[Federal Register Volume 72, Number 107 (Tuesday, June 5, 2007)]
[Notices]
[Pages 31097-31108]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-10590]
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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 May 11, 2007, to May 23, 2007. The last
biweekly notice was published on May 22, 2007 (72 FR 28717).
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
[[Page 31098]]
how that interest may be affected by the results of the proceeding. The
petition should specifically explain the reasons why intervention
should be permitted with particular reference to the following general
requirements: (1) The name, address, and telephone number of the
requestor or petitioner; (2) the nature of the requestor's/petitioner's
right under the Act to be made a party to the proceeding; (3) the
nature and extent of the requestor's/petitioner's property, financial,
or other interest in the proceeding; and (4) the possible effect of any
decision or order which may be entered in the proceeding on the
requestor's/petitioner's interest. The petition must also set forth the
specific contentions which the petitioner/requestor seeks to have
litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by 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].
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Pope County, Arkansas
Date of amendment request: April 24, 2007.
Description of amendment request: The proposed amendment will
delete the Fuel Handling Area Ventilation System (FHAVS) and associated
Ventilation Filter Testing Program (VFTP) requirements that are
included in the ANO-1 Technical Specifications (TSs) 3.7.12 and 5.5.11
and the ANO-2 TSs 3.9.11 and 6.5.11. These requirements will be
relocated to a licensee-controlled document, the unit-specific
Technical Requirements Manuals (TRM), which are controlled under 10 CFR
50.59, ``Changes, tests, and experiments.''
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 change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The FHAVS is not involved in the initiation of any accidents.
The system maintains a suitable environment for equipment operation
and personnel access. They are also designed to filter any gaseous
radioactivity that may occur during normal or accident conditions
(i.e., a fuel handling accident). On this basis, the system is
currently classified and designed as an Engineered Safety Features
(ESF) air cleanup system. The FHAVS is used during movement of
irradiated fuel, crane operation with loads over the Spent Fuel Pool
(SFP), fuel shipments, and spent resin transfer to pull possible
airborne radioactivity from the Train Bay by re-positioning manual
dampers.
Revised ANO-1 and ANO-2 analysis of the dose consequences of
a[n] FHA, to both the public and to the control room operator,
demonstrate that doses remain well within regulatory acceptance
limits without crediting filtration.
Thus there is no required safety function for the ANO-1 or ANO-2
FHAVS.
Therefore, the proposed change[s] [do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The FHAVS is not involved in the initiation of any accidents. It
was designed to
[[Page 31099]]
filter any gaseous radioactivity that may occur during normal or
accident conditions (i.e., a fuel handling accident). No physical
modifications are planned to the ANO-1 or ANO-2 FHAVS.
Revised ANO-1 and ANO-2 analysis of the dose consequences of
a[n] FHA, to both the public and to the control room operator,
demonstrate that doses remain well within regulatory acceptance
limits without crediting filtration. Thus, there is no required
safety function for the ANO-1 or ANO-2 FHAVS.
Therefore, the proposed change[s] [do] not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The FHAVS was designed to filter any gaseous radioactivity that
may occur during normal or accident conditions (i.e., a fuel
handling accident). No physical modifications are planned to the
ANO-1 or ANO-2 FHAVS.
Revised ANO-1 and ANO-2 analysis of the dose consequences of
a[n] FHA, to both the public and to the control room operator,
demonstrate that doses remain well within regulatory acceptance
limits without crediting filtration. The margin of safety, as
defined in Standard Review Plan 15.7.4, Revision 1, and GDC [General
Design Criterion] 19 has not been significantly reduced.
Therefore, the proposed change[s] [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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: April 24, 2007.
Description of amendment request: The proposed amendment will
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification
(TS) 5.2.1, ``Fuel Assemblies,'' to add Optimized ZIRLO\TM\ as an
acceptable fuel rod cladding material.
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 NRC approved topical report WCAP-12610-P-A and CENPD-404-P-
A, Addendum 1-A ``Optimized ZIRLO\TM\,'' prepared by Westinghouse
Electric Company, LLC (Westinghouse), addresses Optimized ZIRLO\TM\
and demonstrates that Optimized ZIRLO\TM\ has essentially the same
properties as currently licensed ZIRLO\TM\. The fuel cladding itself
is not an accident initiator and does not affect accident
probability. Use of Optimized ZIRLO\TM\ fuel cladding has been shown
to meet all 10 CFR 50.46 design criteria and, therefore, will not
increase the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Use of Optimized ZIRLO\TM\ clad fuel will not result in changes
in the operation or configuration of the facility. Topical report
WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material
properties of Optimized ZIRLO\TM\ are similar to those of standard
ZIRLO\TM\. Therefore, Optimized ZIRLO\TM\ fuel rod cladding will
perform similarly to those fabricated from standard ZIRLO\TM\, thus
precluding the possibility of the fuel becoming an accident
initiator and causing a new or different type of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in
the margin of safety because it has been demonstrated that the
material properties of the Optimized ZIRLO\TM\ are not significantly
different from those of standard ZIRLO\TM\. Optimized ZIRLO\TM\ is
expected to perform similarly to standard ZIRLO\TM\ for all normal
operating and accident scenarios, including both loss-of-coolant
accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where
the slight difference in Optimized ZIRLO\TM\ material properties
relative to standard ZIRLO\TM\ could have some impact on the overall
accident scenario, plant-specific LOCA analyses using Optimized
ZIRLO\TM\ properties will be performed prior to the use of fuel
assemblies with fuel rods containing Optimized ZIRLO\TM\. These LOCA
analyses will demonstrate that the acceptance criteria of 10 CFR
50.46 will be satisfied when Optimized ZIRLO\TM\ fuel rod cladding
is implemented.
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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 8, 2007.
Description of amendment request: The proposed amendment will
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification
(TS) 3.1.1.4, ``Moderator Temperature Coefficient (MTC),'' to change
the surveillance frequency to be based on effective full-power days
instead of boron concentration.
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 continues to perform the SRs [surveillance
requirements] to determine MTC at test intervals associated with the
beginning and middle of the cycle. The results of the test[s] will
continue to verify that the predicted MTC is consistent with the
measured [MTC].
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in any plant modifications
or changes in the way the plant is operated. The revised SRs for
confirming the MTC predicted values will continue to be performed at
intervals associated with the beginning and middle of the cycle.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not result in any changes to the test
method or to the frequency of the test. The change of the test
interval to use EFPD [effective full-power
[[Page 31100]]
days] instead of RCS [reactor coolant system] boron concentration
still provides assurance that the predicted MTC is consistent with
the measured [MTC].
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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1
and 2, Ogle 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 No. 50-352 and 50-353, Limerick Generating Station, Units 1
and 2, Montgomery County, Pennsylvania.
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos.
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,
York and Lancaster Counties, Pennsylvania.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois.
Docket No. 50-219, Oyster Creek Nuclear Generating Station, Ocean
County, New Jersey.
Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1,
Dauphin County, Pennsylvania.
Date of amendment request: April 12, 2007.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements related to control
room envelope (CRE) habitability in accordance with Technical
Specification Task Force (TSTF) Traveler TSTF-448, Revision 3,
``Control Room Habitability.''
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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed 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 accident previously
evaluated.
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. 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: Russell Gibbs.
Exelon Generation Company, LLC, 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, Units 1
and 2, Will County, Illinois.
Date of amendment request: April 4, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.16, ``Containment Leakage Rate
Testing Program,'' to reflect a one-time deferral of the containment
Type A, integrated leak rate test from once in 10 years to once in 15
years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes will revise Braidwood Station and Byron
Station TS 5.5.16, ``Containment Leakage Rate Testing Program'' to
reflect a one-time, five-year extension of the containment Type A
test date to enable the implementation of a 15-year test interval.
The containment is designed to contain radioactive material that
may be released from the reactor core following a design basis
[[Page 31101]]
Loss of Coolant Accident (LOCA). The test interval associated with
Type A testing is not a precursor of any accident previously
evaluated. Type A testing does provide assurance that the
containment will not exceed allowable leakage rate criteria
specified in the TS and will continue to perform its design function
following an accident. A risk assessment of the proposed changes has
concluded that there is an insignificant increase in total
population dose rate and an insignificant increase in the
conditional containment failure probability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes for a one-time, five-year extension of the
Type A tests for Braidwood Station and Byron Station will not affect
the control parameters governing unit operation or the response of
plant equipment to transient and accident conditions. The proposed
changes do not introduce any new equipment, modes of system
operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The Braidwood Station and Byron Station containment consists of
the concrete containment building, its steel liner, and the
penetrations through this structure. The structure is designed to
contain radioactive material that may be released from the reactor
core following a design basis LOCA. Additionally, this structure
provides shielding from the fission products that may be present in
the containment atmosphere following accident conditions.
The containment is a reinforced concrete structure with a
cylindrical wall, a flat foundation mat, and a shallow dome roof.
The inside surface of the containment is lined with a carbon steel
liner to ensure a high degree of leak tightness during operating and
accident conditions. The cylinder wall is pre-stressed with a post[-
] tensioning system in the vertical and horizontal directions, and
the dome roof is pre-stressed utilizing a three way post-tensioning
system.
The concrete containment building is required for structural
integrity of the containment under Design Basis Accident (DBA)
conditions. The steel liner and its penetrations establish the
leakage limiting boundary of the containment. Maintaining the
containment OPERABLE limits the leakage of fission product
radioactivity from the containment to the environment.
The integrity of the containment penetrations and isolation
valves is verified through Type B and Type C local leak rate tests
(LLRTs) and the overall leak tight integrity of the containment is
verified by a Type A integrated leak rate test (ILRT) as required by
10 CFR 50, Appendix J, ``Primary Reactor Containment Leakage Testing
for Water-Cooled Power Reactors.'' These tests are performed to
verify the essentially leak tight characteristics of the containment
at the design basis accident pressure.
The existing 10-year Type A test interval is based on past
performance. Previous Type A leakage tests conducted at Braidwood
Station Units 1 and 2, and Byron Station Units 1 and 2 indicate that
leakage from containment has been less than the 10 CFR 50 Appendix J
leakage limit.
The proposed changes for a one-time extension of the Type A
tests do not affect the method for Type A, B or C testing or the
test acceptance criteria. Type B and C testing will continue to be
performed at the frequency required by the Braidwood Station and
Byron Station Technical Specifications. The containment inspections
that are performed in accordance with the requirements of the ASME
Boiler and Pressure Vessel Code, Section XI and 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' provide a high degree of assurance that the
containment will not degrade in a manner that is only detectable by
Type A testing.
In NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' the NRC indicated that a 20-year extension for Type A
testing resulted in an imperceptible increase in risk to the public.
The NUREG-1493 study also concluded that, generically, the design
containment leak rate contributes a very small amount to the
individual risk [and] have a minimal affect on this risk. EGC has
conducted risk assessments to determine the impact of a change to
the Braidwood Station and Byron Station Type A test schedule from a
baseline value of once in 10 years to once in 15 years for the risk
measures of Large Early Release Frequency (LERF), Total Population
Dose, and Conditional Containment Failure Probability (CCFP). The
results of the risk assessments indicate that the proposed changes
to the Braidwood Station and Byron Station Type A test schedule has
a minimal impact on public risk.
Therefore, based on previous Type A test results for the
Braidwood Station and Byron Station containments, the current
containment surveillance programs at each station, and the results
of the EGC risk assessments, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
FPL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: August 7, 2006, as supplemented by
letters dated January 22, and May 14, 2007, which included a revised no
significant hazards consideration determination (NSHCD). This NSHCD is
from the May 14, 2007, supplement.
Description of amendment request: The proposed amendment would
revise the Seabrook Station Unit No. 1 (Seabrook) Facility Operating
License (FOL) and Technical Specifications (TSs). The proposed changes
would correct a joint-owner name in the operating license, remove a
license condition from Appendix C to the FOL that is no longer
applicable, and remove the list of Bases sections from the TS Index.
Additionally, the proposed amendment would remove two manual valves
from TS table 3.3.9, ``Remote Shutdown System,'' and add the
requirement that only one charging pump is permitted to be aligned for
injection into the reactor coolant system (RCS) in Modes 4, 5, and 6 to
TS 3.4.9.3, ``Overpressure Protection Systems.'' The additional
requirement proposed for TS 3.4.9.3 would allow for two pumps to be
aligned for injection under administrative controls for up to one hour
to permit swap over operations. The proposed changes would also remove
a 1-hour reporting requirement for portable makeup pump system storage
from TS 3.7.4, ``Service Water System/Ultimate Heat Sink,'' correct an
error in TS 4.7.4.3, related to the service water pumphouse water level
and delete a footnote from TS 3.7.6.2, ``Air Conditioning,'' that was
only applicable to Cycle 7. The proposed changes would also delete a
redundant reporting requirement in TS 6.6, ``Safety Limit Violation.''
Lastly, the proposed amendment would modify TS 6.7.6, ``Radioactive
Effluent Controls Program,'' to clarify the TS with respect to the
performance of dose projections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The probability or consequences of accidents previously
evaluated in the UFSAR [Updated Final Safety Analysis Report] are
unaffected by this proposed change. There is no change to any
equipment response or accident mitigation scenario, and this change
[[Page 31102]]
results in no additional challenges to fission product barrier
integrity. The proposed change does not alter the design,
configuration, operation, or function of any plant system,
structure, or component. As a result, the outcomes of previously
evaluated accidents are unaffected.
This change limits availability of the charging pumps to one
pump when in Mode 4 with the temperature of any RCS cold leg is less
than or equal to 290 [deg]F, in Mode 5, and in Mode 6 with the
reactor vessel head on and the vessel head closure bolts not fully
de-tensioned. Nonetheless, imposing this limitation does not alter
the configuration or operation of the charging pumps from that
specified in current administrative controls. Technical
Specification (TS) 3/4.5.3, ECCS [Emergency Core Cooling System]
Subsystems--Tavg Less Than 350 [deg]F, presently stipulates that
only one charging pump is maintained operable in Mode 4. Similarly,
Technical Requirement 26, Boration Systems, requires that all but
one operable charging pump be demonstrated inoperable in Modes 4, 5,
and 6. Also, the Seabrook Station Updated Final Safety Analysis
Report (UFSAR) describes the configuration of the charging pumps
during shutdown conditions: Prior to decreasing RCS temperature
below 350 [deg]F, the safety injection pumps and the non-operating
charging pumps are made inoperable. Consequently, the change does
not alter the configuration or operation of the charging pumps from
the procedures presently described in the UFSAR; rather, it only
relocates an existing limitation from the UFSAR to the technical
specifications. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
This proposed change also revises the minimum water level in the
service water system pump house required for operability of the
service water system. The value currently specified in the technical
specifications has been in error since 1986 and will be corrected
with this change. Increasing the minimum required water level from
five feet to 25.1 feet does not alter the configuration or operation
of the service water system. Following discovery of this
discrepancy, administrative controls established a minimum water
level of approximately 25 feet. Moreover, monitoring of the service
water pump house level during 2005 observed that the level, which is
controlled by the ocean tides, is normally greater than 26 feet.
During this period the minimum and maximum pump house water levels
were 26.3 and 48.57 feet, respectively. This administrative change
has no affect on the actual operation or configuration of the
service water system. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed revision to TS Table 3.3-9, Remote Shutdown System,
eliminates valves MS-V127 and MS-V128 from the table. Located in the
main steam supply line to the turbine-driven emergency feedwater
(TDEFW) pump, these are locked open, manually operated, valves.
Supplement 4 of NUREG 0896, Safety Evaluation Report, discusses the
modifications made to the Emergency Feedwater System (EFW) to
address problems experienced with the EFW steam supply lines during
hot functional testing. A design change, installed in 1991, changed
MS-V127 and MS-V128 to normally open valves, replaced the valves'
pneumatic actuators with gear-operated manual operators, and re-
assigned the EFW actuation and containment isolation functions of
these valves to new automatic isolation valves (MS-V393 and MS-V394)
in the TDEFW pump steam supply line. As a result, the elimination of
MS-V127 and MS-V128 from TS Table 3.3-9 does not alter the design,
configuration, operation, or function of these valves with regard to
operation of the EFW system because in the existing design these
normally open valves are not required to re-position to support
operation of the TDEFW pump. Automatic valves MS-V393 and MS-V394,
which actuate to initiate operation of the TDEFW pump, are
appropriately under the control of TS Table 3.3-9. This proposed
change does not alter the design, configuration, operation, or
function of the EFW steam supply valves. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The other changes in this proposed amendment correct errors,
remove an outdated license condition, remove an inconsistency
between indexes, and revise a reporting requirement. These changes
are administrative in nature and do not impact the design,
configuration, operation, or function of any plant system,
structure, or component. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes (1) relocate an existing limitation from
the UFSAR to the technical specifications regarding availability of
the charging pumps, (2) revise the minimum water level in the
service water system pump house required for operability of the
service water system, (3) eliminate valves MS-V127 and MS-V128 from
TS Table 3.3-9, and (4) make administrative changes to the TS that
correct errors, remove an outdated license condition and an
inconsistency between indexes and revises a reporting requirement.
No new accident scenarios, failure mechanisms, or limiting single
failures are introduced as a result of the proposed change. The
proposed change does not challenge the performance or integrity of
any safety-related system. The ability of any operable structure,
system, or component to perform its designated safety function is
unaffected by this change. The proposed change neither installs or
removes any plant equipment, nor alters the design, physical
configuration, or mode of operation of any plant structure, system,
or component. No physical changes are being made to the plant, so no
new accident causal mechanisms are being introduced. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of safety-related
systems and components. The proposed change relocates an existing
limitation from the UFSAR to the technical specifications regarding
availability of the charging pumps during operation in Mode 4 with
the temperature of any RCS cold leg is less than or equal to 290
[deg]F, in Mode 5, and in Mode 6 with the reactor vessel head on and
the vessel head closure bolts not fully de-tensioned. Nonetheless,
imposing this limitation does not alter the configuration or
operation of the charging pumps from those specified in current
administrative controls and the UFSAR. The proposed change includes
revising the minimum water level in the service water system pump
house required for operability of the service water system. This
change replaces a non-conservative, incorrect value in the TS with a
minimum required water level that is consistent with the design
basis for the system. The elimination of MS-V127 and MS-V128 from TS
Table 3.3-9 does not alter the design, configuration, operation, or
function of these valves with regard to operation of the EFW system
because in the existing design these normally open valves are not
required to re-position to support operation of the TDEFW pump.
Automatic valves MS-V393 and MS-V394, which actuate to initiate
operation of the TDEFW pump, are appropriately under the control of
TS Table 3.3-9. Last, the proposed amendment makes administrative
changes to the TS that correct errors, remove an outdated license
condition and an inconsistency between indexes and revises a
reporting requirement.
The proposed changes do not alter the design, configuration,
operation, or function of any plant system, structure, or component.
The ability of any operable structure, system, or component to
perform its designated safety function is unaffected by this change.
Therefore, the margin of safety as defined in the TS is not reduced
and the proposed change does not involve a significant reduction in
a margin of safety.
Based upon the reasoning presented above 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Harold K. Chernoff.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California.
Date of amendment request: April 4, 2007.
[[Page 31103]]
Description of amendment request: The licensee has proposed
amending the existing license to allow the results of near-term
surveys, performed on a portion of the plant site, to be included in
the eventual Final Status Survey (FSS) for license termination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow survey results for a specific
area within the licensed site area, performed prior to Humboldt Bay
Power Plant (HBPP) Unit 3 decommissioning and dismantlement
activities, to be used in the overall licensed site area Final
Status Survey (FSS) for license termination. The FSS will be
performed following completion of HBPP Unit 3 decommissioning and
dismantlement activities. This proposed change would not change
plant systems or accident analysis, and as such, would not affect
initiators of analyzed events or assumed mitigation of accidents.
Therefore, the proposed change does not increase the probability or
consequences of an accident previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
The proposed change does not involve a physical alteration to
the plant or require existing equipment to be operated in a manner
different from the present design. Implementation of a cross
contamination prevention and monitoring plan will be done in
accordance with plant procedures and licensing bases documents.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed change has no effect on existing plant equipment,
operating practices, or safety analysis assumptions. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Antonio Fernandez, Esquire, Pacific Gas
& Electric Company, Post Office Box 7442, San Francisco, CA 94120.
NRC Acting Branch Chief: Kristina Banovac.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: April 17, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) and license to establish more
effective and appropriate action, surveillance, and administrative
requirements related to ensuring the habitability of the control room
envelop (CRE) in accordance with Nuclear Regulatory Commission (NRC)
approved TS Task Force (TSTF) Standard Technical Specification change
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' The NRC
staff issued a ``Notice of Availability of Technical Specification
Improvement to Modify Requirements Regarding Control Room Envelope
Habitability Using the Consolidated Line Item Improvement Process''
associated with TSTF-448, Revision 3, in the Federal Register on
January 17, 2007 (72 FR 2022). The notice included a model safety
evaluation, a model no significant hazards consideration (NSHC)
determination, and a model license amendment request. In its
application dated April 17, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
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
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
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 the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
[[Page 31104]]
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: April 15, 2007.
Description of amendment request: The proposed amendments would
modify the Technical Specifications (TSs) and license to establish more
effective and appropriate action, surveillance, and administrative
requirements related to ensuring the habitability of the control room
envelop (CRE) in accordance with Nuclear Regulatory Commission (NRC)
approved TS Task Force (TSTF) Standard Technical Specification change
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' The NRC
staff issued a ``Notice of Availability of Technical Specification
Improvement to Modify Requirements Regarding Control Room Envelope
Habitability Using the Consolidated Line Item Improvement Process''
associated with TSTF-448, Revision 3, in the Federal Register on
January 17, 2007 (72 FR 2022). The notice included a model safety
evaluation, a model no significant hazards consideration (NSHC)
determination, and a model license amendment request. In its
application dated April 15, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
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
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. Therefore, this
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 the Margin of Safety
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition. 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: March 30, 2007.
Description of amendment requests: The proposed amendment revises
Technical Specifications (TSs) 3.8.1, ``AC [alternating current]
Sources--Operating,'' 3.8.4, ``DC [direct current] Sources--
Operating,'' 3.8.5, ``DC Sources--Shutdown,'' 3.8.6, ``Battery Cell
Parameters,'' 3.8.7, ``Inverters--Operating,'' and 3.8.9,
``Distribution Systems--Operating.'' This change will also add a new
Battery Monitoring and Maintenance Program, Section 5.5.2.16. The
proposed TS changes will provide operational flexibility supported by
DC electrical subsystem design upgrades that are in progress. These
upgrades will provide increased capacity batteries, additional battery
chargers, and the means to cross-connect DC subsystems while meeting
all design battery loading requirements. With these modifications in
place, it will be feasible to perform routine surveillances as well as
battery replacements online.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed changes to Technical Specifications (TS) 3.8.4 and
3.8.6 would allow extension of the Completion Time (CT) for
inoperable Direct Current (DC) distribution subsystems to manually
cross-connect DC distribution buses of the same safety train of the
operating unit for a period of 30 days. Currently the CT only allows
for 2 hours to ascertain the source of the problem before a
controlled shutdown is initiated. Loss of a DC subsystem is not an
initiator of an event. However, complete loss of a Train A
(subsystems A and C) or Train B (subsystems B and D) DC system would
initiate a plant transient/plant trip.
Operation of a DC Train in cross-connected configuration does
not affect the quality of DC control and motive power to any system.
Therefore, allowing the cross-connect of DC distribution systems
does not significantly increase the probability of an accident
previously evaluated in Chapter 15 of the Updated Final Safety
Analysis Report (UFSAR).
The above conclusion is supported by Probabilistic Risk
Assessment (PRA) evaluation which encompasses all accidents,
[[Page 31105]]
including UFSAR Chapter 15. The Frequency for Surveillance
Requirements in TS 3.8.4.3 is changed from 24 months to 30 months.
San Onofre Nuclear Generating Station (SONGS) experience has
indicated that there have been no battery failures using the 24-
month test frequency for battery service tests, and extending the
interval to 30 months is not expected to affect SONGS' capability to
detect battery health and capacity. Also, the routine test frequency
of 30 months will better dove-tail with the scheduling of the more
rigorous 60-month interval battery performance of modified
performance discharge tests.
Enhancements from TSTF-360, Rev. 1 and IEEE 450 have been
incorporated into Limiting Conditions for Operation (LCOs) 3.8.4,
3.8.5, and 3.8.6. These changes do not impact the probability or
consequences of an accident previously evaluated.
Further changes are made of an editorial nature or provide
clarification only. For example, discussions regarding electrical
`Trains' and `Subsystems' will be in more conventional terminology.
LCOs affected by editorial changes include 3.8.1, 3.8.4, 3.8.5,
3.8.6, 3.8.7, and 3.8.9.
The changes being proposed in the TS do not affect assumptions
contained in other safety analyses or the physical design of the
plant, nor do they affect other Technical Specifications that
preserve safety analysis assumptions.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Will operation of the facility in accordance with this
proposed change create the possibility of [a] new or different kind
of accident from any accident previously evaluated?
Response: No.
The proposed change modifies surveillances and LCOs for
batteries and chargers to meet the requirements of IEEE 450-2002
whose intent is to maintain the same equipment capability as
previously assumed in our commitment to IEEE 450-1980.
The proposed change will allow the cross-tie of DC subsystems
and allow extension of the CT for an inoperable subsystem to 30
days. Failure of the cross-tied DC buses and/or associated
battery(ies) is bounded by existing evaluations for the failure of
an entire electrical train.
Swing battery chargers are added to increase the overall DC
system reliability. Administrative and mechanical controls will be
in place to ensure the design and operation of the DC systems
continue to meet the UFSAR design basis.
LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 revisions are
editorial clarifications and do not affect plant design.
Therefore, operation of the facility in accordance with this
proposed change will not create the possibility of [a] new or
different kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Changes in accordance with IEEE 450 and TSTF-360, Rev. 1
maintain the same level of equipment performance stated in the UFSAR
and the current Technical Specifications.
Swing battery chargers are added to increase the overall DC
system reliability. Administrative and mechanical controls will be
in place to ensure the design and operation of the DC systems
continue to meet the UFSAR design basis.
The addition of the DC cross-tie capability proposed for LCO
3.8.4 has been evaluated, as described previously, using PRA and
determined to be of acceptable risk as long as the duration while
cross-tied is limited to 30 days. An LCO has been included as part
of this proposed change to ensure that plant operation, with DC
buses cross-tied, will not exceed 30 days.
All remaining changes are editorial.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Thomas G. Hiltz.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: April 25, 2007.
Description of amendment request: The proposed amendment would
revise the technical specifications to increase the maximum number of
tritium producing burnable absorber rods (TPBARs) that can be
irradiated in the reactor from 240 to 400.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the maximum number of TPBARs in the
core. The required boron concentration for the cold leg accumulators
(CLAs) and RWST [Refueling Water Storage Tank] remains unchanged.
The current boron concentration has been demonstrated to maintain
the required accident mitigation safety function for the CLAs and
RWST with the higher number of TPBARs and this will be verified for
each core that contains TPBARs as part of the normal reload
analysis. The CLAs and RWST safety function is to mitigate accidents
that require the injection of borated water to cool the core and to
control reactivity. These functions are not potential sources for
accident generation and the modification of the number of TPBARs
will not increase the potential for an accident. Therefore, the
possibility of an accident is not increased by the proposed changes.
The current boron concentration levels are supported by the proposed
number of TPBARs in the core. Since the current boron concentration
levels will continue to maintain the safety function of the CLAs and
RWST in the same manner as currently approved, the consequences of
an accident are not increased by the proposed changes.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change only modifies the maximum number of TPBARs
in the core. The boron concentrations for accident mitigation
functions of the CLAs and RWST remain unchanged. These functions do
not have a potential to generate accidents as they only serve to
perform mitigation functions associated with an accident. The
proposed modification will maintain the mitigation function in an
identical manner as currently approved. There are no plant equipment
or operational changes associated with the proposed revision.
Therefore, since the CLA and RWST functions are not altered and the
plant will continue to operate without change, the possibility of a
new or different kind of an accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This change proposes a change to the maximum number of TPBARs in
the core. The boron concentration requirements that support the
accident mitigation functions of the CLAs and RWST remain unchanged.
The proposed change does not alter any plant equipment or components
and does not alter any setpoints utilized for the actuation of
accident mitigation system or control functions. The proposed number
of TPBARs, in conjunction with the current boron concentration
values, has been demonstrated to provide an adequate level of
reactivity control for accident mitigation and this will be verified
for each core that contains TPBARs as part of the normal reload
analysis. Therefore, the proposed change will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
[[Page 31106]]
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(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: August 16, 2006, as
supplemented by letters dated January 25 and March 8, 2007.
Brief description of amendments: The amendments revised Technical
Specifications (TS) requirements in Surveillance Requirements (SRs) to
allow for surveillances to be performed in modes that are not currently
allowed in TS and to require certain SRs to be performed at a power
factor of <=0.89 if performed with the emergency diesel generators
synchronized to the grid unless grid conditions do not permit.
Date of issuance: May 16, 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--167, Unit 2--167, Unit 3--167.
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: October 24, 2006 (71 FR
62307). The supplements dated January 25 and March 8, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register on October 24, 2006
(71 FR 62307).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 16, 2007.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: June 28, 2006, as supplemented
by letter dated November 2, 2006.
Brief description of amendment: The amendment changes Kewaunee
Power Station Technical Specifications 3.3.b.3.B and 3.3.b.4.A to
increase the minimum required boron concentration in the refueling
water storage tank from 2400 parts per million (ppm) to 2500 ppm.
Date of issuance: May 18, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 192.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43530). The supplemental letter contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 18, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: January 26, 2006, as
supplemented by letter dated December 21, 2006.
Brief description of amendment: The amendment will allow additional
startup and operating flexibility and an expanded operating domain
resulting from the proposed implementation of the Average Power Range
Monitor, Rod Block Monitor Technical Specification improvement program
concurrently with the proposed implementation of the Maximum Extended
Operating Domain Analysis, which is the combination of the power/flow
operating map expansion with Maximum Extended Load Line Limit Analysis
and increased core flow.
Date of issuance: May 17, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 287.
Facility Operating License No. DPR-59: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13171). The supplemental letter dated December 21, 2006, 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 May 17, 2007.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: November 2, 2006.
[[Page 31107]]
Description of amendment request: The proposed amendment revised
Technical Specifications requirements for inoperable snubbers
consistent with the Technical Specification Task Force 372, Revision 4.
Date of issuance: May 14, 2007.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 229.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 30, 2007 (72 FR
4307). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 14, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 31, 2006, as supplemented by
letter dated January 31, 2007.
Brief description of amendment: The amendment relocated TS 3.8.7
requirements associated with 120 volt (V) inverter Y-28 and TS 3.8.9
requirements associated with the 120 V alternating current electrical
power distribution subsystem panel C-540 to the Technical Requirements
Manual.
Date of issuance: May 15, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 230.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications/license.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65142). The supplement dated January 31, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 15, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 2, 2006.
Brief description of amendment: The amendment change deletes the
augmented testing requirement for containment purge supply and exhaust
isolation valves with resilient seal materials and allows the
surveillance intervals to be set in accordance with the Containment
Leakage Rate Testing Program.
Date of issuance: May 23, 2007.
Effective date: As of the date of issuance and shall be implemented
120 days from the date of issuance.
Amendment No.: 213.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 26, 2006 (71
FR 56191). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 23, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 6, 2006.
Brief description of amendment: This amendment revised the
Ventilation Filter Test Program (VFTP) in Technical Specification
5.5.7, to correct the flow rate units specified in the VFTP, from
standard cubic feet per minute to cubic feet per minute.
Date of issuance: May 9, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 143.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51228).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 9, 2007.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: May 25, 2006, as supplemented by
letters dated December 21, 2006, March 14, 2007, and March 30, 2007.
Brief description of amendment: The amendment revises the Technical
Specification Steam Generator tube Surveillance Program to one modeled
after Technical Specification Task Force (TSTF) Traveler TSTF-449,
``Steam Generator Tube Integrity.''
Date of issuance: May 16, 2007.
Effective date: Date of issuance, to be implemented within 90 days.
Amendment No.: 223.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51229). The supplements dated December 21, 2006, March 14 and 30, 2007,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 16, 2006.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: September 29, 2006, as
supplemented by letter dated December 7, 2006, and February 12, 2007.
Brief description of amendment: The amendment revises Technical
Specification 3.7.8, ``Service Water (SW) System,'' from an electrical
train-based specification to a pump-based specification. Revisions to
the Limiting Conditions for Operation, Required Actions, Completion
Times, and Surveillance Requirements have been made to require a
specific number of SW water pumps to be operable rather than SW trains.
Date of issuance: May 16, 2007.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 102.
Renewed Facility Operating License No. DPR-18: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: November 7, 2006 (71 FR
65144).
The letters dated December 7, 2006, and February 12, 2007, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the 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 May 16, 2007.
[[Page 31108]]
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: May 24, 2006, as supplemented on
February 15, 2007.
Brief description of amendment: The amendment revises the Virgil C.
Summer Nuclear Station Technical Specifications and provides associated
Bases that are modeled after Technical Specification Task Force (TSTF)
traveler, TSTF-449, Revision 4, ``Steam Generator Tube Integrity.'' A
notice of availability for this TS improvement using the consolidated
line item improvement process was published in the Federal Register on
May 6, 2005 (70 FR 24126).
Date of issuance: May 15, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 179.
Renewed Facility Operating License No. NPF-12: Amendment revises
the TSs.
Date of initial notice in Federal Register: June 20, 2006 (71 FR
35458). The supplemental letter provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration. The Commission's
related evaluation of the amendment is contained in a safety evaluation
dated May 15, 2007.
No significant hazards consideration comments received: No.
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 application for amendments: February 2, 2007.
Brief description of amendments: The amendments revised the
Technical Specifications Limiting Condition for Operation (LCO) 3.10.1
to be consistent with TSTF-484, Revision 0, ``Use of Technical
Specification 3.10.1 for Scram Time Testing Activities.''
Date of issuance: May 17, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 251, 195.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11395).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 17, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of May 2007.
For the Nuclear Regulatory Commission.
Timothy J. McGinty,
Acting Director, Division of Operating Reactor Licensing Office of
Nuclear Reactor Regulation.
[FR Doc. E7-10590 Filed 6-4-07; 8:45 am]
BILLING CODE 7590-01-P