[Federal Register Volume 72, Number 29 (Tuesday, February 13, 2007)]
[Notices]
[Pages 6780-6795]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-2323]
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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 January 19, 2007, to February 1, 2007. The
last biweekly notice was published on January 30, 2007 (72 FR 4304).
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
[[Page 6781]]
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 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:
[[Page 6782]]
(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)7ndash;(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 No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 26, 2006.
Description of amendment request: The proposed change deletes
reference to the containment fan cooler (CFC) condensate flow switch
from Technical Specification (TS) 3.4.5.1, ``Reactor Coolant System
Leakage--Leakage Detection Instrumentation,'' and to modify or delete
associated actions.
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 Reactor Coolant System (RCS) leakage detection systems are
passive monitoring systems therefore the proposed changes do not
affect reactor operations or accident analyses and have no
radiological consequences. The proposed change continues to require
diverse methods of monitoring leakage. The gaseous radioactivity
monitor, although not included in the TSs and the CFC condensate
flow switches, which are proposed for removal from the TSs, will be
maintained functional and available.
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 introduces no new mode of plant operation or
any plant modification. The RCS leakage detection instrumentation is
used solely for monitoring purposes and is not part of plant control
instruments or engineered safety feature actuation circuits. The
change does not vary or affect any plant operating condition or
parameter.
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 modify any of the RCS leakage
detection instrumentation. The proposed change continues to require
diverse methods of monitoring leakage. In addition, although not
required by TS, multiple means of diverse monitoring RCS leakage
will remain functional and available.
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: David Terao.
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 amendment request: January 18, 2007.
Description of amendment request: The proposed change will revise
the description of Grand Gulf Nuclear Station Technical Specification
4.2.2, ``Control Rod Assemblies,'' to allow to the use of hafnium as an
additional type of control 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 has specifically approved the use of hafnium as neutron
absorbing material for use in BWR [boiling-water reactor] control
rod assemblies. The use of hafnium in control rods as a neutron
absorber material does not significantly alter the neutronic or
mechanical functional characteristics of the control rods. Control
rod designs using hafnium have been successfully used in other BWRs.
Since control rods that utilize hafnium have a longer lifetime, the
probability of some accidents involving the handling, on-site
storage, and shipping of irradiated rods will actually be reduced.
The proposed change does not alter the required number of control
rods nor does it affect any of the specifications related to the
control rods (e.g., the shutdown margin and scram timing
requirements are unaffected).
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 application of a control rod design using hafnium as an
absorber material does not produce any new mode of plant operation
or alter the control rods in such a way as to affect their function
or operability since the new control rods are designed to be
compatible with the existing control rods.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
[[Page 6783]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not significantly affect the neutronic
or mechanical characteristics of the control rods since the hafnium
containing controls rods are designed to be compatible with the
existing design and reload licensing criteria; therefore, there is
no significant change in the margin of safety. It does not change
the required number of existing control rods. It does not affect the
existing Technical Specifications related to control rods (e.g.,
required shutdown margin and scram time, etc.).
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: David Terao.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: October 11, 2006.
Description of amendment request: The proposed amendment would
modify the plant Improved Technical Specifications (ITSs) to implement
a more conservative requirement in ITS 3.7.7, ``Nuclear Services Closed
Cycle Cooling Water (SW) System.'' The current Action A allows the
plant to operate for up to 72 hours before initiating a shutdown when
one required SW heat exchanger is inoperable. The proposed revision
will only allow operation to continue for 8 hours before initiating a
shutdown when one required SW heat exchanger is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
(1) Does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The limiting design basis accident for CR-3 includes, as an
assumption, adequate heat removal capability by the SW system. The
amendment is being proposed to ensure the SW system performs its
design basis function. Adequate heat removal is provided by three
OPERABLE SW heat exchangers. The 8 hour completion time will reduce
the window that the plant can operate with only two SW heat
exchangers before a shutdown is required. The proposed change does
not increase the probability of an accident previously evaluated
since the amendment is not a modification to plant systems, nor a
change to plant operation that could initiate an accident.
Therefore, granting the LAR [license amendment request] does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The dose consequences of all
design basis accidents are unchanged by this proposed amendment.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously evaluated?
The function of the SW system considered in the design basis is
to remove process and operating heat from safety-related components
during normal as well as transient conditions. The proposed
amendment to limit the allowed ACTION Completion Time to 8 hours
will ensure the function of the SW system is consistent with the
design basis and will not result in changes to the design, physical
configuration of the plant or the assumptions made in the safety
analysis. The requirement does not change the function of the system
nor its ability to perform its design function. No alteration to
plant configuration or operation is proposed. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated.
(3) Does not involve a significant reduction in a margin of
safety?
CR-3's design basis considers adequate heat removal by the SW
system to cool the containment fan assembly cooling coils and fan
motors, spent fuel pool, SW pump motors and other equipment which
must function following an accident. This proposed amendment will
not alter the current design basis. By limiting the allowed ACTION
Completion Time to 8 hours, the proposed amendment to ITS 3.7.7 will
limit the time the safety function of the SW system can be
compromised. Therefore, the amendment does not result in a reduction
of the margin of safety.
The NRC staff has reviewed the analysis provided for Florida Power
Corporation 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 (Acting): Margaret H. Chernoff.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: December 13, 2006.
Description of amendment requests: The amendment application
proposes to delete Technical Specification (TS) 6.8.1.3, which provides
the requirement for submittal of the annual occupational radiation
exposure report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No
The proposed change eliminates the Technical Specification
reporting requirement for occupational radiation exposure
information, which is in excess to that required to be submitted by
regulations. The proposed change involves no changes to plant
systems or accident analyses. As such, the change is administrative
in nature and does not affect initiators of analyzed events or
assumed mitigation of accidents. 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?
No
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety? No
This change is an administrative change to reporting
requirements of occupational radiation exposure data and will not
reduce a margin of safety because it has no effect on any safety
analyses assumptions. Hence, this change is administrative in
nature. For these reasons, the proposed change does not involve a
significant reduction 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.
NRC Branch Chief: Claudia Craig.
[[Page 6784]]
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 amendment request: December 21, 2006.
Description of amendment request: The proposed amendment revises
the licensing basis to reflect a revision to the spent fuel pool
criticality analysis methodology and a new criticality analysis. In
addition, associated changes are proposed to Technical Specifications
3.7.12, ``Spent Fuel Storage,'' and 4.3.1, ``Criticality,'' to reflect
the results of the new criticality analysis.
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
Operation of the facility in accordance with the proposed
amendment request does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
presence of soluble boron in the Spent Fuel Pool (SFP) water being
used for criticality control does not increase the probability of a
dropped fuel assembly accident within the pool. The handling of the
fuel assemblies in the SFP has always been performed and will
continue to be performed in borated water.
There is no increase in the probability of the accidental
misloading of fuel assemblies into the SFP fuel storage racks when
considering the presence of soluble boron in the pool water for
criticality control. Fuel assembly placement will continue to be
controlled pursuant to approved fuel handling procedures and in
accordance with the spent fuel storage rack limitations specified in
the Technical Specifications (TS). There is no increase in the
consequences for an accidental misloading of fuel assemblies in the
SFP fuel storage racks because the criticality analyses demonstrate
that the pool will remain subcritical following an accidental
misloading.
Soluble boron credit is used to provide margin to offset
uncertainties, tolerances, and off-normal/accident conditions, and
to provide subcritical margin such that the SFP keff
[effective neutron multiplication constant] is maintained less than
or equal to 0.95. The plant-specific criticality analysis results
demonstrate that the spent fuel rack keff will remain<1.0
(at a 95/95 percent probability and confidence level) even with the
SFP flooded with unborated water.
There is no increase in the probability of the loss of normal
cooling to the SFP water when considering the presence of soluble
boron in the pool water for subcriticality control since a high
concentration of soluble boron has always been maintained in the SFP
water.
A loss of normal cooling to the SFP water causes an increase in
the temperature of the water passing through the stored fuel
assemblies. This causes a decrease in water density, which would
result in a net increase in reactivity when soluble boron is present
in the water. However, the additional negative reactivity provided
by the 2100 ppm [parts per million] boron concentration limit, above
that provided by the concentration required (805 ppm) to maintain
keff less than or equal to 0.95, will compensate for the
increased reactivity which could result from a loss of SFP cooling
event. Because adequate soluble boron will be maintained in the SFP
water the consequences of a loss of normal cooling to the SFP will
not be increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No
Under the proposed amendment, no changes are being made to the
fuel storage racks themselves, to any other systems, or to the
physical structures of the Primary Auxiliary Building. Therefore,
there are no changes proposed to the plant configuration, equipment
design, or installed equipment.
Criticality accidents in the SFP are not new or different types
of accidents. They have been analyzed in the FSAR [Final Safety
Analysis Report] and in fuel storage criticality analysis reports
associated with specific licensing amendments. The proposed new SFP
storage limitations are consistent with the assumptions made in the
new criticality analysis, and will not have any significant effect
on normal SFP operations and maintenance, and do not create the
possibility of a new or different kind of accident. Verifications
will continue to be performed to ensure that the SFP loading
configuration meets specified requirements.
The current TS includes a SFP boron concentration limit that
conservatively bounds the boration assumption of the new criticality
analysis. Since soluble boron has always been maintained in the SFP
water, implementation of this requirement for SFP criticality
control purposes has have no effect on normal pool operations and
maintenance. Also, since soluble boron has always been present in
the SFP, a dilution event has always been a possibility. The loss of
substantial amounts of soluble boron from the SFP that could lead to
keff exceeding 0.95 was evaluated as part of the analyses
in support of this license amendment request. The evaluation
demonstrates that a dilution of the SFP boron concentration from the
minimum TS concentration of 2100 to 805 ppm is not credible.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
(3) Does the proposed amendment result in a significant
reduction in a margin of safety?
Response: No
The proposed Technical Specification changes providing the
resulting spent fuel storage operation limits provide adequate
safety margin to ensure that the stored fuel assembly array always
remains subcritical. These limits are based on a plant-specific
criticality analysis performed in accordance with the present
Westinghouse spent fuel rack criticality analysis methodology which
allows credit for soluble boron.
The criticality analysis takes credit for soluble boron to
ensure that keff will be less than or equal to 0.95 under
normal circumstances. While the criticality analysis used credit for
soluble boron, storage configurations have been defined using 95/95
keff calculations to ensure that the spent fuel rack
keff is less than unity (0.995) with no soluble boron.
Soluble boron credit is used to provide safety margin to offset
uncertainties, tolerances, and off-normal/accident conditions, and
to provide subcritical margin such that the SFP keff is
maintained less than or equal to 0.95.
The loss of substantial amounts of soluble boron from the SFP
that could lead to keff exceeding 0.95 was evaluated as
part of the analyses in support of this license amendment request.
The evaluation demonstrates that a dilution of the SFP boron
concentration from the minimum TS concentration of 2100 to 805 ppm
is not credible. Also, the plant-specific criticality analysis
results demonstrate that even if a complete dilution were to occur
the spent fuel rack keff would remain <1.0 (at a 95/95
percent probability and confidence level) with the SFP flooded with
unborated water. The plant-specific criticality analysis performed
in accordance with the conservative analysis methodology of the
Westinghouse licensing topical report demonstrates that the
requirements of 10 CFR 50.68 and 10 CFR 50, Appendix A, General
Design Criterion 62 will be satisfied. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Patrick D. Milano.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: December 29, 2006.
Description of amendment request: The proposed amendments would
[[Page 6785]]
revise Technical Specification (TS) 5.5.8 to indicate that the
Inservice Testing Program shall include testing frequencies applicable
to the American Society of Mechanical Engineers Code for Operations and
Maintenance (ASME OM Code), and to indicate that there may be some non-
standard frequencies specified as 2 years or less in the Inservice
Testing Program to which the provisions of Surveillance Requirement
(SR) 3.0.2 are applicable. The proposed changes are consistent with
NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-
479, Revision 0, ``Changes to Reflect Revision of 10 CFR 50.55a,'' and
TSTF-497, Revision 0, ``Limit Inservice Testing Program SR 3.0.2
Application to Frequencies of 2 Years or Less.''
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 changes revise TS 5.5.8, ``lnservice Testing
Program,'' for consistency with 10 CFR 50.55a(f)(4) requirements
regarding inservice testing of pumps and valves. The proposed change
incorporates revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves.
The proposed changes do not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. They do not involve the addition or removal of any
equipment, or any design changes to the facility. Therefore, the
proposed changes do not represent 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 proposed changes do not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure. Therefore, this proposed change
does not create the possibility of an accident of a different kind
than previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No
The proposed changes revise TS 5.5.8, ``lnservice Testing
Program,'' for consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the inservice testing of pumps and valves.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves. The safety function of the affected pumps and valves will be
maintained. Therefore, this 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 requests involve no significant hazards consideration.
Attorney for licensee: Antonio Fernandez, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: December 29, 2006.
Description of amendment requests: The proposed amendments will
revise Technical Specification (TS) 5.5.16 for consistency with the
requirements of 10 CFR 50.55a(g)(4) for components classified as Code
Class CC. This regulation requires licensees to update their
containment inservice inspection requirements in accordance with
Subsections IWE and IWL of Section XI, Division I of the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code
as limited by 10 CFR 50.55a(b)(2)(vi) and modified by 10 CFR
50.55a(b)(2)(viii) and 10 CFR 50.55a(b)(2)(ix). This license amendment
request is consistent with NRC-approved Industry/Technical
Specification Task Force (TSTF) Traveler number TSTF-343, ``Containment
Structural Integrity.''
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 revises the Technical Specification (TS)
administrative controls programs for consistency with the
requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components
classified as Code Class CC.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Containment Leakage Rate Testing Program. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The frequency of visual examinations of the concrete
surfaces of the containment and the mode of operation during which
those examinations are performed has no relationship to or adverse
impact on the probability of any of the initiating events assumed in
the accident analyses. The proposed change would allow visual
examinations that are performed pursuant to NRC-approved ASME
[Code,] Section XI requirements (except where relief has been
granted by the NRC) to meet the intent of visual examinations
required by Regulatory Guide 1.163, without requiring additional
visual examinations pursuant to the Regulatory Guide. The intent of
early detection of deterioration will continue to be met by the more
rigorous requirements of the Code-required visual examinations. As
such, the safety function of the containment as a fission product
barrier is maintained.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. It does not involve the addition or removal of any
equipment, or any design changes to the facility.
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 accident from any accident previously evaluated?
Response: No.
The proposed change revises the TS Administrative Controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a(g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces of the containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or a change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or a malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released offsite and there is no increase in individual or
cumulative occupational exposure.
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?
[[Page 6786]]
Response: No.
The proposed change revises the TS Administrative Controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a(g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces of the containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The safety function of the containment as a fission product
barrier will be maintained.
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 requests involve no significant hazards consideration.
Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: December 29, 2006.
Description of amendment requests: The proposed amendments will
revise Technical Specification (TS) 3.4.1, ``RCS [Reactor Coolant
System] Pressure, Temperature, and Flow Departure from Nucleate Boiling
(DNB) Limits,'' and TS 5.6.5, ``CORE OPERATING LIMITS REPORT (COLR).
This license amendment request proposes to relocate the RCS DNB
parameters for pressurizer pressure and RCS average temperature to the
COLR. This relocation is consistent with Technical Specification Task
Force Traveler TSTF-339, Revision 2, ``Relocate TS Parameters to
COLR.'' TS 5.6.5 is revised to add topical reports WCAP-8567-P-A,
``Improved Thermal Design Procedure,'' and WCAP-11596-P-A,
``Qualification of the PHOENIX-P/ANC Nuclear Design System for
Pressurized Water Reactor Cores,'' by name and title only. These
changes are consistent with TSTF-363, Revision 0, ``Revise Topical
Report References in ITS 5.6.5, COLR.''
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 proposed changes are programmatic and administrative in
nature, and do not physically alter safety-related systems or affect
the way in which safety-related systems perform their functions. The
proposed changes relocate cycle-specific parameters from Technical
Specification (TS) 3.4.1 to the Core Operating Limits Report (COLR).
This does not change plant design or affect system operating
parameters. The proposed changes do not, by themselves, alter any of
the parameters. Removal of the cycle-specific parameters from the TS
does not eliminate existing requirements to comply with the
parameters. Also, TS 5.6.5 is revised to add topical reports WCAP-
8567-P-A, ``Improved Thermal Design Procedure,'' and WCAP-11596-P-A,
``Qualification of the PHOENIX-P/ANC Nuclear Design System for
Pressurized Water Reactor Cores,'' as they are approved analytical
methods for determining core operating limits.
Although relocation of the cycle-specific parameters to the COLR
would allow revision of the affected parameters without prior NRC
approval, there is no significant effect on the probability or
consequences of an accident previously evaluated. Future changes to
the COLR parameters could result in event consequences that are
either slightly less or slightly more severe than the consequences
for the same event using the present parameters. The differences
would not be significant and would be bounded by the existing
requirement of TS 5.6.5c to meet the applicable limits of the safety
analyses.
The cycle-specific parameters being transferred from the TS to
the COLR will continue to be controlled under existing programs and
procedures. The Final Safety Analysis Report Update (FSARU) accident
analyses will continue to be examined with respect to changes in the
cycle-dependent parameters obtained using NRC reviewed and approved
reload design methodologies to ensure that the transient evaluation
of new reload designs are bounded by previously accepted analyses.
This examination will continue to be performed pursuant to 10 CFR
50.59 requirements, ensuring that future reload designs use NRC-
approved methodologies and do not involve more than a minimal
increase in the probability or consequences of an accident
previously evaluated in the FSARU.
The proposed changes do not allow for an increase in plant power
levels, do not increase the production, and do not alter the flow
path or method of disposal of radioactive waste or byproducts.
Therefore, the proposed changes do not change the type or increase
the amount of effluents released offsite.
The proposed changes to TS 5.6.5b to reference only the topical
report number and title for five of the topical reports do not alter
the analytical methods that have been previously reviewed and
approved by the NRC. This method of referencing topical reports
would allow the use of current topical reports to support limits in
the COLR without having to submit a request for an amendment to the
operating license. Implementation of revisions to these topical
reports would still be reviewed in accordance with 10 CFR 50.59 and,
where required, revisions would be submitted to the NRC for approval
prior to implementation.
Therefore, the proposed change does 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 accident from any accident previously evaluated?
Response: No.
The proposed changes that relocate cycle-specific parameters
from the TS to the COLR, thus removing the requirement for prior NRC
approval of revisions to those parameters, do not involve a physical
change to the plant. No new equipment is being introduced, and
installed equipment is not being operated in a new or different
manner. No changes are being made to the parameters within which the
plant is operated, other than their relocation to the COLR. No
protective or mitigative action setpoints are affected by the
proposed changes. The proposed changes will not alter the manner in
which equipment operation is initiated, nor will the functional
demands on credited equipment be changed. No change to procedures
that ensure the plant remains within analyzed limits are being
proposed, and no change is being made to procedures relied upon to
respond to an off-normal event. As such, no new failure modes are
being introduced.
Relocation of cycle-specific parameters does not influence,
impact, or contribute in any way to the possibility of a new or
different kind of accident. The relocated cycle-specific parameters
will continue to be calculated using the NRC-reviewed and approved
methodology. The proposed changes do not alter assumptions made in
the safety analysis, and operation within the core operating limits
will continue.
The proposed changes to reference only the topical report number
and title do not alter the use of the analytical methods that have
been previously reviewed and approved by the NRC. This method of
referencing topical reports would allow the use of current topical
reports to support limits in the COLR without having to submit a
request for an amendment to the operating license. Implementation of
revisions to topical reports would still be reviewed in accordance
with 10 CFR 50.59 and, where required, would receive NRC review and
approval.
The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is
a clarification to provide a complete listing of approved analytical
methods used for determining core operating limits.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are
[[Page 6787]]
initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety-related
systems perform their functions. No protective or mitigative action
setpoints are affected by the proposed changes. Therefore,
sufficient equipment remains available to actuate upon demand for
the purpose of mitigating an analyzed event. As the proposed changes
to relocate cycle-specific parameters to the COLR will not affect
plant design or system operating parameters, there is no detrimental
impact on any equipment design parameter, and the plant will
continue to be operated within prescribed limits.
The development of cycle-specific parameters for future reload
designs will continue to conform to NRC-reviewed and approved
methodologies, and will be performed pursuant to 10 CFR 50.59 to
assure that the plant operates within cycle-specific parameters.
The proposed changes to reference only the topical report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing topical reports would allow
the use of current NRC-approved topical reports to support limits in
the COLR without having to submit a request for an amendment to the
operating license. Implementation of revisions to topical reports
would still be reviewed in accordance with 10 CFR 50.59 and, where
required, receive NRC review and approval.
The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is
a clarification to provide a complete listing of approved analytical
methods used for determining core operating limits.
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 requests involve no significant hazards consideration.
Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: January 11, 2007.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TSs) to support replacement of the
steam generators (SGs) at Diablo Canyon Power Plant, Unit Nos. 1 and 2.
Revisions are proposed to TS 3.3.2, ``Engineered Safety Feature
Actuation System (ESFAS) Instrumentation,'' TS 5.5.9, ``Steam Generator
(SG) Program,'' and TS 5.6.10, ``Steam Generator (SG) Tube Inspection
Report.'' The replacement SGs are to be installed during the Diablo
Canyon Power Plant, Unit No. 2, 14th refueling outage (2R14), currently
scheduled for February 2008, and the Unit No. 1, 15th refueling outage
(1R15), currently scheduled for January 2009.
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 revised engineered safety feature actuation system (ESFAS)
steam generator (SG) Water Level-High High feedwater isolation
Nominal Trip Setpoint and Allowable Value have been determined using
the existing setpoint methodology approved for Diablo Canyon Power
Plant. The setpoint analysis for the replacement steam generators
(RSGs) accounts for the setpoint uncertainties specific to the RSG
design. The revised Feedwater Isolation SG Water Level-High High (P-
14) Nominal Trip Setpoint and Allowable Value are applied using a
conservative surveillance requirement methodology. The function of
the ESFAS instrumentation is unchanged. The Feedwater Isolation SG
Water Level-High High (P-14) ESFAS instrumentation will continue to
function in a manner consistent with the plant design basis and
satisfy all the requirements of the safety analyses.
The probability and consequences of accidents previously
evaluated in the Final Safety Analysis Report (FSAR) Update are not
adversely affected because the revised Feedwater Isolation SG Water
Level-High High (P-14) Nominal Trip Setpoint and Allowable Value
continue to assure a conservative plant response to high SG level,
consistent with the safety analyses and licensing basis.
The proposed changes revise and clarify the surveillance
requirements for ESFAS Function 5.b, Feedwater Isolation SG Water
Level-High High (P-14). These changes ensure that this function will
actuate as assumed in the safety analyses.
The proposed changes to TS 5.5.9 delete the alternate repair
criteria (ARC) for the existing SGs, incorporate tube inspection
periods applicable to Alloy 690 thermally treated tubes, and delete
the TS 5.6.10 reporting requirements for ARC. The TS 5.5.9 SG
structural integrity, accident induced leakage, and operational
leakage performance criteria will continue to be met for the RSGs.
Meeting the SG performance criteria provides reasonable assurance
that the SG tubes will remain capable of maintaining reactor coolant
pressure boundary integrity throughout each operating cycle and in
the unlikely event of a design basis accident. Removal of the ARC
for the existing SGs will ensure that all tubes found by inservice
inspection to contain flaws with a depth equal to or exceeding 40
percent of the nominal tube wall thickness will be plugged as
required by TS 5.5.9.c. With the revised SG tube inspection period,
the SGs will continue to meet the SG program defined by NEI [Nuclear
Energy Institute] 97-06, ``Steam Generator Program Guidelines,''
which incorporates a balance of prevention, inspection, evaluation,
repair, and leakage monitoring.
Removal of the ARC will reduce the allowable accident induced
leakage following a main steamline break accident. The proposed
changes do not have any impact on the accident induced leakage
assumed in the other design basis accidents. The changes do not have
any impact on the allowable SG operational leakage, allowable
reactor coolant system activity, or the allowable SG secondary
activity.
The proposed changes will not affect the probability of any
accident initiators. There will be no degradation in the performance
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident. There will
be no change to accident mitigation performance. The proposed
changes will not alter any assumptions or change any mitigation
actions in the radiological consequence evaluations in the FSAR
Update.
Therefore the proposed changes 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 accident from any accident previously evaluated?
Response: No.
The proposed changes will not affect the normal method of plant
operation or create new methods of plant operation related to the
Feedwater Isolation SG Water Level-High High (P-14) ESFAS setpoints.
The proposed changes to the Feedwater Isolation SG Water Level-High
High (P-14) instrumentation surveillance requirements will provide
assurance that the plant will operate within the limits assumed in
the safety analyses. The assumptions made in the setpoint analyses
for the Feedwater Isolation SG Water Level-High High (P-14) ESFAS
instrument do not create any new accidents, accident initiators, or
failure mechanisms.
The proposed changes, which delete the TS 5.5.9 ARC for the
existing SGs, incorporate tube inspection periods for Alloy 690
thermally-treated tubes in TS 5.5.9, and delete the ARC reporting
requirements in TS 5.6.10, will not introduce any adverse changes to
the plant design basis or postulated accidents resulting from
potential tube degradation. The primary-to-secondary leakage that
may be experienced during all plant conditions will be monitored to
ensure it remains within current safety analysis assumptions. The
proposed changes do not adversely affect the method of operation of
the SGs or the primary or secondary coolant controls and do not
impact other plant systems or components.
[[Page 6788]]
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The FSAR Update Excessive Heat Removal due to Feedwater System
Malfunctions event credits the Feedwater Isolation SG Water Level-
High High (P-14) ESFAS instrumentation. The safety analysis limit
assumed for the Feedwater Isolation SG Water Level-High High (P-14)
ESFAS instrumentation for this event has not changed for the safety
analyses for the RSGs. None of the acceptance criteria for Excessive
Heat Removal due to Feedwater System Malfunctions event are changed
as a result of the revised Feedwater Isolation SG Water Level-High
High (P-14) Nominal Trip Setpoint and Allowable Value. The
instrument surveillance requirement changes for the Feedwater
Isolation SG Water Level-High High (P-14) function ensure that the
instrumentation will actuate as assumed in the safety analysis.
The safety function of the SGs is maintained by ensuring the
integrity of the tubes. SG tube integrity is a function of the
design, environment, and the physical condition of the SG tubes. The
proposed changes, which delete the TS 5.5.9 ARCs for the existing
SGs, incorporate tube inspection periods for Alloy 690 thermally
treated tubes in TS 5.5.9, and delete the ARC reporting requirements
in TS 5.6.10, do not adversely impact the SG tube design or
operating environment. SG tube integrity will continue to be
maintained by implementing the SG Program to manage SG tube
inspection, assessment, and repair. The requirements established by
the SG program are consistent with those in the applicable design
codes and standards.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California
Date of amendment request: May 17, 2006.
Description of amendment request: The licensee has proposed to
modify the Physical Security Plan (PSP) to allow leaving certain
security posts temporarily under emergency conditions requiring
personnel to evacuate occupied plant areas for their health and safety.
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.
Allowing the security posts and monitoring requirements of PSP,
Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously
maintained has no impact on the probability of an accident from
occurring, especially acts of nature such as earthquakes and
tsunamis.
The HBPP Defueled Safety Analysis Report, Appendix A, and NRC
Safety Evaluation Report (SER), Section 10, dated April 29, 1987,
evaluate various accidents at HBPP. Because all fuel has been
removed from the reactor vessel and stored in the spent fuel pool,
the majority of accidents analyzed pertain to events that could only
affect spent fuel or the spent fuel pool. All accidents affecting
spent fuel or the spent fuel pool do not require security personnel
action to protect the public health and safety, or to maintain
offsite radiological doses well within regulatory limits. In
addition, NRC SER, Section 10.7, ``Impact of Tsunami Flooding,''
analyzes the impact of tsunami flooding. That analysis identifies a
likely impact of the tsunami to be a release of the radwaste tank
radionuclide contents to the bay and some damage to the reactor
building. For both situations, no security personnel action is
required to maintain offsite radiological doses well within
regulatory limits.
Allowing the security posts and monitoring requirements of PSP,
Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously
maintained temporarily, under emergency conditions, does not create
problems that could increase the consequences of an accident. The
primary function of the manning and monitoring requirements of PSP,
Sections 3.1.4 and 4.3, and Table 7-1, is to monitor, detect and
assess unauthorized intrusion into the protected area, and has
nothing to do with the probability or consequences of plant
accidents.
If security personnel evacuate PSP, Section 3.1.4 and Table 7-1,
security posts during a tsunami, those security personnel will be
able to return to the PSP, Section 3.1.4 and Table 7-1, security
posts after the tsunami and assess damage or intrusion by observing
alarms and/or physical conditions as well as resume implementation
of security post and monitoring requirements of PSP, Sections 3.1.4
and 4.3, and Table 7-1. In addition, upon evacuation, security
personnel notify offsite security backup personnel of the evacuation
and the need for the offsite personnel to remotely monitor HBPP
security system alarms. Conversely, if security personnel remain at
the PSP, Section 3.1.4 and Table 7-1, security posts during a
tsunami and become injured, those security personnel would be unable
to assist in the resumption of implementation of security post and
monitoring requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-
1. Therefore, not continually manning the PSP, Section 3.1.4 and
Table 7-1, security posts during a tsunami does not increase the
consequences of the tsunami.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
As discussed in the response to Question 1 above, none of the
analyzed accidents require security personnel action to keep offsite
radiological doses well within regulatory limits. In addition,
allowing security personnel to not continuously maintain security
post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and
Table 7-1, after an emergency situation has occurred has no impact
on the possibility of a new or different kind of accident from
occurring. The primary function of the manning and monitoring
requirements of PSP, Sections 3.1.4 and 4.3, and Table 7-1, is to
monitor, detect, and assess unauthorized intrusion into the
protected area, and has nothing to do with the possibility of a
different kind of plant accident occurring.
(3) Does the change involve a significant reduction in a margin
of safety?
Response: No.
NRC SER, Section 10.8, ``Accident Analysis Conclusions,''
summarizes the consequences from accidents in terms of offsite
radiological doses. SER, Section 10.8, includes the statement, ``The
(NRC) staff has determined that offsite radiological consequences
due to a tsunami are within acceptable dose guideline values.'' As
discussed in the response to Question 1 above, none of the analyzed
accidents require security personnel action to keep offsite
radiological doses well within regulatory limits. Therefore,
allowing security personnel to not continuously maintain security
post and monitoring requirements of PSP, Sections 3.1.4 and 4.3, and
Table 7-1, after an emergency situation has occurred has no impact
on 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: Mr. Antonio Fern[aacute]ndez, Esquire,
Pacific Gas & Electric Company, Post Office Box 7442, San Francisco, CA
94120.
NRC Branch Chief: Claudia Craig.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California
Date of amendment request: December 20, 2006.
Description of amendment request: The licensee has proposed to
amend the
[[Page 6789]]
Facility Operating License by deleting paragraph 2.B.3(c), and
replacing it with a new paragraph 2.B.4 to read as follows: ``Pursuant
to the Act and Title 10, CFR, Chapter I, Parts 30, 40, and 70, to
receive, possess, and use in amounts as required any byproduct, source,
or special nuclear material without restriction to chemical or physical
form, for sample analysis or instrument calibration or associated with
radioactive apparatus or components.''
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 eliminates a restriction regarding the type
and limits of byproduct and special nuclear material to be received,
possessed, and used onsite. However, in the proposed change, the
type or amount of byproduct, source, or special nuclear material to
be received, possessed, or used 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 eliminates a restriction regarding the
limits and type of byproduct and special nuclear material to be
received, possessed, and used onsite. 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. Temporary equipment brought onsite for decommissioning
activities would still be required to be operated in accordance with
plant procedures and licensing bases documents, regardless of the
byproduct material content. 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 eliminates a restriction regarding the limit
and type of byproduct and special nuclear material to be received,
possessed, and used onsite. The proposed change has no effect on
existing plant equipment, operating practices, or safety analysis
assumptions. Temporary equipment brought onsite for decommissioning
activities would still be required to be operated in accordance with
plant procedures and licensing bases documents, regardless of the
byproduct material content. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The U.S. Nuclear Regulatory Commission 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 Fern[aacute]ndez, Esquire,
Pacific Gas & Electric Company, Post Office Box 7442, San Francisco, CA
94120.
NRC Branch Chief: Claudia Craig.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: November 15, 2006.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) Table 3.6.3-1, ``Primary
Containment Isolation Valves,'' and relocate the information to the
Technical Requirements Manual. The amendment would also revise other TS
sections that reference TS Table 3.6.3-1. The proposed changes are
based on the guidance in Generic Letter 91-08, ``Removal of Component
Lists from Technical Specifications.''
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 relocation of Technical Specification component
lists of primary containment isolation valves does not alter the
requirements for component operability or surveillance currently in
the Technical Specifications. The proposed change to remove the
component lists from TS and relocate the information to an
administratively controlled document will have no impact on any
safety related structures, systems or components.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of
accidents previously evaluated in the UFSAR [Updated Final Safety
Analysis Report] are not affected because the ability of the
components to perform their required function is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are administrative in nature, conform to
the guidance in Generic Letter 91-08 and do not result in physical
alterations or changes in the method by which any safety related
system performs its intended function. The proposed changes do not
affect any safety analysis assumptions. The proposed changes do not
create any new accident initiators or involve an activity that could
be an initiator of an accident of a different type.
All components will continue to be tested to the same
requirements as defined in the Technical Specification Surveillance
Requirements. The proposed revision does not make changes in any
method of testing or how any safety related system performs its
safety functions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to remove Technical Specification Table
3.6.3-1 from the Technical Specifications and relocate it to the
Technical Requirements Manual does not alter the Technical
Specification requirements for containment integrity and containment
isolation and will not affect the containment isolation capability.
Future revisions to the Technical Requirements Manual Table will be
subject to evaluation pursuant to 10 CFR 50.59 [Title 10 of the Code
of Federal Regulations (10 CFR), Section 50.59].
The proposed change will not affect the current Technical
Specification requirements or the components to which they apply.
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.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of amendment request: April 12, 2006, and supplemented
November 21, 2006.
Description of amendment request: The licensee has proposed to
amend its
[[Page 6790]]
license to incorporate a new license condition addressing the license
termination plan (LTP). This amendment will document the approval of
the LTP, document the criteria for making changes to the LTP which will
and will not require pre-approval by the NRC, and will document any
conditions imposed with the approval of the LTP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed change is administrative. The change allows for
the approval of the LTP and provides the criteria for when changes
to the LTP require prior U.S. Nuclear Regulatory Commission (NRC)
approval. This change does not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operation of the facility. Safety limits, limiting safety system
settings, and limiting control systems are no longer applicable to
Rancho Seco in the permanently defueled mode, and are therefore not
relevant.
The proposed change does not affect the boundaries used to
evaluate compliance with liquid or gaseous effluent limits, and has
no impact on plant operations. Therefore, the proposed license
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. As described above, the proposed change is administrative
and provides the criteria for when changes to the LTP require prior
NRC approval. The safety analysis for the facility remains complete
and accurate. There are no physical changes to the facility as a
result of the proposed amendment and the plant conditions for which
the design basis accidents have been evaluated are still valid.
The operating procedures and emergency procedures are not
affected. The proposed changes do not affect the emergency planning
zone, the boundaries used to evaluate compliance with liquid or
gaseous effluent limits, and have no impact on plant operations.
Consequently, no new failure modes are introduced as the result of
the proposed changes. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
No. As described above, the proposed changes are administrative.
There are no changes to the design or operation of the facility. The
proposed changes do not affect the emergency planning zone, the
boundaries used to evaluate compliance with liquid or gaseous
effluent limits, and have no impact on plant operations.
Accordingly, neither the design basis nor the accident assumptions
in the Defueled Safety Analysis Report, nor the Technical
Specification Bases are affected. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's significant hazards
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: Arlen Orchard, Esq., General Counsel,
Sacramento Municipal Utility District, 6201 S Street, P.O. Box 15830,
Sacramento, CA 95817-1899.
NRC Branch Chief: Claudia M. Craig.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston
County, Alabama
Date of amendment request: January 30, 2007.
Description of amendment request: The proposed amendment would
revise the Farley Nuclear Plant, Units 1 and 2, Technical
Specifications (TSs) to reflect a change to a site vice president
organizational structure. The resulting structure places a vice
president at the plant site. The proposed amendment describes changes
in titles and administrative duties that accompany the reorganization.
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 proposed change to [the] FNP TS involves SNC moving to a
site vice president organizational structure. Since the proposed
change is administrative in nature, it does not involve any physical
changes to any structures, systems, or components, nor will their
performance requirements be altered. The proposed change also does
not affect the operation, maintenance, or testing of the plant.
Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, 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 previously evaluated?
As a result of the proposed change to the FNP TS, the
qualification requirements for the unit staff position[s] will
remain unchanged and the plant staff will continue to meet
applicable regulatory requirements. Also, since no change is being
made to the design, operation, maintenance, or testing of the plant,
no new methods of operation or failure modes are introduced by the
proposed change. Therefore, the possibility of a new or different
kind of accident from any previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The proposed change to the FNP TS will have no adverse impact on
the onsite organizational features necessary to assure safe
operation of the plant since the qualification requirements for the
unit staff remains unchanged. Since the proposed change is
administrative in nature, it does not involve any physical changes
to any structures, systems, or components, nor will their
performance requirements be altered. Therefore, the proposed change
does not involve a significant decrease 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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Evangelos C. Marinos.
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 (HNP), Units 1 and 2, Appling County, Georgia
Date of amendment request: January 30, 2007.
Description of amendment request: The proposed amendments would
revise the Hatch Nuclear Plant, Units 1 and 2, Technical Specifications
(TSs) to reflect a change to a site vice president organizational
structure. The resulting structure places a vice president at the plant
site. The proposed amendment describes changes in titles and
administrative duties that accompany the reorganization.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 6791]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to [the] HNP TS involves SNC moving to a
site vice president organizational structure. Since the proposed
change is administrative in nature, it does not involve any physical
changes to any structures, systems, or components, nor will their
performance requirements be altered. The proposed change also does
not affect the operation, maintenance, or testing of the plant.
Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, 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 previously evaluated?
As a result of the proposed change to the HNP TS, the
qualification requirements for the unit staff position[s] will
remain unchanged and the plant staff will continue to meet
applicable regulatory requirements. Also, since no change is being
made to the design, operation, maintenance, or testing of the plant,
no new methods of operation or failure modes are introduced by the
proposed change. Therefore, the possibility of a new or different
kind of accident from any previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The proposed change to the HNP TS will have no adverse impact on
the onsite organizational features necessary to assure safe
operation of the plant since the qualification requirements for the
unit staff remains unchanged. Since the proposed change is
administrative in nature, it does not involve any physical changes
to any structures, systems, or components, nor will their
performance requirements be altered. Therefore, the proposed change
does not involve a significant decrease 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Branch Chief: Evangelos C. Marinos.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Burke
County, Georgia
Date of amendment request: January 30, 2007.
Description of amendment request: The proposed amendment would
revise the Vogle Electric Generating Plant, Units 1 and 2, Technical
Specifications (TSs) to reflect a change to a site vice president
organizational structure. The resulting structure places a vice
president at the plant site. The proposed amendment describes changes
in titles and administrative duties that accompany the reorganization.
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 proposed change to [the] VEGP TS involves SNC moving to a
site vice president organizational structure. Since the proposed
change is administrative in nature, it does not involve any physical
changes to any structures, systems, or components, nor will their
performance requirements be altered. The proposed change also does
not affect the operation, maintenance, or testing of the plant.
Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, 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 previously evaluated?
As a result of the proposed change to the VEGP TS, the
qualification requirements for the unit staff position[s] will
remain unchanged and the plant staff will continue to meet
applicable regulatory requirements. Also, since no change is being
made to the design, operation, maintenance, or testing of the plant,
no new methods of operation or failure modes are introduced by the
proposed change. Therefore, the possibility of a new or different
kind of accident from any previously evaluated is not created.
3. Does the proposed change involve a significant decrease in
the margin of safety?
The proposed change to the VEGP TS will have no adverse impact
on the onsite organizational features necessary to assure safe
operation of the plant since the qualification requirements for the
unit staff remains unchanged. Since the proposed change is
administrative in nature, it does not involve any physical changes
to any structures, systems, or components, nor will their
performance requirements be altered. Therefore, the proposed change
does not involve a significant decrease 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: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: December 21, 2006 (TS-456).
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.10.1 and the associated TS Bases to expand its scope to include
provisions for temperature excursions greater than 212 [deg]F as a
consequence of inservice leak and hydrostatic testing, and as a
consequence of scram time testing initiated in conjunction with
inservice leak or hydrostatic testing, while considering operational
conditions to be in Mode 4.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 21, 2006 (71 FR 48561), on possible
amendments to revise the plant-specific TS, to expand the scope of TS
LCO 3.10.1, 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, including a model
safety evaluation and model No Significant Hazards Consideration (NSHC)
Determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on October 27, 2006 (71 FR 63050). The licensee affirmed the
applicability of the model NSHC determination in its application dated
December 21, 2006.
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.
Technical Specifications currently allow for operation at
greater than [200] [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
[[Page 6792]]
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 [200] [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 types of
equipment will be installed) or a change in the methods governing
normal plant operation. In addition, the changes 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 [200] [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 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
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.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: July 27, 2006, as supplemented by
letters dated October 4 and October 9, 2006.
Brief description of amendment request: The proposed amendment
would revise Technical Specification (TS) 3.7.14, ``Spent Fuel Pool
Boron Concentration,'' TS 3.7.15, ``Spent Fuel Pool Storage,'' and the
associated Figure 3.7.15-1, and TS 4.3, ``Fuel Storage,'' and the
associated Figure 4.3.1.2-1. In addition, this amendment would add TS
5.5.17, ``Metamic Coupon Sampling Program,'' and Surveillance
Requirement 3.7.15.2 that directs the performance of the coupon
sampling program. The proposed TS changes support a modification to the
ANO-1 spent fuel pool (SFP) that would utilize Metamic[supreg] poison
insert assemblies. In addition to the proposed plant modification, the
licensee would increase the SFP boron concentration and credit boron to
ensure that a 5-percent subcriticality margin is maintained during
normal and accident conditions. This proposed amendment also would
increase the allowable initial fuel assembly uranium-235 (U-235)
enrichment from 4.1 weight percent (wt%) to a maximum U-235 enrichment
of 4.95 wt%.
Date of publication of individual notice in Federal Register:
December 26, 2006 (71 FR 77414).
Expiration date of individual notice: February 26, 2007.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: January 26, 2006, as
supplemented by letter dated December 20, 2006.
Brief description of amendment: The amendment revised the Millstone
Power Station, Unit No. 2 Technical Specifications (TSs) to update the
list of NRC-approved documents specified in the TSs that describe the
analytical methods used to determine the core operating limits. The
proposed change also corrects a typographical error in TS 5.3.1,
``Reactor Core, Fuel Assembly,'' which was introduced in the retyped
pages provided to the NRC for issuance of Amendment No. 280, dated
September, 25, 2003.
Date of issuance: January 23, 2007.
[[Page 6793]]
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 295.
Facility Operating License Nos. DPR-65: The Amendment revised the
TSs.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
26997). The supplement dated December 20, 2006, provided clarifying
information that did not change the scope of the proposed amendment as
described in the original notice of proposed action published in the
Federal Register, and did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 23, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: March 17, 2006.
Brief description of amendment: The amendment changed the Millstone
Power Station, Unit No. 2, Technical Specifications by replacing the
existing maximum and minimum pressurizer water volume and water level
limits with a maximum water level limit. The associated TS bases were
updated to address the proposed changes.
Date of issuance: January 30, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 296.
Facility Operating License No. DPR-65: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 11, 2006 (71
FR 65141).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 2007.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1 (ANO-1), Pope County, Arkansas
Date of amendment request: July 27, 2006, as supplemented by
letters dated October 4, October 9, and December 14, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.14, ``Spent Fuel Pool Boron Concentration,'' TS
3.7.15, ``Spent Fuel Pool Storage,'' and the associated Figure 3.7.15-
1, and TS 4.3, ``Fuel Storage,'' and the associated Figure 4.3.1.2-1.
In addition, this amendment added TS 5.5.17, ``Metamic Coupon Sampling
Program,'' and Surveillance Requirement 3.7.15.2 that directs the
performance of the coupon sampling program. The TS changes support a
modification to the ANO-1 spent fuel pool (SFP) that utilize
Metamic[supreg] poison insert assemblies. In addition to the proposed
plant modification, the licensee increased the SFP boron concentration
and credited boron to ensure that a 5-percent subcriticality margin is
maintained during normal and accident conditions. This amendment also
increased the allowable initial fuel assembly uranium-235 (U-235)
enrichment from 4.1 weight percent (wt%) to a maximum U-235 enrichment
of 4.95 wt%.
Date of issuance: January 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 228.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications/license.
Date of initial notice in Federal Register: December 26, 2006 (71
FR 77414). The supplement dated 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 amendment is contained in a Safety Evaluation dated January 26,
2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-374, LaSalle County
Station, Unit 2, LaSalle County, Illinois
Date of application for amendments: April 21, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.13, ``Primary Containment Leakage Testing
Program,'' to reflect a one-time extension of the LaSalle, Unit 2
primary containment Type A integrated leak rate test (ILRT) from the
current requirement of no later than December 7, 2008, to prior to
startup following the 12th LaSalle, Unit 2 refueling outage.
Date of issuance: January 24, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 166.
Facility Operating License No. NPF-18: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: June 6, 2006 (71 FR
32605). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 24, 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 13, 2005, as supplemented
by letters dated December 22, 2005, June 12, 2006, and January 4, 2007.
Brief description of amendments: The proposed amendment would
extend, on a one-time basis, the completion time (CT) for required
action C.4, ``Restore required Diesel Generators (DGs) OPERABLE
status,'' associated with Technical Specification (TS) Section 3.8.1
from 72 hours to 6 days. This proposed change would only be used during
the upcoming Unit 2--spring 2007 refueling outage, and later during the
Unit 1--spring 2008 refueling outage. The amendment would also extend
the CT from 2 hours to 6 hours in TS Section 3.8.1, Required Action
F.1, ``Restore one required DG to OPERABLE status.'' This proposed
change to be used during the upcoming Unit 2--spring 2007 refueling
outage, and later during the subsequent Unit 1--spring 2008 refueling
outage.
Date of issuance: January 29, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 180/167.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications/License.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33210). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 29, 2007.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: April 24, 2006, as supplemented
September 14, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) consistent with the NRC-approved Revision 4 to TS
Task Force (TSTF) Standard TS Change
[[Page 6794]]
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
Date of Issuance: January 30, 2007.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 200.
Renewed Facility Operating License No. DPR-67: Amendment revised
the TSs.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40746). The September 14, 2006, supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated: January 30, 2007.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: December 16, 2005, as
supplemented by letter dated October 25, 2006.
Brief description of amendment: The amendment relocates Technical
Specification Surveillance Requirement 4.1.4d for core spray header
differential pressure instrumentation to the Updated Final Safety
Analysis Report.
Date of issuance: January 31, 2007.
Effective date: January 31, 2007.
Amendment No.: 192.
Facility Operating License No. DPR-63: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15484). The supplemental letter dated October 25, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2007.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: January 25, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 1.1, ``Definitions,'' and TS 3.4.16, ``RCS [Reactor
Coolant System] Specific Activity.'' The amendments replaced the
current TS 3.4.16 limit on RCS gross specific activity with a new limit
on RCS noble gas specific activity. The noble gas specific activity
limit is based on a new dose equivalent Xe-133 definition that would
replace the current E-Bar average disintegration energy definition. In
addition, the current dose equivalent I-131 definition is revised to
allow the use of alternate thyroid dose conversion factors.
Date of issuance: January 19, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-192; Unit 2-193.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications and Operating Licenses.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13176). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 19, 2007.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: October 13, 2005, as
supplemented on May 18, September 15 (PLA-6112 and PLA-6114), September
29, October 20, November 14, December 13, and December 14, 2006.
Brief description of amendments: The amendments revise the SSES 1
and 2 Technical Specifications (TSs) to incorporate a full-scope
application of an alternate source term methodology in accordance with
Title 10 of the Code of Federal Regulations, section 50.67.
Date of issuance: January 31, 2007.
Effective date: As of the date of issuance and to be implemented by
October 30, 2007.
Amendment Nos.: 239 and 216.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and license.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51231). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 31, 2007.
The supplements dated September 15 (PLA-6112 and PLA-6114),
September 29, October 20, November 14, December 13, 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.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: October 26, 2006 (TS-457).
Brief description of amendments: The amendments revise Technical
Specification (TS) Action 3.8.1.B.4 for Browns Ferry Nuclear Plant
Units 2 and 3. The revision changes the restoration time of an
inoperable Emergency Diesel Generator from 14 to 7 days.
Date of issuance: January 26, 2007.
Effective date: Within 60 days of NRC approval or prior to changing
Unit 1 reactor mode to startup, whichever is earlier.
Amendment Nos.: 298 and 256.
Renewed Facility Operating License Nos. DPR-52 and DPR-68:
Amendments revised the TSs.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67398).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 26, 2007.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 25, 2006.
Brief description of amendment: The amendment revised TSs by adding
Limiting Condition for Operation (LCO) 3.0.8. This change is consistent
with NRC-approved Revision 4 to Technical Specification Task Force
(TSTF) Standard Technical Specification Traveler, TSTF-372, ``Addition
of LCO 3.0.8, Inoperability of Snubbers.''
Date of issuance: January 31, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 179.
Facility Operating License No. NPF-30: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40755).
The Commission's related evaluation of the amendment is contained
in a
[[Page 6795]]
Safety Evaluation dated January 31, 2007.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: May 22, 2006.
Brief description of amendment: These amendments revise the
existing steam generator tube surveillance program to be consistent
with the Technical Specification Task Force (TSTF) Standard TS Change
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
Date of issuance: October 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 248, 228.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43537)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 16, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of February 2007.
For the Nuclear Regulatory Commission.
John W. Lubinski,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-2323 Filed 2-12-07; 8:45 am]
BILLING CODE 7590-01-P