
channel(s) to address industry concern that compliance with SR 3.3.1.2
and SR 3.3.1.3 may result in a non-conservative channel calibration
during reduced-power operations. The changes resolved the issue of non-
conservatism.
Date of issuance: February 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: NPF-87-133, NPF-89-133.
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: March 28, 2006 (71 FR
15490).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 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 30, 2006, as supplemented by
letters dated November 22 and December 19, 2006.
Brief description of amendment: The amendment revised Surveillance
Requirements (SRs) 3.5.2.8 and 3.6.7.1 due to (1) the future
replacement of the existing containment recirculation sump suction
inlet trash racks and screens with strainers, (2) the resulting
relocation of the recirculation fluid pH control (RFPC) system from the
sump, and (3) the removal of details from SR 3.6.7.1, including the
relocation of the name of the RFPC chemical to a license condition in
Appendix C to the license. The modifications will be done in the
refueling outage scheduled for the spring of 2007. The amendment also
deleted the footnote to the frequency for SR 3.5.2.5 because it is no
longer applicable.
Date of issuance: February 21, 2007.
Effective date: As of its date of issuance, and shall be
implemented prior to entry into Mode 4 during the plant startup from
the refueling outage scheduled for the spring of 2007.
Amendment No.: 180.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46940). The supplemental letters dated November 22 and December 19,
2006, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 21, 2007.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
[[Page 11402]]
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (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/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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, 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 identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
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Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory
[[Page 11403]]
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 or 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).
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: February 2, 2007.
Description of amendment request: The amendment revised Technical
Specification 3.6.1.7, ``Suppression Chamber-to-Drywell Vacuum
Breakers,'' to allow a one-time extension to the current closure
verification surveillance requirement for one of two redundant disks in
one of nine vacuum breakers until reliable position indication can be
restored in the main control room during the next refueling outage (R-
18), which is scheduled to begin on May 12, 2007.
Date of issuance: February 27, 2007.
Effective date: As of its date of issuance and shall be implemented
within 14 days from the date of issuance.
Amendment No.: 202.
Facility Operating License No.: NPF-21: Amendment revises the
technical specifications and license.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. 72 FR 6606, published February 12, 2007. The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received. The notice
also provided an opportunity to request a hearing within 60 days after
the date of publication of the notice, but indicated that if the
Commission makes a final NSHC determination, any such hearing would
take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated February 27, 2007.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Dated at Rockville, Maryland, this 2nd day of March 2007.
For the Nuclear Regulatory Commission.
Michael C. Cheok,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-4251 Filed 3-12-07; 8:45 am]
BILLING CODE 7590-01-P